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Sample records for element zirconium cladding

  1. 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

  2. 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)

  3. 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

  4. Electrochemical Studies on Important Elements for Zirconium Recovery Form Irradiated Zircaloy-4 Cladding

    International Nuclear Information System (INIS)

    Park, J.; Sohn, S.; Hwang, I.S.

    2015-01-01

    Since Zircaloy cladding accounts for about 16 wt. % of used nuclear fuel assembly, decontamination process is required to reduce the final waste volume from spent nuclear fuel. To develop Zircaloy-4 electrorefining process as an irradiated Zircaloy cladding decontamination process, electrochemical studies on Sn, Cr, Fe and Co which are major or important elements in the irradiated cladding were conducted based on cyclic voltammetry in LiCl-KCl at 500 deg. C. Cyclic voltammetry for Sn, Fe, Cr and Co elements that should be eliminated was conducted and revealed that redox reactions of these ions are much simpler than Zr and more reductive than Zr. The reliability of cyclic voltammetry was verified by comparing diffusion coefficients and formal reduction potentials of these ions obtained in this study to previous studies. (authors)

  5. Zirconium alloy fuel cladding resistant to PCI crack propagation

    International Nuclear Information System (INIS)

    Boyle, R.F.; Foster, J.P.

    1987-01-01

    A nuclear fuel element is described cladding tube comprising: concentric tubular layers of zirconium base alloys; the concentric tubular layers including an inner layer and outer layer; the outer layer metallurgically bonded to the inner layer; the outer layer composed of a first zirconium base alloy characterized by excellent resistance to corrosion caused by exposure to high temperature and pressure aqueous environments; the inner layer composed of a second zirconium base alloy consisting of: about 0.2 to 0.6 wt.% tin, about 0.03 to 0.11 wt.% iron, less than about 0.02 wt.% chromium, up to about 350 ppm oxygen and the remainder being zirconium and incidental impurities, and the inner layer characterized by improved resistance to crack propagation under reactor operating conditions compared to the first zirconium alloy

  6. Mechanism for iodine cracking of zirconium claddings

    International Nuclear Information System (INIS)

    Novikov, V.V.

    1991-01-01

    The mechanism of iodine cracking of zirconium cladding is analyzed taking into account the effect of stresses on diffusion. A decisive effect of the stress gradiemt on crack propagation in an agressive medium is shown. The experimental data are compared with the proposed model

  7. Fuel elements (uranium clad with zirconium) produced by co-extrusion; Les elements combustibles uranium gaine de zirconium obtenus par coextrusion

    Energy Technology Data Exchange (ETDEWEB)

    Montagne, R.; Winogradzki, A.; Sauve, C.; Buffet, J. [Commissariat a l' energie atomique et aux energies alternatives - CEA, Departement de Metallurgie et de Chimie Appliquee (France)

    1959-07-01

    In this paper a description is given of a process for making fuel elements for atomic reactors. Contact of the most intimate possible kind is achieved between the fuel and the sheath by the simultaneous extrusion of the two elements. Genuine welding between the two metals is thus effected. This can be subsequently improved by a heat-treatment which causes diffusion. Tests made on these co-extruded elements are described in the paper. Reprint of a paper published in Revue de Metallurgie, LV, no. 11, 1958.

  8. Automatic measuring system of zirconium thickness for zirconium liner cladding tubes

    International Nuclear Information System (INIS)

    Matsui, K.; Yamaguchi, H.; Hiroshima, T.; Sakamoto, T.; Murayama, R.

    1985-01-01

    An automatic system of pure zirconium liner thickness for zirconium-zircaloy cladding tubes has been successfully developed. The system consists of three parts. (1) An ultrasonic thickness measuring method for mother tubes before cold rolling. (2) An electromagnetic thickness measuring method for the manufactured tubes. (3) An image processing method for the cross sectional view of the manufactured cut tube samples. In Japanese nuclear industry, zirconium-zircaloy cladding tubes have been tested in order to realize load following operation in the atomic power plant. In order to provide for the practical use in the near future, Sumitomo Metal Industries, Ltd. has been studied and established the practical manufacturing process of the zirconium liner cladding tubes. The zirconium-liner cladding tube is a duplex tube comprising an inner layer of pure zirconium bonded to zircaloy metallurgically. The thickness of the pure zirconium is about 10 % of the total wall thickness. Several types of the automatic thickness measuring methods have been investigated instead of the usual microscopic viewing method in which the liner thickness is measured by the microscopic cross sectional view of the cut tube samples

  9. 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

  10. 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

  11. 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

  12. 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

  13. Thermal creep behavior of N36 zirconium alloy cladding tube

    International Nuclear Information System (INIS)

    Wang, P.; Zhao, W.; Dai, X.

    2015-01-01

    N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

  14. 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.

  15. 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.

  16. 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

  17. Delayed hydride cracking of zirconium alloy fuel cladding

    International Nuclear Information System (INIS)

    2010-10-01

    This report describes the work performed in a coordinated research project on Hydrogen and Hydride Degradation of the Mechanical and Physical Properties of Zirconium Alloys. It is the second in the series. In 2005-2009 that work was extended within a new CRP called Delayed Hydride Cracking in Zirconium Alloy Fuel Cladding. The project consisted of adding hydrogen to samples of Zircaloy-4 claddings representing light water reactors (LWRs), CANDU and Atucha, and measuring the rates of delayed hydride cracking (DHC) under specified conditions. The project was overseen by a supervisory group of experts in the field who provided advice and assistance to participants as required. All of the research work undertaken as part of the CRP is described in this report, which includes details of the experimental procedures that led to a consistent set of data for LWR cladding. The participants and many of their co-workers in the laboratories involved in the CRP contributed results and material used in this report, which compiles the results, their analysis, discussions of their interpretation and conclusions and recommendations for future work. The research was coordinated by an advisor and by representatives in three laboratories in industrialized Member States. Besides the basic goal to transfer the technology of the testing technique from an experienced laboratory to those unfamiliar with the methods, the CRP was set up to harmonize the experimental procedures to produce consistent sets of data, both within a single laboratory and between different laboratories. From the first part of this project it was demonstrated that by following a standard set of experimental protocols, consistent results could be obtained. Thus, experimental vagaries were minimized by careful attention to detail of microstructure, temperature history and stress state in the samples. The underlying idea for the test programme was set out at the end of the first part of the project on pressure tubes. The

  18. Creep behavior under internal pressure of zirconium alloy cladding oxidized in steam at high temperature

    International Nuclear Information System (INIS)

    Chosson, Raphael

    2014-01-01

    During hypothetical Loss-Of-Coolant-Accident (LOCA) scenarios, zirconium alloy fuel cladding tubes creep under internal pressure and are oxidized on their outer surface at high temperature (HT). Claddings become stratified materials: zirconia and oxygen-stabilized α phase, called α(O), are formed on the outer surface of the cladding whereas the inner part remains in the β domain. The strengthening effect of oxidation on the cladding creep behavior under internal pressure has been highlighted at HT. In order to model this effect, the creep behavior of each layer had to be determined. This study focused on the characterization of the creep behavior of the α(O) phase at HT, through axial creep tests performed under vacuum on model materials, containing from 2 to 7 wt.% of oxygen and representative of the α(O) phase. For the first time, two creep flow regimes have been observed in this phase. Underlying physical mechanisms and relevant microstructural parameters have been discussed for each regime. The strengthening effect due to oxygen on the α(O) phase creep behavior at HT has been quantified and creep flow equations have been identified. A ductile to brittle transition criterion has been also suggested as a function of temperature and oxygen content. Relevance of the creep flow equations for each layer, identified in this study or from the literature, has been discussed. Then, a finite element model, describing the oxidized cladding as a stratified material, has been built. Based on this model, a fraction of the experimental strengthening during creep is predicted. (author) [fr

  19. 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.

  20. 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.

  1. 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

  2. Management of waste cladding hulls. Part II. An assessment of zirconium pyrophoricity and recommendations for handling waste hulls

    International Nuclear Information System (INIS)

    Kullen, B.J.; Levitz, N.M.; Steindler, M.J.

    1977-11-01

    This report reviews experience and research related to the pyrophoricity of zirconium and zirconium alloys. The results of recent investigations of the behavior of Zircaloy and some observations of industrial handling and treatment of Zircaloy tubing and scrap are also discussed. A model for the management of waste Zircaloy cladding hulls from light water reactor fuel reprocessing is offered, based on an evaluation of the reviewed information. It is concluded that waste Zircaloy cladding hulls do not constitute a pyrophoric hazard if, following the model flow sheet, finely divided metal is oxidized during the management procedure. Steps alternative to the model are described which yield zirconium in deactivated form and also accomplish varying degrees of transuranic decontamination. Information collected into appendixes is (1) a collation of zirconium pyrophoricity data from the literature, (2) calculated radioactivity contents in Zircaloy cladding hulls from spent LWR fuels, and (3) results of a laboratory study on volatilization of zirconium from Zircaloy using HCl or Cl 2

  3. Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

    2014-08-15

    Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

  4. Effect of high hydrogen content on metallurgical and mechanical properties of zirconium alloy claddings after heat-treatment at high temperature

    International Nuclear Information System (INIS)

    Turque, Isabelle

    2016-01-01

    Under hypothetical loss-of-coolant accident conditions, fuel cladding tubes made of zirconium alloys can be exposed to steam at high temperature (HT, up 1200 C) before being cooled and then quenched in water. In some conditions, after burst occurrence the cladding can rapidly absorb a significant amount of hydrogen (secondary hydriding), up to 3000 wt.ppm locally, during steam exposition at HT. The study deals with the effect, poorly studied up to date, of high contents of hydrogen on the metallurgical and mechanical properties of two zirconium alloys, Zircaloy-4 and M5, during and after cooling from high temperatures, at which zirconium is in its β phase. A specific facility was developed to homogeneously charge in hydrogen up to ∼ 3000 wt.ppm cladding tube samples of several centimeters in length. Phase transformations, chemical element partitioning and hydrogen precipitation during cooling from the β temperature domain of zirconium were studied by using several techniques, for the materials containing up to ∼ 3000 wt.ppm of hydrogen in average: in-situ neutron diffraction upon cooling from 700 C, X-ray diffraction, μ-ERDA, EPMA and electron microscopy in particular. The results were compared to thermodynamic predictions. In order to study the effect of high hydrogen contents on the mechanical behavior of the (prior-)μ phase of zirconium, axial tensile tests were performed at various temperatures between 20 and 700 C upon cooling from the β temperature domain, on samples with mean hydrogen contents up to ∼ 3000 wt.ppm. The results show that metallurgical and mechanical properties of the (prior-)β phase of zirconium alloys strongly depend on temperature and hydrogen content. (author) [fr

  5. Zirconium cladding - the long way towards a mechanistic understanding of processing and performance

    International Nuclear Information System (INIS)

    Preuss, Michael

    2011-01-01

    Zirconium alloys are the material of choice to encapsulate nuclear fuel in light and heavy water-cooled reactors due to their low neutron absorption, excellent corrosion resistance and sufficient mechanical properties. Despite these advantageous physical and mechanical properties a more physically based understanding of microstructure and texture evolution during processing is highly desirable in order to improve our understanding of formability during thermomechanical processing and performance variability of cladding material. In addition, the purely empirical understanding of aqueous zirconium corrosion, hydrogen pick up, hydride precipitation as well as irradiation growth and creep limits the accuracy of life predictions and therefore the level of burnup that is obtained from current fuel assemblies. The presentation aims at giving examples of new research strategies that will enable the development of a new physical understanding of processing and performance aspects in zirconium cladding material, which is required to develop new predictive models. Particular emphasis will be placed on using novel research tools and large-scale research facilities such as neutron spallation and synchrotron radiation sources to undertake very detailed and often in-situ studies of deformation mechanisms and microstructure evolution as well as determining stress states in grain families, oxides and hydrides. The results will be presented in the view of how they might help us to improve our understanding and enable the development of better predictive models

  6. Development of new zirconium alloys for PWR fuel rod claddings

    International Nuclear Information System (INIS)

    Zhao Wenjin; Zhou Bangxin; Miao Zhi; Li Cong; Jiang Hongman; Yu Xiaowei; Jiang Yourong; Huang Qiang; Gou Yuan; Huang Decheng

    2001-01-01

    An advanced zirconium alloys containing Sn, Nb, Fe and Cr have been developed. The relationships between manufacturing, microstructure and corrosion performance for the new alloys have been studied. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in lithia water at 633 K and high-temperature steam at 773 K. Analytical electron microscopy demonstrated that the best out-of-pile corrosion performance was obtained for microstructure containing a fine and uniform distribution of β-Nb and Zr(Fe, Nb) 2 particles. Autoclave testing in LiOH solution indicated that two kinds of alloys (N18, N36) showed the lower corrosion rate than the reference Zr-4 tested, and especially, the corrosion resistance in superheated steam at 773 K was much better. Moreover, the mechanical properties were superior to Zr-4. And the hydrogen absorption data for all of alloys from corrosion reactions under various corrosion conditions showed a linear increase with the oxide thickness

  7. Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar [University of Wisconsin-Madison; Mariani, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States); Xu, Peng [Westinghouse Electric Company; Lahoda, Ed [Westinghouse Electric Company

    2018-03-31

    Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi2 coating) during clad quenching experiments is discussed in detail. Oxidation of bulk ZrSi2 was found to be negligible compared to Zircaloy-4 (a common Zr-alloy cladding material) and mechanical integrity of ZrSi2 was superior to that of bulk Zr2Si at high temperatures in ambient air. Very interesting and unique multi-nanolayered composite of ZrO2 and SiO2 were observed. Physical model for the oxidation has been proposed wherein Zr–Si–O mixture undergoes a spinodal phase decomposition into ZrO2 and SiO2, which is manifested as a nanoscale assembly of alternating layer of the two oxides. Steam corrosion at high pressure (10.3 MPa) led to weight loss of ZrSi2 and produced oxide scale with depletion of silicon, possibly attributed to volatile silicon hydroxide, gaseous silicon monoxide, and a solubility of silicon dioxide in water. Only Zircon phase (ZrSiO4) formed during oxidation of ZrSi2 at 1400°C in air, and allowed for immobilization silicon species in oxide scale in the aqueous environments. Zirconium-silicide coatings (on zirconium-alloy substrates) investigated in this study were deposited primarily using magnetron sputter deposition method and slurry method, although powder spray deposition processes cold spray and thermal spray methods were also investigated. The optimized ZrSi2 sputtered coating exhibited a highly protective nature at elevated temperatures in ambient air by mitigating oxygen permeation to the underlying zirconium alloy substrate. The high oxidation resistance of the coating has been shown to be due to nanocrystalline SiO2 and ZrSiO4 phases in the amorphous

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Louthan, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); PNNL, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-05-29

    This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to

  10. 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)

  11. Phase evolution and dendrite growth in laser cladding of aluminium on zirconium

    International Nuclear Information System (INIS)

    Yue, T.M.; Xie, H.; Lin, X.; Yang, H.O.

    2011-01-01

    Research highlights: → Laser cladding of Al on pure Zr. → A series of phase evolutions occurred across the laser-clad coating. → Epitaxial crystal growth, backward dendrite growth and two-phase eutectic dendritic growth. → Phase and microstructure evolution is discussed. - Abstract: Aluminium was laser clad on a pure zirconium substrate using the blown powder method. The microstructure across the laser-clad coating was studied. Starting from the bottom to the top surface of the coating, a series of phase evolutions had occurred: (Zr) → (Zr) + AlZr 2 + AlZr 3 → Al 4 Zr 5 + Al 3 Zr 2 → Al 3 Zr 2 + AlZr 2 → Al 2 Zr → Al 2 Zr + Al 3 Zr. This resulted in an epitaxial columnar crystal growth at the re-melt substrate boundary, a band of backward growth Al 3 Zr 2 dendrites towards the lower half of the coating, and a two-phase eutectic dendritic growth of Al 2 Zr + Al 3 Zr towards the top of the coating. The evolution of the various phases and microstructures is discussed in conjunction with the Al-Zr phase diagram, the criteria for planar interface instability, and the theory of eutectic growth under rapid solidification conditions (the TMK model).

  12. Application of non-destructive liner thickness measurement technique for manufacturing and inspection process of zirconium lined cladding tube

    International Nuclear Information System (INIS)

    Nakazawa, Norio; Fukuda, Akihiro; Fujii, Noritsugu; Inoue, Koichi

    1986-01-01

    Recently, in order to meet the difference of electric power demand owing to electric power situation, large scale load following operation has become necessary. Therefore, the development of the cladding tubes which withstand power variation has been carried out, as the result, zirconium-lined zircaloy 2 cladding tubes have been developed. In order to reduce the sensitivity to stress corrosion cracking, these zirconium-lined cladding tubes require uniform liner thickness over the whole surface and whole length. Kobe Steel Ltd. developed the nondestructive liner thickness measuring technique based on ultrasonic flaw detection technique and eddy current flaw detection technique. These equipments were applied to the manufacturing and inspection processes of the zirconium-lined cladding tubes, and have demonstrated superiority in the control and assurance of the liner thickness of products. Zirconium-lined cladding tubes, the development of the measuring technique for guaranteeing the uniform liner thickness and the liner thickness control in the manufacturing and inspection processes are described. (Kako, I.)

  13. 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)

  14. The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Northwood, D.O.

    1979-01-01

    A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  16. An evaluated neutronic data file for elemental zirconium

    International Nuclear Information System (INIS)

    Smith, A.B.; Chiba, S.

    1994-09-01

    A comprehensive evaluated neutronic data file for elemental zirconium is derived and presented in the ENDF/B-VI formats. The derivation is based upon measured microscopic nuclear data, augmented by model calculations as necessary. The primary objective is a quality contemporary file suitable for fission-reactor development extending from conventional thermal to fast and innovative systems. This new file is a significant improvement over previously available evaluated zirconium files, in part, as a consequence of extensive new experimental measurements reported elsewhere

  17. Separate-effect tests on zirconium cladding degradation in air ingress situations

    Energy Technology Data Exchange (ETDEWEB)

    Duriez, C. [Institut de Radioprotection et de Surete Nucleaire, IRSN, Direction de Prevention des Accidents Majeurs, Centre de Cadarache, 13115 St Paul Lez Durance (France)], E-mail: christian.duriez@irsn.fr; Steinbrueck, M. [Forschungszentrum Karlsruhe, FZK, Institut fuer Materialforschung, Postfach 3640, 76021 Karlsruhe (Germany); Ohai, D.; Meleg, T. [Institute for Nuclear Research, INR, Nuclear Material and Corrosion Department, Pitesti, 115400 Mioveni Arges (Romania); Birchley, J.; Haste, T. [Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)

    2009-02-15

    In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident. A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available. New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program-ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the 'as-received' state, or after pre-oxidation in steam. 'Analytical' tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 deg. C up to 1500 deg. C. Systematic post-test metallographic inspections are performed. The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of

  18. Separate-effect tests on zirconium cladding degradation in air ingress situations

    International Nuclear Information System (INIS)

    Duriez, C.; Steinbrueck, M.; Ohai, D.; Meleg, T.; Birchley, J.; Haste, T.

    2009-01-01

    In the event of air ingress during a reactor or spent fuel pond low probability accident, the fuel rods will be exposed to air-containing atmospheres at high temperatures. In comparison with steam, the presence of air is expected to result in a more rapid escalation of the accident. A state-of-the-art review performed before SARNET started showed that the existing data on zirconium alloy oxidation in air were scarce. Moreover, the exact role of zirconium nitride on the cladding degradation process was poorly understood. Regarding the cladding behaviour in air + steam or nitrogen-enriched atmospheres (encountered in oxygen-starved conditions), almost no data were available. New experimental programmes comprising small-scale tests have therefore been launched at FZK, IRSN (MOZART programme in the frame of the International Source Term Program-ISTP) and INR. Zircaloy-4 cladding in PWR (FZK, IRSN) and in CANDU (INR) geometry are investigated. On-line kinetic data are obtained on centimetre size tube segments, by thermogravimetry (FZK, IRSN and INR) or by mass spectrometry (FZK). Plugged tubes 15 cm long (FZK) are also investigated. The samples are air-oxidised either in the 'as-received' state, or after pre-oxidation in steam. 'Analytical' tests at constant temperature and gas composition provide basic kinetic data, while more prototypical temperature transients and sequential gas compositions are also investigated. The temperature domains extend from 600 deg. C up to 1500 deg. C. Systematic post-test metallographic inspections are performed. The paper gives a synthesis of the results obtained, comparing them in terms of kinetics and oxide scale structure and composition. A comparative analysis is performed with results of the QUENCH-10 (Q-10) bundle test, which included an air ingress phase. It is shown how the data contribute to a better understanding of the cladding degradation process, especially regarding the role of nitrogen. For modelling of the oxide scale

  19. Zirconium

    Science.gov (United States)

    Bedinger, G.M.

    2013-01-01

    Zirconium is the 20th most abundant element in the Earth’s crust. It occurs in a variety of rock types and geologic environments but most often in igneous rocks in the form of zircon (ZrSiO4). Zircon is recovered as a coproduct of the mining and processing of heavy mineral sands for the titanium minerals ilmenite and rutile. The sands are formed by the weathering and erosion of rock containing zircon and titanium heavy minerals and their subsequent concentration in sedimentary systems, particularly in coastal environments. A small quantity of zirconium, less than 10 kt/a (11,000 stpy), compared with total world production of 1.4 Mt (1.5 million st) in 2012, was derived from the mineral baddeleyite (ZrO2), produced from a single source in Kovdor, Russia.

  20. Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

    International Nuclear Information System (INIS)

    Mazeres, Benoit

    2013-01-01

    One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

  1. Zirconium alloy barrier having improved corrosion resistance

    International Nuclear Information System (INIS)

    Adamson, R.B.; Rosenbaum, H.S.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and is comprised of zirconium and a metal selected from the group consisting of iron, chromium, iron plus chromium, and copper. The dilute zirconium alloy liner shields the substrate from impurities or fission products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking. The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water or steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy. (author)

  2. High-resolution characterization of oxidation mechanism of zirconium nuclear fuel cladding alloys

    International Nuclear Information System (INIS)

    Hu, J.; Lozano-Perez, S.; Grovenor, C.

    2015-01-01

    Full text of publication follows. Zirconium alloys are used extensively as cladding materials in modern light water reactors to separate the uranium dioxide (UO 2 ) fuel rods and the coolant water in order to prevent the escape of radioactive fission products whilst maintaining heat transfer to the coolant. With increasing demand for high burn-up in modern nuclear reactors, environmental degradation of these alloys is now the life limiting factor for fuel assemblies. As part of the MUZIC-2 collaboration studying oxidation and hydrogen pickup in Zr alloys, several high resolution analysis techniques have been used to study the microstructure of a range of commercial and developmental Zr alloys. The sample used for this investigation was prepared from a Westinghouse TM developmental alloy with composition of Zr-0.9Nb-0.01Sn-0.08Fe (wt %) in the recrystallized condition. The sample was oxidised in an autoclave at EDF Energy under simulated PWR water conditions at 360 C. degrees for 360 days. Using Transmission Electron Microscope (TEM), we have studied the development of the equiaxed-columnar-equiaxed grain structure, and observe that the columnar grains are both longer and show a stronger preferred texture in more corrosion-resistant alloys. Fresnel imaging revealed the existence of both parallel interconnected pores and some vertically interconnected pores along the columnar oxide grain boundaries, which become more disconnected near the metal-oxide interface. Electron Energy Loss Spectroscopy (EELS) provided accurate quantitative analysis of the oxygen concentration across the interface, identifying the existence of local regions of stoichiometric ZrO and Zr 3 O 2 with varying thickness. These observations will be discussed in the context of current models for oxidation in zirconium alloys. (authors)

  3. 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.

  4. DISSOLUTION OF ZIRCONIUM AND ALLOYS THEREFOR

    Science.gov (United States)

    Swanson, J.L.

    1961-07-11

    The dissolution of zirconium cladding in a water solution of ammonium fluoride and ammonium nitrate is described. The method finds particular utility in processing spent fuel elements for nuclear reactors. The zirconium cladding is first dissolved in a water solution of ammonium fluoride and ammonium nitrate; insoluble uranium and plutonium fiuorides formed by attack of the solvent on the fuel materiai of the fuel element are then separated from the solution, and the fuel materiai is dissolved in another solution.

  5. Evaluation of steam corrosion and water quenching behavior of zirconium-silicide coated LWR fuel claddings

    Science.gov (United States)

    Yeom, Hwasung; Lockhart, Cody; Mariani, Robert; Xu, Peng; Corradini, Michael; Sridharan, Kumar

    2018-02-01

    This study investigates steam corrosion of bulk ZrSi2, pure Si, and zirconium-silicide coatings as well as water quenching behavior of ZrSi2 coatings to evaluate its feasibility as a potential accident-tolerant fuel cladding coating material in light water nuclear reactor. The ZrSi2 coating and Zr2Si-ZrSi2 coating were deposited on Zircaloy-4 flats, SiC flats, and cylindrical Zircaloy-4 rodlets using magnetron sputter deposition. Bulk ZrSi2 and pure Si samples showed weight loss after the corrosion test in pure steam at 400 °C and 10.3 MPa for 72 h. Silicon depletion on the ZrSi2 surface during the steam test was related to the surface recession observed in the silicon samples. ZrSi2 coating (∼3.9 μm) pre-oxidized in 700 °C air prevented substrate oxidation but thin porous ZrO2 formed on the coating. The only condition which achieved complete silicon immobilization in the oxide scale in aqueous environments was the formation of ZrSiO4 via ZrSi2 coating oxidation in 1400 °C air. In addition, ZrSi2 coatings were beneficial in enhancing quenching heat transfer - the minimum film boiling temperature increased by 6-8% in the three different environmental conditions tested. During repeated thermal cycles (water quenching from 700 °C to 85 °C for 20 s) performed as a part of quench tests, no spallation and cracking was observed and the coating prevented oxidation of the underlying Zircaloy-4 substrate.

  6. Modeling of mechanical behavior of quenched zirconium-based nuclear fuel claddings after a high temperature oxidation

    International Nuclear Information System (INIS)

    Cabrera-Salcedo, A.

    2012-01-01

    During the second stage of Loss Of Coolant Accident (LOCA) in Pressurized Water Reactors (PWR) zirconium-based fuel claddings undergo a high temperature oxidation (up to 1200 C), then a water quench. After a single-side steam oxidation followed by a direct quench, the cladding is composed of three layers: an oxide (Zirconia) outer layer (formed at HT), always brittle at Room Temperature (RT), an intermediate oxygen stabilized alpha layer, always brittle at RT, called alpha(O), and an inner 'prior-beta' layer, which is the only layer able to keep some significant Post Quench (PQ) ductility at RT. However, hydrogen absorbed because of service exposure or during the LOCA transient, concentrates in this layer and may leads to its embrittlement. To estimate the PQ mechanical properties of these materials, Ring Compression Tests (RCT) are widely used because of their simplicity. Small sample size makes RCTs advantageous when a comparison with irradiated samples is required. Despite their good reproducibility, these tests are difficult to interpret as they often present two or more load drops on the engineering load-displacement curve. Laboratories disagree about their interpretation. This study proposes an original fracture scenario for a stratified PQ cladding tested by RCT, and its associated FE model. Strong oxygen content gradient effect on layers mechanical properties is taken into account in the model. PQ thermal stresses resulting from water quench of HT oxidized cladding are investigated, as well as progressive damage of three layers during an RCT. The proposed scenario is based on interrupted RCT analysis, post- RCT sample's outer layers observation for damage evaluation, RCTs of prior-beta single-layer rings, and mechanical behavior of especially chemically adjusted samples. The force displacement curves appearance is correctly reproduced using the obtained FE model. The proposed fracture scenario elucidates RCTs of quenched zirconium-based nuclear fuel

  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. Fast-neutron scattering cross sections of elemental zirconium

    International Nuclear Information System (INIS)

    Smith, A.B.; Guenther, P.T.

    1982-12-01

    Differential neturon-elastic-scattering cross sections of elemental zirconium are measured from 1.5 to 4.0 MeV at intervals of less than or equal to 200 keV. Inelastic-neutron-scattering cross sections corresponding to the excitation of levels at observed energies of: 914 +- 25, 1476 +- 37, 1787 +- 23, 2101 +- 26, 2221 +- 17, 2363 +- 14, 2791 +- 15 and 3101 +- 25 keV are determined. The experimental results are interpreted in terms of the optical-statistical model and are compared with corresponding quantities given in ENDF/B-V

  9. Standard recommended practice for examination of fuel element cladding including the determination of the mechanical properties

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Guidelines are provided for the post-irradiation examination of fuel cladding and to achieve better correlation and interpretation of the data in the field of radiation effects. The recommended practice is applicable to metal cladding of all types of fuel elements. The tests cited are suitable for determining mechanical properties of the fuel elements cladding. Various ASTM standards and test methods are cited

  10. High-temperature irradiation of niobium-1 w/o zirconium-clad UO/sub 2/. [Compatibility with lithium

    Energy Technology Data Exchange (ETDEWEB)

    Kangilaski, M.; Fromm, E.O.; Lozier, D.H.; Storhok, V.W.; Gates, J.E.

    1965-06-28

    Twenty-four 0.225-in.-diameter and six 0.290-in.-diameter UO/sub 2/ specimens clad with 80 mils of niobium-1 w/o zirconium were irradiated to burnups of 1.4 to 6.0 at. % of uranium at surface temperatures of 900 to 1400/sup 0/C. UO/sub 2/ and lithium were found to be incompatible at these temperatures, and the thick cladding was used primarily to minimize the chances of contact of UO/sub 2/ and the lithium coolant. The thickly clad specimens did not undergo any dimensional changes as a result of irradiation, although it was found that movement of UO/sub 2/ took place in the axial direction by a vaporization-redeposition mechanism. It was found that 32 to 87% of the fission gases was released from the fuel, depending on the temperature of the specimen. Metallographic examination of longitudinal and transverse sections of the specimens indicated the usual UO/sub 2/ microstructure with columnar grains. Grain-boundary thickening was observed in the UO/sub 2/ at higher burnups. The oxygen/uranium ratio of UO/sub 2/ increased with increasing burnup.

  11. 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

  12. Mechanical behaviour and failure of fuel cladding zirconium alloys in nuclear power plants under accidental RIA-type situation

    International Nuclear Information System (INIS)

    Doan, D.T.

    2009-01-01

    In French Nuclear Pressurized Water Reactors (PWRs), most of structural parts of the fuel assembly consist of zirconium alloy tubes and plates. Optimizing the management of fuel in nuclear power plants led to the increase in the duration of fuel cycles and power. The use of high fuel burnups requires drastic changes in the rules for reactor design in the nuclear safety. The evaluation of nuclear reactors in accident situations is based on reference accident scenarios. One of these hypothetical accidents, examined in this study, is the 'Reactivity Initiated Accident'. In order to assess the structural integrity of these parts it is necessary to characterize both the plastic flow and fracture behaviour of the materials at various stages of the life cycle, (i.e. at increasing levels of hydriding, irradiation, oxidation or thermal mechanical loading). The purpose of this work is to provide experimental data and to develop a model of the thermo-mechanical behaviour and to propose a design analysis method in the case of non-irradiated clads, in RIA-type situations. Mechanical tests were conducted on Cold-Worked-Stress-Relieved and on Recrystallized Zircaloy-4 sheets using various kinds of samples including smooth and notched tensile specimens and small punch tests. Temperature was set to 25, 250 and 600 C with hydrogen contents between 0 and 1000 ppm. The model is based on a simplified description of a Zircaloy polycrystal in which scalar isotropic ductile damage including void nucleation and growth is added. The model is also physically based to easily transfer parameters determined for one material state to another (e.g. transfer between sheet and tube or between different levels of irradiation). The model was implemented in the Finite Element software Zebulon using either an explicit or an implicit time integration scheme. Uniaxial tension tests were used to tune the model parameters for both materials, considering various values of temperature and hydrogen levels

  13. 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.

  14. Critical assessment of finite element analysis applied to metal–oxide interface roughness in oxidising zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Platt, P., E-mail: Philip.Platt@manchester.ac.uk [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Frankel, P. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Gass, M. [AMEC, Walton House, Faraday Street, Birchwood Park, Risley, Warrington WA3 6GA (United Kingdom); Preuss, M. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom)

    2015-09-15

    As a nuclear fuel cladding material, zirconium alloys act as a barrier between the fuel and pressurised steam or lithiated water environment. Controlling degradation mechanisms such as oxidation is essential to extending the in-service lifetime of the fuel. At temperatures of ∼360 °C zirconium alloys are known to exhibit cyclical, approximately cubic corrosion kinetics. With acceleration in the oxidation kinetics occurring every ∼2 μm of oxide growth, and being associated with the formation of a network of lateral cracks. Finite element analysis has been used previously to explain the lateral crack formation by the development of localised out-of-plane tensile stresses at the metal–oxide interface. This work uses the Abaqus finite element code to assess critically current approaches to representing the oxidation of zirconium alloys, with relation to undulations at the metal–oxide interface and localised stress generation. This includes comparison of axisymmetric and 3D quartered modelling approaches, and investigates the effect of interface geometry and plasticity in the metal substrate. Particular focus is placed on the application of the anisotropic strain tensor used to represent the oxidation mechanism, which is typically applied with a fixed coordinate system. Assessment of the impact of the tensor showed that 99% of the localised tensile stresses originated from the out-of-plane component of the strain tensor, rather than the in-plane expansion as was previously thought. Discussion is given to the difficulties associated with this modelling approach and the requirements for future simulations of the oxidation of zirconium alloys.

  15. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed

  16. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors

    International Nuclear Information System (INIS)

    Raepsaet, C.; Bossis, Ph.; Hamon, D.; Bechade, J.L.; Brachet, J.C.

    2007-01-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  17. 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)

  18. Spectral interference of zirconium on 24 analyte elements using CCD based ICP-AES technique

    International Nuclear Information System (INIS)

    Adya, V.C.; Sengupta, Arijit; Godbole, S.V.

    2014-01-01

    In the present studies, the spectral interference of zirconium on different analytical lines of 24 critical analytes using CCD based ICP-AES technique is described. Suitable analytical lines for zirconium were identified along with their detection limits. The sensitivity and the detection limits of analytical channels for different elements in presence of Zr matrix were calculated. Subsequently analytical lines with least interference from Zr and better detection limits were selected for their determinations. (author)

  19. Fabrication of the fuel elements cladding for utilization in the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Schaeffer, L.; Sefidvash, F.

    1986-01-01

    A method for the fabrication of cladding of the spherical fuel elements for the utilization in the fluidized bed nuclear reactor is presented. Some prelimminary experiments were performed to adopt a method which adapt itself to mass production with the desired high quality. Still methods for cladding fabrication are under study. (Author) [pt

  20. Pellet-clad interaction observations in boiling water reactor fuel elements

    International Nuclear Information System (INIS)

    Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R.

    1981-01-01

    Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. (author)

  1. 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.

  2. 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

  3. 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.)

  4. Critical stability conditions of the fuel element cladding; Kriticni uslovi stabilnosti kosuljice G.E

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, M; Savic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-12-15

    The role of the fuel element cladding being the first safety barrier, is to prevent contamination by the fission products. Construction of the fuel element cladding depends on the reactor type, coolant type, fuel type, technology of material fabrication, influence of the material on the neutron economy, thermal conditions, etc. That is why an optimum solution has to be found. This paper deals with mechanical properties of ceramic natural UO{sub 2} sintered fuel pellets in the zircaloy-2 cladding. This type of fuel is used in heavy water reactors.

  5. Technique for mass-spectrometric determination of moisture content in fuel elements and fuel element claddings

    International Nuclear Information System (INIS)

    Kurillovich, A.N.; Pimonov, Yu.I.; Biryukov, A.S.

    1988-01-01

    A technique for mass-spectroimetric determination of moisture content in fuel elements and fuek claddings in the 2x10 -4 -1.5x10 -2 g range is developed. The relative standard deviation is 0.13. A character of moisture extraction from oxide uranium fuels in the 20-700 deg C temperature range is studied. Approximately 80% of moisture is extracted from the fuels at 300 deg C. The moisture content in fuel elements with granular uranium oxide fuels is measured. Dependence of fuel element moisture content on conditions of hot vacuum drying is shown. The technique permits to optimize the fuel element fabrication process to decrease the moisture content in them. 4 refs.; 3 figs.; 2 tabs

  6. Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

    Energy Technology Data Exchange (ETDEWEB)

    Patra, Anirban [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-23

    This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

  7. 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

  8. Long-time corrosion and high-temperature oxidation of zirconium alloys applied on NPP like fuel elements cover

    International Nuclear Information System (INIS)

    Vrtilkova, V.; Novotny, L.; Lingart, S.; Doukha, R.; Yarosh, Ya.; Kolenchik, Ya.

    2007-01-01

    Zirconium is applying in nuclear energy since 50-th of last century in capacity of material for cover production for fuel elements, reactor fuel and structural parts, and mainly due to both corrosion stability and low effective cross section for thermal neutrons capture. Impurities in doping elements form and alloy production technology has influence on mechanical and corrosion properties of finite alloy. Long-time corrosion tests for several zirconium alloys in forcing autoclave under different reaction conditions were carried out. After that process kinetics was studied, mass increase, hydrogen formation, zirconium hydride forming morphology, zirconium oxide layer thickness have been determined as well

  9. Plastic deformation of the cladding of Fortissimo fuel elements

    International Nuclear Information System (INIS)

    Marbach, G.; Millet, P.; Blanchard, F.

    1979-07-01

    A study of a large number of standard Fortissimo pins, clad in solution treated 316 steel, shows that the plastic strain depends linearly on the fission gas pressure and the dose (in dpaF). The derived modulus of irradiation creep ranges from 1 to 2 x 10 -6 (MPa dpaF) -1 at 450 0 C and increases steadily with temperature. (author)

  10. Finite element simulation of a novel composite light-weight microporous cladding panel

    Science.gov (United States)

    Tian, Lida; Wang, Dongyan

    2018-04-01

    A novel composite light-weight microporous cladding panel with matched connection detailing is developed. Numerical simulation on the experiment is conducted by ABAQUS. The accuracy and rationality of the finite element model is verified by comparison between the simulation and the experiment results. It is also indicated that the novel composite cladding panel is of desirable bearing capacity, stiffness and deformability under out-of-plane load.

  11. Nuclear fuel element

    International Nuclear Information System (INIS)

    Iwano, Yoshihiko.

    1993-01-01

    Microfine cracks having a depth of less than 10% of a pipe thickness are disposed radially from a central axis each at an interval of less than 100 micron over the entire inner circumferential surface of a zirconium alloy fuel cladding tube. For manufacturing such a nuclear fuel element, the inside of the cladding tube is at first filled with an electrolyte solution of potassium chloride. Then, electrolysis is conducted using the cladding tube as an anode and the electrolyte solution as a cathode, and the inner surface of the cladding tube with a zirconium dioxide layer having a predetermined thickness. Subsequently, the cladding tube is laid on a smooth steel plate and lightly compressed by other smooth steel plate to form microfine cracks in the zirconium dioxide layer on the inner surface of the cladding tube. Such a compressing operation is continuously applied to the cladding tube while rotating the cladding tube. This can inhibit progress of cracks on the inner surface of the cladding tube, thereby enabling to prevent failure of the cladding tube even if a pellet/cladding tube mechanical interaction is applied. Accordingly, reliability of the nuclear fuel elements is improved. (I.N.)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  13. Stochastic thermal stress analysis of clad cylindrical fuel elements

    International Nuclear Information System (INIS)

    Barrett, P.R.

    1975-01-01

    After a review of deterministic elastic thermal stress analysis by means of the displacement method for a cylindrical system in which the temperature distribution is not only radially variable but azimuthally and axially variable also, a method is shown for the determination of the statistical moments of the stress components when (a) the outer boundary of the cladding is a stochastic quantity, and (b) the uncertainties in the elastic and thermal constants of the materials and in the magnitude of the heat generation term are taken into account. A typical model is proposed for describing the statistics of the outer radius of the cladding which is a stochastic variable owing to uncertainties produced by the extrusion process. The theory is illustrated by means of a simple example by examining a meaningful reliability index and the relative importance of each of the uncertainties. (Auth.)

  14. Heat treated tube for cladding nuclear fuel element

    International Nuclear Information System (INIS)

    Eddens, F.C.; White, D.W.; Harmon, J.L.

    1983-01-01

    The zirconium alloy tube comprises a metallurgical gradient across the width of the tube wall wherein the tube has a more corrosion-resistant metallurgical condition at the outer circumference and a less corrosion-resistant metallurgical condition at the inner circumference. The metallurgical gradient can be generated by heating an outer circumferential portion of the tube to the high alpha or mixed alpha plus beta range while maintaining the inner surface at a lower temperature, followed by cooling of the tube. Preferably the tube is made of Zircaloy. (author)

  15. 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

  16. 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

  17. Effects of Cooling Rates on Hydride Reorientation and Mechanical Properties of Zirconium Alloy Claddings under Interim Dry Storage Conditions

    International Nuclear Information System (INIS)

    Min, Su-Jeong; Kim, Myeong-Su; Won, Chu-chin; Kim, Kyu-Tae

    2013-01-01

    As-received Zr-Nb cladding tubes and 600 ppm hydrogen-charged tubes were employed to evaluate the effects of cladding cooling rates on the extent of hydride reorientation from circumferential hydrides to radial ones and mechanical property degradations with the use of cooling rates of 2, 4 and 15 °C/min from 400 °C to room temperature simulating cladding cooling under interim dry storage conditions. The as-received cladding tubes generated nearly the same ultimate tensile strengths and plastic elongations, regardless of the cooling rates, because of a negligible hydrogen content in the cladding. The 600 ppm-H cladding tubes indicate that the slower cooling rate generated the larger radial hydride fraction and the longer radial hydrides, which resulted in greater mechanical performance degradations. The cooling rate of 2 °C/min generates an ultimate tensile strength of 758 MPa and a plastic elongation of 1.0%, whereas the cooling rate of 15 °C/min generates an ultimate tensile strength of 825 MPa and a plastic elongation of 15.0%. These remarkable mechanical property degradations of the 600 ppm-H cladding tubes with the slowest cooling rate may be characterized by cleavage fracture surface appearance enhanced by longer radial hydrides and their higher fraction that have been precipitated through a relatively larger nucleation and growth rate.

  18. Solidification technique of radioactive elements. Research using zirconium phosphates

    International Nuclear Information System (INIS)

    Nakayama, Susumu; Ito, Katsuhiko

    2005-01-01

    Proton type zirconium phosphates HZr 2 (PO 4 ) 3 , NASICON type three-dimensional net work structure, is used for solidification of Cs in the high level radioactive waste. Two kinds of solidification methods such as the dry method and autoclave method are explained. Cs ion entered into 0.6nm space of HZr 2 (PO 4 ) 3 , and formed ionic bonding, which made the difficult situation to remove. When mixture of HZr 2 (PO 4 ) 3 and 23 kinds of M(NO 3 )n (M= Li, Na, K, Pb, Sr, Bi, Y, Mg, Ca, Sc, Mn, Fe, Co, Ni, Cu, Zn, Ag, Cd, Ba, La, Ce, Tl, and Pb; n=1,2 or 3) was treated at 400-700degC by dry method, solidification of the subject metals was succeeded. Amount of solidification of Cs by autoclave at 250degC is almost same as the dry method and its leachability resistance increased 40 times than that of dry method after heat treatment in atmosphere at 700degC. (S.Y.)

  19. MULTILAYER COMPOSITE PLASMA COATINGS ON SCREEN PROTECTION ELEMENTS BASED ON ZIRCONIUM DIOXIDE

    Directory of Open Access Journals (Sweden)

    V. A. Okovity

    2017-01-01

    Full Text Available The paper contains results of investigations pertaining to an influence of plasma jet parameters (current, spraying distance, consumption of plasma formation gas (nitrogen, fractional composition of initial powder and degree of cooling with compressed air on anti-meteoric coating characteristics. Optimum modes (arc current 600 A; spray distance of 110 mm; consumption of plasma formation gas (nitrogen – 50 l/min; fractional composition of zirconium dioxide powder <50 μm; compressed air consumption for cooling – 1 m3/min; p = 4 bar make it possible to obtain anti-meteoric coatings based on zirconium dioxide with material utilization rate of 62 %, total ceramic layer porosity of 6 %. After exposure of compression plasma flows on a coating in the nitrogen atmosphere a cubic modification of zirconium oxide is considered as the main phase being present in the coating. The lattice parameter of cubic zirconium oxide modification is equal to 0.5174 nm. Taking into consideration usage of nitrogen as plasma formation substance its interaction with zirconium coating atoms occurs and zirconium nitride (ZrN is formed with a cubic crystal lattice (lattice parameter 0.4580 nm. Melting of pre-surface layer takes place and a depth of the melted layer is about 8 μm according to the results of a scanning electron microscopy. Pre-surface layer being crystallized after exposure to compression plasma flows is characterized by a homogeneous distribution of ele-ments and absence of pores formed in the process of coating formation. The coating structure is represented by a set of lar- ge (5–7 μm and small (1–2 μm zirconium oxide particles sintered against each other. Melting of coating surface layer and speed crystallization occur after the impact of compression plasma flows on the formed coating. Cracking of the surface layer arises due to origination of internal mechanical stresses in the crystallized part. While using a scanning electron microscopy a

  20. Performance testing of refractory alloy-clad fuel elements for space reactors

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

    1985-01-01

    Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

  1. 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

  2. The role of a fuel element and its cladding in water cooled reactor dynamics

    International Nuclear Information System (INIS)

    Randles, J.

    1963-10-01

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO 2 fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

  3. The role of a fuel element and its cladding in water cooled reactor dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1963-10-15

    To clarify the role of fuel element cladding in water reactor dynamics, the heat diffusion and transfer equations are solved in slab geometry for (a) an oscillatory fission power, (b) an oscillatory coolant temperature. From the resulting transfer functions a clear description of the effect of the cladding on the heat flow is obtained. A Mercury autocode programme for evaluating the transfer functions is described. In addition to the slab element, the heat diffusion equations are also solved for a cylindrical system exposed to an oscillatory fission power. The solutions are expressed as transfer functions and are obtainable numerically from another autocode programme. Both of these programmes are used to obtain the power out/ power in transfer function for a typical cylindrical and slab UO{sub 2} fuel pellet clad in zircaloy. The results give a further indication of the effect of the cladding heat capacity over a wide frequency range. It is shown also that the effect of the geometrical difference between a slab and cylindrical fuel element is unimportant provided the surface to volume ratio of the fuel is the same in each case. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  5. Separation of zirconium and hafnium from acompanying elements by paper chromatography in the systems of alcohol-acid

    International Nuclear Information System (INIS)

    Lebedeva, G.G.; Viktorova, M.E.

    1980-01-01

    Solvents have been chosen and investigated which provide expressed separation of zirconium and hafnium at the analysis of mineral raw materials by paper chromatography. The systems with HNO 3 and HCl containing methyl, ethyl, propyl and butyl alcohols have been studied as mobile phases for separation of zirconium and hafnium. It has been shown that alcohol contents and solvent acidity affect the Rsub(f) value of these elements. The C 2 H 5 OH-5MHCl and C 2 H 5 OH-5MHNO 3 systems are most optimal for pre-concentration of zirconium and hafnium

  6. PCI resistant light water reactor fuel cladding

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  7. Radiochemical neutron activation analysis of zirconium and zirconium-niobium alloys

    International Nuclear Information System (INIS)

    Tashimova, F.A.; Sadikov, I.I.; Salimov, M.

    2004-01-01

    Full text: Zirconium and zirconium-niobium alloys are used on nuclear technology, as fuel cladding of nuclear reactors. Their nuclear-physical, mechanical and thermophysical properties are influenced them matrix and impurity composition, therefore determination of matrix and impurity content of these materials is a very important task. Neutron activation analysis is one from multielemental and high sensible techniques that are widely applied in analysis of high purity materials. Investigation of nuclear-physical characteristics of zirconium has shown that instrumental variant NAA is unusable for analysis due to high radioactivity of a matrix. Therefore it is necessary carrying out radiochemical separation of impurity radionuclides from matrix. Study of the literature datum have shown, that zirconium and niobium are very well extracted from muriatic solution with 5% tributyl phosphineoxide (TBPO) solution in toluene and 0,75 M solution of di-2-ethyl hexyl phosphoric acid (HDEHP) in cyclohexanone. Investigation of these elements extraction in these systems has shown that more effective and selective separation of matrix radionuclides is achieved in HDEHP-3M HCI system. This system is also extracted and hafnium, witch is an accompanying element of zirconium and its high content prevented determination of other impurity elements in sample. Therefore we used extraction system HDEHP-3M HCl for analysis of zirconium and zirconium-niobium alloys in chromatographic variant. By measurement of distribution profile of a matrix and of elution curve of determined elements is established, that for effective separation of impurity and matrix radionuclides there is enough chromatographic column with diameter 1 cm and height of a sorbent layer 7 cm, thus volume of elute, necessary for complete elution of determinate elements is 35-40 ml. On the basis of the carried out researches the technique of radiochemical NAA of high purity zirconium and zirconium-niobium alloy, which allows to

  8. Fuel element cladding state change mathematical model for a WWER-1000 plant operated in the mode of varying loading

    Directory of Open Access Journals (Sweden)

    S. N. Pelykh

    2010-09-01

    Full Text Available Main features of a fuel element cladding state change mathematical model for a WWER-1000 reactor plant operated in the mode of varying loading are listed. The integrated model is based on the energy creep theory, uses the finite element method for imultaneous solution of the fuel element heat conduction and mechanical deformation equa-tions. Proposed mathematical model allows us to determine the influence of the WWER-1000 regime parameters and fuel assembly design characteristics on the change of cladding properties under different loading conditions of normal operation, as well as the cladding limiting state at variable loading depending on the length, depth and number of cycles.

  9. Chemical aspects of pellet-cladding interaction in light water reactor fuel elements

    International Nuclear Information System (INIS)

    Olander, D.R.

    1982-01-01

    In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI

  10. 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

  11. Effects of alloying elements on nodular and uniform corrosion resistance of zirconium-based alloys

    International Nuclear Information System (INIS)

    Abe, Katsuhiro

    1992-01-01

    The effects of alloying and impurity elements (tin, iron, chromium, nickel, niobium, tantalum, oxygen, aluminum, carbon, nitrogen, silicon, and phosphorus) on the nodular and uniform corrosion resistance of zirconium-based alloys were studied. The improving effect of iron, nickel and niobium in nodular corrosion resistance were observed. The uniform corrosion resistance was also improved by nickel, niobium and tantalum. The effects of impurity elements, nitrogen, aluminum and phosphorus were negligibly small but increasing the silicon content seemed to improve slightly the uniform corrosion resistance. Hydrogen pick-up fraction were not changed by alloying and impurity elements except nickel. Nickel addition increased remarkably hydrogen pick-up fraction. Although the composition of secondary precipitates changed with contents of alloying elements, the correlation of composition of secondary precipitates to corrosion resistance was not observed. (author)

  12. Introduction of polycrystal constitutive laws in a finite element code with applications to zirconium forming

    International Nuclear Information System (INIS)

    Maudlin, P.J.; Tome, C.N.; Kaschner, G.C.; Gray, G.T. III

    1998-01-01

    In this work the authors simulate the compressive deformation of heavily textured zirconium sheet using a finite element code with the constitutive response given by a polycrystal self-consistent model. They show that the strong anisotropy of the response can be explained in terms of the texture and the relative activity of prismatic (easy) and pyramidal (hard) slip modes. The simulations capture the yield anisotropy observed for so-called through-thickness and in-plane compression tests in terMs of the loading curves and final specimen geometries

  13. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a metal liner disposed between the cladding and the nuclear fuel material and a high lubricity material in the form of a coating disposed between the liner and the cladding. The liner preferably has a thickness greater than the longest fission product recoil distance and is composed of a low neutron capture cross-section material. The liner is preferably composed of zirconium, an alloy of zirconium, niobium or an alloy of niobium. The liner serves as a preferential reaction site for volatile impurities and fission products and protects the cladding from contact and reaction with such impurities and fission products. The high lubricity material acts as an interface between the liner and the cladding and reduces localized stresses on the cladding due to fuel expansion and cracking of the fuel

  14. Analytical functions used for description of the plastic deformation process in Zirconium alloys WWER type fuel rod cladding under designed accident conditions

    International Nuclear Information System (INIS)

    Fedotov, A.

    2003-01-01

    The aim of this work was to improve the RAPTA-5 code as applied to the analysis of the thermomechanical behavior of the fuel rod cladding under designed accident conditions. The irreversible process thermodynamics methods were proposed to be used for the description of the plastic deformation process in zirconium alloys under accident conditions. Functions, which describe yielding stress dependence on plastic strain, strain rate and temperature may be successfully used in calculations. On the basis of the experiments made and the existent experimental data the dependence of yielding stress on plastic strain, strain rate, temperature and heating rate for E110 alloy was determined. In future the following research work shall be made: research of dynamic strain ageing in E635 alloy under different strain rates; research of strain rate influence on plastic strain in E635 alloy under test temperature higher than 873 K; research of deformation strengthening of E635 alloy under high temperatures; research of heating rate influence n phase transformation in E110 and E635 alloys

  15. Quantification of the distribution of hydrogen by nuclear microprobe at the Laboratory Pierre Sue in the width of zirconium alloy fuel clad of PWR reactors; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur de tubes de gainage du combustible des REP en alliage de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, C. [CEA Saclay, Dept. de Recherche sur l' Etat Condense, les Atomes et les Molecules (DSM/DRECAM/LPS-CNRS) UMR9956, 91 - Gif sur Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMULM2E), 91 - Gif-sur-Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMALA2M), 91 - Gif-sur-Yvette (France)

    2007-07-01

    Among the analysis techniques by ions beams, the micro ERDA (Elastic Detection Analysis) is an interesting technique which allows the quantitative distribution of the hydrogen in materials. In particular, this analysis has been used for hydride zirconium alloys, with the nuclear microprobe of the Laboratory Pierre Sue. This probe allows the characterization of radioactive materials. The technique principles are recalled and then two examples are provided to illustrate the fuel clad behavior in PWR reactors. (A.L.B.)

  16. Finite element analysis of the tetragonal to monoclinic phase transformation during oxidation of zirconium alloys

    Science.gov (United States)

    Platt, P.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

    2014-11-01

    Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal-oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations.

  17. Finite element analysis of the tetragonal to monoclinic phase transformation during oxidation of zirconium alloys

    International Nuclear Information System (INIS)

    Platt, P.; Frankel, P.; Gass, M.; Howells, R.; Preuss, M.

    2014-01-01

    Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal–oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations

  18. Finite element analysis of the tetragonal to monoclinic phase transformation during oxidation of zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Platt, P., E-mail: Philip.Platt@manchester.ac.uk [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Frankel, P. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom); Gass, M.; Howells, R. [AMEC, Walton House, Faraday Street, Birchwood Park, Risley, Warrington WA3 6GA (United Kingdom); Preuss, M. [University of Manchester, School of Materials, Materials Performance Centre, Manchester M13 9PL (United Kingdom)

    2014-11-15

    Corrosion is a key limiting factor in the degradation of zirconium alloys in light water reactors. Developing a mechanistic understanding of the corrosion process offers a route towards improving safety and efficiency as demand increases for higher burn-up of fuel. Oxides formed on zirconium alloys are composed of both monoclinic and meta-stable tetragonal phases, and are subject to a number of potential mechanical degradation mechanisms. The work presented investigates the link between the tetragonal to monoclinic oxide phase transformation and degradation of the protective character of the oxide layer. To achieve this, Abaqus finite element analysis of the oxide phase transformation has been carried out. Study of the change in transformation strain energy shows how relaxation of oxidation induced stress and fast fracture at the metal–oxide interface could destabilise the tetragonal phase. Central to this is the identification of the transformation variant most likely to form, and understanding why twinning of the transformed grain is likely to occur. Development of transformation strain tensors and analysis of the strain components allows some separation of dilatation and shear effects. Maximum principal stress is used as an indication of fracture in the surrounding oxide layer. Study of the stress distributions shows the way oxide fracture is likely to occur and the differing effects of dilatation and shape change. Comparison with literature provides qualitative validation of the finite element simulations.

  19. Consequences of metallic fuel-cladding liquid phase attack during over-temperature transient on fuel element lifetime

    International Nuclear Information System (INIS)

    Lahm, C.E.; Koenig, J.F.; Seidel, B.R.

    1990-01-01

    Metallic fuel elements irradiated in EBR-II at temperatures significantly higher than design, causing liquid phase attack of the cladding, were subsequently irradiated at normal operating temperatures to first breach. The fuel element lifetime was compared to that for elements not subjected to the over-temperature transient and found to be equivalent. 1 ref., 3 figs

  20. Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Bibilashvili, Yu K; Nekrasova, G A; Sukhanov, G I

    1989-01-01

    In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified.

  1. Influences of in-fuel physical-chemical processes on serviceability of energy reactor fuel elements

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Nekrasova, G.A.; Sukhanov, G.I.

    1989-01-01

    In-fuel physico-chemical processes and their effect on stress corrosion cracking of fuel element zirconium cladding are considered in the review. The mechanism of fission product release from the fuel is studied and the negative role of primarily iodine on the cladding corrosion process is demonstrated. Directions for improving the fuel element claddings and fuel to increase the fuel element serviceability are specified

  2. Preparation, microstructural evolution and properties of Ni–Zr intermetallic/Zr–Si ceramic reinforced composite coatings on zirconium alloy by laser cladding

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Kun; Li, Yajiang, E-mail: yajli@sdu.edu.cn; Wang, Juan; Ma, Qunshuang; Li, Jishuai; Li, Xinyue

    2015-10-25

    NiZr{sub 2}–ZrSi–Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}-ZrC intermetallic/ceramic reinforced composite coatings were in situ synthesized by laser cladding the pre-placed Ni–Cr–B–Si powder on zirconium substrate. Microstructure and phase constituents were investigated by X-ray diffraction (XRD), optical microscope (OM), scanning electron microscope (SEM) and energy dispersive spectroscopy (EDS). Microhardness tester and block-on-ring wear tester were employed to measure the hardness distribution and wear resistance of the intermetallic/ceramic reinforced composite coating. Results indicated that the multiphase of reinforcements includes Ni–Zr intermetallic compounds (e.g., NiZr and NiZr{sub 2}) and Zr–Si(C) ceramic phases (e.g., ZiSi, Zr{sub 5}Si{sub 4} and ZrC). Ni–Si clusters transforming to Zr–Si–Ni clusters at high temperature facilitated the forming of Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4} and during the growth of Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}, the consumption of Zr atoms at the lateral interface of liquid/Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4} resulted into developing Zr-poor zone near Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}. The microhardness and wear resistance of the coating were significantly improved by various reinforced phases in comparison to zirconium substrate. - Highlights: • NiZr{sub 2}–ZrSi–Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}-ZrC compostie coating was in-situ synthesized. • Ni–Si clusters transforming resulted into developing Zr-poor zone near Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4}. • Reinforced phases significantly improve wear resistance of the coating.

  3. Fuel-to-cladding heat transfer coefficient into reactor fuel element

    International Nuclear Information System (INIS)

    Lassmann, K.

    1979-01-01

    Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

  4. Influence of alloying elements on the dislocation loops created by Zr+ ion irradiation in alpha-zirconium

    International Nuclear Information System (INIS)

    Hellio, C.; Novion, C.H. de; Boulanger, L.

    1987-01-01

    Pure zirconium and four (annealed) α - zirconium based alloys (Zr-1760 ppm weight 0, Zr - 1% Nb - 430 ppm 0, Zr-1% Nb-1800 ppm 0, zircaloy 4) have been studied by transmission electron microscopy after 500 keV Zr + ion or 1 MeV electron irradiation performed at high temperature. Type of burgers vectors of the dislocation loops are given; in the case of electron irradiated Zr-1760 ppm 0, the larger loops were found of interstitial type. Alloying elements increase the loop density. The kinetic of loop growth was observed in-situ during 1 MeV electron irradiation between 400 and 700 0 C: oxygen was found to reduce considerably the growth speed of loops. In-situ annealing at 450 or 500 0 C after ion irradiation led to a large coalescence of loops in the case of pure zirconium, but modified only slightly the defect structure of the alloys

  5. 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.

  6. Analysis of hafnium in zirconium alloys

    International Nuclear Information System (INIS)

    Kondo, Isao; Sakai, Fumiaki; Ohuchi, Yoshifusa; Nakamura, Hisashi

    1977-01-01

    It is required to analyse alloying components and impurity elements in the acceptance analysis of zirconium alloys as the material for fuel cladding tubes and pressure tubes for advanced thermal reactors. Because of extreme similarity in chemical properties between zirconium and hafnium, about 100 ppm of hafnium is usually contained in zirconium alloys. Zircaloy-2 alloy and 2.5% Nb-zirconium with the addition of hafnium had been prepared as in-house standard samples for rapid analysis. Study was made on fluorescent X-ray analysis and emission spectral analysis to establish the analytical method. By using these in-house standard samples, acceptance analysis was successfully carried out for the fuel cladding tubes for advanced thermal reactors. Sulfuric acid solution was prepared from JAERI-Z 1, 2 and 3, the standard sample for zircaloy-2 prepared by the Analytical Committee on Nuclear Fuel and Reactor Materials, JAERI, and zirconium oxide (Hf 1 ppm/Zr). Standard Hf solution was added to the sulfuric acid solution step by step, to make up a series of the standard oxide samples by the precipitation process. By the use of these standard samples, the development of the analytical method and joint analysis were made by the three-member analytical technique research group including PNC. The analytical precision for the fluorescent X-ray analysis was improved by attaching a metallic yttrium filter to the window of an X-ray tube so as to suppress the effect due to zirconium matrix. The variation factor of the joint analysis was about 10% to show good agreement, and the indication value was determined. (Kobatake, H.)

  7. 3D finite element analysis of a nuclear fuel rod with gap elements between the pellet and the cladding

    International Nuclear Information System (INIS)

    Kang, Chang-Hak; Lee, Sung-Uk; Yang, Dong-Yol; Kim, Hyo-Chan; Yang, Yong-Sik

    2016-01-01

    Nuclear fuel rods which comprises an important component of a nuclear power plant are composed of nuclear fuel and cladding. Simulating the nuclear fuel rod using a computer program is the universal method to verify its safety. The computer program used for this is called the fuel performance code. The main objective of this study is to simulate the nuclear fuel rod behavior considering the gap conductance using three-dimensional gap elements. Gap elements are used because, unlike other methods, this approach does not require special methods or other variables such as the Lagrange multiplier. In this work, a nuclear fuel rod has been simulated and the results are compared with the experimental results. (author)

  8. Axisym finite element code: modifications for pellet-cladding mechanical interaction analysis

    International Nuclear Information System (INIS)

    Engelman, G.P.

    1978-10-01

    Local strain concentrations in nuclear fuel rods are known to be potential sites for failure initiation. Assessment of such strain concentrations requires a two-dimensional analysis of stress and strain in both the fuel and the cladding during pellet-cladding mechanical interaction. To provide such a capability in the FRAP (Fuel Rod Analysis Program) codes, the AXISYM code (a small finite element program developed at the Idaho National Engineering Laboratory) was modified to perform a detailed fuel rod deformation analysis. This report describes the modifications which were made to the AXISYM code to adapt it for fuel rod analysis and presents comparisons made between the two-dimensional AXISYM code and the FRACAS-II code. FRACAS-II is the one-dimensional (generalized plane strain) fuel rod mechanical deformation subcode used in the FRAP codes. Predictions of these two codes should be comparable away from the fuel pellet free ends if the state of deformation at the pellet midplane is near that of generalized plane strain. The excellent agreement obtained in these comparisons checks both the correctness of the AXISYM code modifications as well as the validity of the assumption of generalized plane strain upon which the FRACAS-II subcode is based

  9. Online study of cracks during laser cladding process based on acoustic emission technique and finite element analysis

    International Nuclear Information System (INIS)

    Wang Fujun; Mao Huaidong; Zhang Dawei; Zhao Xingyu; Shen Yu

    2008-01-01

    This paper presents a novel method by using acoustic emission technique to online study of the crack generation and expansion during laser cladding process. The cracks during laser cladding processes with five different laser cladding powder materials were online tested, respectively, with this method. Through finite element analysis (FEA), the temperature ranges of crack generation and expansion were figured out. The main forms and extended forms of the cracks were investigated by using optical microscope and scanning electron microscope (SEM). The experiment and analysis results show that the amount of cracks increases with the area and thickness of coating and the cooling rate increasing. A majority of cracks occur in the bonding zone and extend cross the whole cladding coating along the perpendicular direction of laser scanning during laser cladding, and few cracks generate in the cooling process. There are mainly four kinds of crack forms, and the forms of crack expansion are related to the stress states of coating and substrate. The method and conclusions in this paper provide important information for reducing the cracks during large area laser cladding process.

  10. A finite element elastic-plastic analysis of residual stresses due to clad welding in reactor vessels

    International Nuclear Information System (INIS)

    Buchalet, C.; Riccardella, P.C.

    1972-01-01

    Residual stresses due to weld deposited cladding on the inside of a typical Westinghouse pressurized water reactor vessel are investigated using an axisymmetric finite element elastic-plastic analysis. At the beginning of the analysis, one head of the weld cladding is assumed to lie on the reactor vessel wall at melting temperature (2600degF), but in the solid phase, while the vessel remains at 300degF (preheat temperature). All material properties used in the calculations are taken as temperature-dependent. Temperature profiles are obtained in the cladding and base metal at several discrete time intervals. These temperatures profiles are used to obtain the stress distribution for the same time intervals. Residual hoop tensile stresses of approximately 25 ksi were found to exist in the cladding. Peak tensile stresses in the hoop direction occur in the base metal near the cladding interface and reach a value of 60 ksi at the end of the transient. The tensile stress decreases very rapidly through the thickness of the base metal and becomes insignificant at about two inches from the inside surface. In order to lower residual stresses, a post-weld heat treatment is performed by uniformly heating the vessel to 1100degF, holding at that temperature for a specified period of time and then cooling slowly. The analysis shows that after this treatment, the peak stresses in the base metal decrease from 60 ksi to 32 ksi, while the stress in the cladding does not change significantly. (author)

  11. Nuclear fuel element and container

    International Nuclear Information System (INIS)

    Grubb, W.T.; King, L.H.

    1981-01-01

    The invention is based on the discovery that a substantial reduction in metal embrittlement or stress corrosion cracking from fuel pellet-cladding interaction can be achieved by the use of a copper layer or liner in proximity to the nuclear fuel, and an intermediate zirconium oxide barrier layer between the copper layer and the zirconium cladding substrate. The intermediate zirconia layer is a good copper diffusion barrier; also, if the zirconium cladding surface is modified prior to oxidation, copper can be deposited by electroless plating. A nuclear fuel element is described which comprises a central core of fuel material and an elongated container using the system outlined above. The method for making the container is again described. It comprises roughening or etching the surface of the zirconium or zirconium alloy container, oxidizing the resulting container, activating the oxidized surface to allow for the metallic coating of such surfaces by electroless deposition and further coating the activated-oxidized surface of the zirconium or zirconium alloy container with copper, iron or nickel or an alloy thereof. (U.K.)

  12. Study on fracture of fuel element cladding for naval reactor during typical accidents

    International Nuclear Information System (INIS)

    Zhang Fan; Shang Xueli; Zheng Zhongliang; Yu Lei

    2011-01-01

    Aiming at defining the grade of nuclear emergency response, the best estimate model has been adopted; the simulation of large break loss of coolant accident (LBLOCA) has been carried out by the radioactive analysis software coupled with relap5/mod 3.2 and core physics model. First, the peak clad temperature of the critical failure channel is calculated in relap5 code, and simultaneously its power factor is obtained. Second, pin power distribution of the fuel assemblies has been calculated in coarse-mesh nodal method. According to the pin power distribution in the whole core and the result gained above, the fraction of fuel element fracture is calculated. Finally, the radioactive analysis has been carried out and the reasonable source term is gotten, which can offer reference for the nuclear emergency decision making. (authors)

  13. The pellet-cladding contact in a fuel rod and its simulation by finite elements

    International Nuclear Information System (INIS)

    Tanajura, C.A.S.

    1988-01-01

    A model to analyse the mechanical behavior of a fuel rod of a PWR is presented. We drew our attention to the phenomenon of pellet-pellet and pellet-cladding contact by taking advantage of a model which assumes the hypotheses of axisymmetry, elastic behavior with infinitesimal deformations and changes of the material properties with temperature. It also includes the effects of swelling and initial relocation. The problem of contact gives rise to a variational formulation which employs Lagrangian multipliers. With this approach an iterative scheme is constructed to obtain the solution. The finite element method is applied to space discretization. The model sensibility to some parameters and its performance concerning fuel rod behavior is discussed by means of numerical simulations. (author) [pt

  14. Metallurgy of zirconium and hafnium

    International Nuclear Information System (INIS)

    Baryshnikov, N.V.; Geger, V.Eh.; Denisova, N.D.; Kazajn, A.A.; Kozhemyakin, V.A.; Nekhamkin, L.G.; Rodyakin, V.V.; Tsylov, Yu.A.

    1979-01-01

    Considered are those properties of zirconium and of hafnium, which are of practical interest for the manufacture of these elements. Systematized are the theoretical and the practical data on the procedures for thermal decomposition of zirconia and for obtaining zirconium dioxide and hafnium dioxide by a thermal decomposition of compounds and on the hydrometallurgical methods for extracting zirconium and hafnium. Zirconium and hafnium fluorides and chlorides production procedures are described. Considered are the iodide and the electrolytic methods of refining zirconium and hafnium

  15. PECITIS-II, a computer program to predict the performance of collapsible clad UO2 fuel elements

    International Nuclear Information System (INIS)

    Anand, A.K.; Anantharaman, K.; Sarda, V.

    1978-01-01

    The Indian power programme envisages the use of PHWRs, which use collapsible clad UO 2 fuel elements. A computer code, PECITIS-II, developed for the analysis of this type of fuel is described in detail. The sheath strain and fission gas pressure are evaluated by this method. The pellet clad gap conductance is calculated by Ross and Solute model. The pellet thermal expansion is calculated by assuming a two zone model, i.e. a plastic core surrounded by an elastic cracked annulus. (author)

  16. Thermodynamic Database for Zirconium Alloys

    International Nuclear Information System (INIS)

    Jerlerud Perez, Rosa

    2003-05-01

    For many decades zirconium alloys have been commonly used in the nuclear power industry as fuel cladding material. Besides their good corrosion resistance and acceptable mechanical properties the main reason of using these alloys is the low neutron absorption. Zirconium alloys are exposed to a very severe environment during the nuclear fission process and there is a demand for better design of this material. To meet this requirement a thermodynamic database is developed to support material designers. In this thesis some aspects about the development of a thermodynamic database for zirconium alloys are presented. A thermodynamic database represents an important facility in applying thermodynamic equilibrium calculations for a given material providing: 1) relevant information about the thermodynamic properties of the alloys e.g. enthalpies, activities, heat capacity, and 2) significant information for the manufacturing process e.g. heat treatment temperature. The basic information in the database is first the unary data, i.e. pure elements; those are taken from the compilation of the Scientific Group Thermodata Europe (SGTE) and then the binary and ternary systems. All phases present in those binary and ternary systems are described by means of the Gibbs energy dependence on composition and temperature. Many of those binary systems have been taken from published or unpublished works and others have been assessed in the present work. All the calculations have been made using Thermo C alc software and the representation of the Gibbs energy obtained by applying Calphad technique

  17. 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.)

  18. The Effect of Peak Temperatures and Hoop Stresses on Hydride Reorientations of Zirconium Alloy Cladding Tubes under Interim Dry Storage Condition

    International Nuclear Information System (INIS)

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

    2016-01-01

    In this study, the effect of peak temperatures and hoop tensile stresses on hydride reorientation in cladding was investigated. It was shown that the 250ppm-H specimens generated larger radial hydride fractions and longer radial hydrides than the 500ppm-H ones. The precipitated hydride in radial direction severely degrades mechanical properties of spent fuel rod. Hydride reorientation is related to cladding material, cladding temperature, hydrogen contents, thermal cycling, hoop stress and cooling rate. US NRC established the regulation on cladding temperature during the dry storage, which is the maximum fuel cladding temperature should not exceed 400 .deg. C for all fuel burnups under normal conditions of storage. However, if it is proved that the best estimate cladding hoop stress is equal to or less than 90MPa for the temperature limit proposed, a higher short-term temperature limit is allowed for low burnup fuel. In this study, 250ppm and 500ppm hydrogen-charged Zr-Nb alloy cladding tubes were selected to evaluate the effect of peak temperatures and hoop tensile stresses on the hydride reorientation during the dry storage. In order to evaluate threshold stresses in relation to various peak temperatures, four peak temperatures of 250, 300, 350, and 400 .deg. C and three tensile hoop stresses of 80, 100, 120MPa were selected.

  19. 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.)

  20. Fast-neutron interaction with elemental zirconium, and the dispersive optical model

    Energy Technology Data Exchange (ETDEWEB)

    Chiba, S.; Guenther, P.T.; Smith, A.B.; Sugimoto, M.; Lawson, R.D.

    1991-06-01

    Differential neutron elastic- and inelastic-scattering cross sections of elemental zirconium are measured from {approx} 1.5 to 10 MeV. Below 3 MeV the measurements are made at incident-neutron energy intervals of {approx} 100 keV, from 3 to 4 MeV at intervals of {approx} 200 keV, and at intervals of {approx} 500 keV at higher energies. the angular range of the measurements is {approx} 18{degrees} to 160{degrees}, with up to more than 100 differential values per distribution. this comprehensive data base, augmented with a 24-MeV elastic-scattering distribution from the literature, is used to develop two phenomenological optical-statistical models which both describe the data very well. First, the parameters of the conventional spherical optical model (SOM) are deduced. Secondly, the model in which the change in the real potential brought about by the dispersion relationship (DOM) is examined. the SOM parameters are consistent with systematics previously reported from this laboratory, and the volume integral-integral-per-nucleon of the real potential strength, J{sub v}, and the radius, r{sub v}, are energy dependent. When the DOM is used, a substantial part of the enrgy dependence of J{sub v}({approx} 30%) disappears. However, the change in the energy dependence of r{sub v} is small, so that a significant energy dependence remains when the DOM is used. Both models are extrapolated to the bound-state regime where they have modest success in predicting the binding energies of the single-particle and single-hole states in {sup 90}Zr. 55 refs., 20 figs., 7 tabs.

  1. Formation and Role of Gel Fractions in the Corrosion Layer of Zirconium Cladding as the First-stage Protection of the Nuclear Power Plant Fuel

    Czech Academy of Sciences Publication Activity Database

    Weishauptová, Zuzana; Vrtílková, V.; Bláhová, O.; Maixner, J.

    -, č. 16 (2007), s. 29-38 ISSN 1214-9691 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : zirconium alloys * corrosion layer * hydrated ZrO2 Subject RIV: CF - Physical ; Theoretical Chemistry

  2. 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.

  3. 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.

  4. Modelling of Zirconium and Hafnium separation using continuous annular chromatography

    International Nuclear Information System (INIS)

    Moch-Setyadji; Endang Susiantini

    2014-01-01

    Nuclear degrees of zirconium in the form of a metal alloy is the main material for fuel cladding of NPP. Zirconium is also used as sheathing UO 2 kernel in the form of ZrC as a substitute of SiC in the fuel elements of High Temperature Reactor (HTR). Difficulty separating hafnium from zirconium because it has a lot of similarities in the chemical properties of Zr and Hf. Annular chromatography is a device that can be used for separating of zirconium and hafnium to obtain zirconium nuclear grade. Therefore, it is necessary to construct the mathematical modelling that can describe the separation of zirconium and hafnium in the annular chromatography containing anion resin dowex-1X8. The aim of research is to perform separation simulation by using the equilibrium model and mass transfer coefficient resulted from research. Zr and Hf feed used in this research were 26 and 1 g/l, respectively. Height of resin (L), angular velocity (ω) and the superficial flow rate (uz) was varied to determine the effect of each parameter on the separation of Zr and Hf. By using Kd and Dv values resulted previous research. Simulation results showed that zirconium and hafnium can be separated using a continuous annular chromatography with high resin (long bed) 50 cm, superficial flow rate of 0.001 cm/s, the rotation speed of 0.006 rad/min and 20 cm diameter annular. In these conditions the results obtained zirconium concentration of 10,303.226 g/m 3 and hafnium concentration of 12.324 g/m 3 (ppm). (author)

  5. environmental studies for removal of some radioactive elements using zirconium silicate as inorganic ion exchange material

    International Nuclear Information System (INIS)

    El-Aryan, Y.F.A.

    2007-01-01

    inorganic ion exchangers have a good potential than the commonly used organic ones for removal and separation of radionuclides from irradiated nuclear fuel solutions. therefore, the main aim of this work is directed to find the optimum conditions for removal of some radionuclides such as Cs + ,Co 2+ ,and Eu 3+ by the prepared zirconium silicate as cation exchanger. the following items will be involved:-1- preparation of zirconium silicate as a cation exchanger. 2- characterization of the prepared exchanger using IR spectra, X-ray diffraction patterns, DTA and TG analyses. 3-chemical stability, capacity and equilibrium measurements will be determined on the materials using at different conditions (heating temperature and reaction temperature). 4- ion exchange isotherms. 5- breakthrough curves for removal of the investigated metal ions on the prepared exchanger under certain conditions

  6. Steady-state fission gas behavior in uranium-plutonium-zirconium metal fuel elements

    International Nuclear Information System (INIS)

    Steele, W.G.; Wazzan, A.R.; Okrent, D.

    1989-01-01

    An analysis of fission gas release and induced swelling in steady state irradiated U-Pu-Zr metal fuels is developed and computer coded. The code is used to simulate, with fair success, some gas release and induced swelling data obtained under the IFR program. It is determined that fuel microstructural changes resulting from zirconium migration, anisotropic swelling, and thermal variations are major factors affecting swelling and gas release behavior. (orig.)

  7. Nuclear fuel element, and method of producing same

    International Nuclear Information System (INIS)

    Armijo, J.S.; Esch, E.L.

    1986-01-01

    This invention relates to an improvement in nuclear fuel elements having a composite container comprising a cladding sheath provided with a protective barrier of zirconium metal covering the inner surface of the sheath, rendering such fuel elements more resistant to hydrogen accumulation in service. The invention specifically comprises removing substantially all zirconium metal of the barrier layer from the part of the sheath surrounding and defining the plenum region. Thus the protective barrier of zirconium metal covers only the inner surface of the fuel container in the area immediately embracing the fissionable fuel material

  8. Zirconium Micro-Arc Oxidation as a Method for Producing Heat Insulation Elements in Spacecraft

    Directory of Open Access Journals (Sweden)

    V. K. Shatalov

    2014-01-01

    Full Text Available Application of coatings on the surface of materials as well as their composition and structure control in the near-surface layer enables us to use properties of base material and modified layers in the most rational and profitable way and save expensive and rare metals and alloys.The space telescope of T-170M will be the main tool of the international space observatory "Spektr-UF".It is being understood that the main mirror shade, which is in the outer space and has a considerable height will act as a radiator cooling a unit (cage of the main mirror. Therefore it is necessary to create heat insulation between the shade of the main mirror and the frame of the main mirror unit. From the thermal calculations a detail to provide heat insulation must possess thermal conductivity, at most, 2,5 and a conditional limit of fluidity for compression, at least, 125 MPas to ensure that the shade diaphragms position of the main mirror is stable with respect to the optical system of telescope.Considering that oxide of zirconium possesses one of the lowest thermal conductivities among oxides of metals, it is offered to use zirconium, as a material of base, and to put the MAO-covering (micro-arc oxide on its surface.As a result of studying the features of MAO-coverings on zirconium it is:1 found that the composite material consisting of zirconium and MAO-covering on it, has low thermal conductivity (less than 2 , and thus, because of small oxide layer thickness against the thickness of base material, possesses the mechanical properties which are slightly different from the pure zirconium ones;2 found that the composite material possesses the low gas release, allowing its use in the outer space conditions; the material processed in two electrolytes i.e. phosphate and acid ones has the lowest gas release;3 found that with growing thickness of MAO-covering its porosity decreases, thus the average pore diameter grows thereby leading to increasing thermal

  9. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    1980-01-01

    The invention is of a nuclear fuel element which comprises a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium and mixtures thereof, and an elongated composite cladding container comprising a zirconium alloy tube containing constituents other than zirconium in an amount greater than about 5000 parts per million by weight and an undeformed metal barrier of moderate purity zirconium bonded to the inside surface of the alloy tube. The container encloses the core so as to leave a gap between the container and the core during use in a nuclear reactor. The metal barrier is of moderate purity zirconium with an impurity level on a weight basis of at least 1000ppm and less than 5000ppm. Impurity levels of specific elements are given. Variations of the invention are also specified. The composite cladding reduces chemical interaction, minimizes localized stress and strain corrosion and reduces the likelihood of a splitting failure in the zirconium alloy tube. Other benefits are claimed. (U.K.)

  10. UKAEA fast reactor project research and development programme on fuel element cladding and sub-assembly wrapper materials

    International Nuclear Information System (INIS)

    Harries, D.R.

    1977-01-01

    Research and development work on fuel element component (cladding, subassembly wrappers, etc.) materials for the U.K. sodium cooled fast reactor programme has been conducted at the United Kingdom Atomic Energy Authority (UKAEA) establishments at Dounreay, Harwell, Risley, and Springfields during the past fifteen years or so. This work has formed an integral part of, and has been co-ordinated by, the UKAEA Fast Reactor Project and has involved close liaison with the Nuclear Power Company (NPC) and the Central Electricity Generating Board (CEGB). The research and development were initially related to the Prototype Fast Reactor (PFR) but the scope has now been extended to cover the first Civil Fast Reactor (CFR1), which has recently been re-designated the Civil Demonstration Fast Reactor (CDFR). The paper outlines the present status of the development of sodium cooled fast reactors in the U.K. and proceeds to summarize the principal PFR and CDFR core and fuel element parameters which have determined the planning and direction of the fuel element materials programme. The current position on the fuel element cladding and wrapper research and development programme is reviewed, and the facilities and future irradiation programme to be carried out in PFR are described

  11. Elemental profiling of laser cladded multilayer coatings by laser induced breakdown spectroscopy and energy dispersive X-ray spectroscopy

    Science.gov (United States)

    Lednev, V. N.; Sdvizhenskii, P. A.; Filippov, M. N.; Grishin, M. Ya.; Filichkina, V. A.; Stavertiy, A. Ya.; Tretyakov, R. S.; Bunkin, A. F.; Pershin, S. M.

    2017-09-01

    Multilayer tungsten carbide wear resistant coatings were analyzed by laser induced breakdown spectroscopy (LIBS) and energy dispersive X-ray (EDX) spectroscopy. Coaxial laser cladding technique was utilized to produce tungsten carbide coating deposited on low alloy steel substrate with additional inconel 625 interlayer. EDX and LIBS techniques were used for elemental profiling of major components (Ni, W, C, Fe, etc.) in the coating. A good correlation between EDX and LIBS data was observed while LIBS provided additional information on light element distribution (carbon). A non-uniform distribution of tungsten carbide grains along coating depth was detected by both LIBS and EDX. In contrast, horizontal elemental profiling showed a uniform tungsten carbide particles distribution. Depth elemental profiling by layer-by-layer LIBS analysis was demonstrated to be an effective method for studying tungsten carbide grains distribution in wear resistant coating without any sample preparation.

  12. Temperature distribution determination of JPSR power reactor fuel element and cladding

    International Nuclear Information System (INIS)

    Sudarmono

    1996-01-01

    In order to utilize of fuel rod efficiency, a concept of JAERI passive Safety Reactor (JPSR) has been developed in Japan Atomic Energy Research Institute. In the JPSR design, UO 2 . are adopted as a fuel rod. The temperature distribution in the fuel rod and cladding in the hottest channel is a potential limiting design constraint of the JPSR. In the present determination, temperature distribution of the fuel rod and cladding for JPSR were PET:formed using COBRA-IV-I to evaluate the safety margin of the present JPSR design. In this method, the whole core was represented by the 1/4 sector and divided into 50 subchannels and 40 axial nodes. The temperature become maximum at the elevation of 1.922 and 2.196 m in the typical cell under operating condition. The maximum temperature in the center of the fuel rod surface of the fuel rod and cladding were 1620,4 o C, 722,8 o C, and 348,6 o C. The maximum results of temperature in the center of the fuel rod and cladding; were 2015,28 o C and 550 o C which were observed at 3.1 second in the typical cell

  13. Nuclear fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1977-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed which has a composite cladding having a substrate, a metal barrier metallurgically bonded to the inside surface of the substrate and an inner layer metallurgically bonded to the inside surface of the metal barrier. In this composite cladding, the inner layer and the metal barrier shield the substrate from any impurities or fission products from the nuclear fuel material held within the composite cladding. The metal barrier forms about 1 to about 4 percent of the thickness of the cladding and is comprised of a metal selected from the group consisting of niobium, aluminum, copper, nickel, stainless steel, and iron. The inner layer and then the metal barrier serve as reaction sites for volatile impurities and fission products and protect the substrate from contact and reaction with such impurities and fission products. The substrate and the inner layer of the composite cladding are selected from conventional cladding materials and preferably are a zirconium alloy. Also in a preferred embodiment the substrate and the inner layer are comprised of the same material, preferably a zirconium alloy. 19 claims, 2 figures

  14. 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.

  15. 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

  16. Zirconium - an imported mineral commodity

    International Nuclear Information System (INIS)

    1983-10-01

    This report examines Canada's position in regard to the principal zirconium materials: zircon; fusion-cast zirconium-bearing refractory products; zirconium-bearing chemicals; and zirconium metal, master alloys, and alloys. None of these is produced in Canada except fused alumina-zirconia and certain magnesium-zirconium alloys and zirconium-bearing steels. Most of the 3 000-4 000 tonnes of the various forms of zircon believed to be consumed in Canada each year is for foundry applications. Other minerals, notably chromite, olivine and silica sand are also used for these purposes and, if necessary, could be substituted for zircon. Zirconium's key role in Canada is in CANDU nuclear power reactors, where zirconium alloys are essential in the cladding for fuel bundles and in capital equipment such as pressure tubes, calandria tubes and reactivity control mechanisms. If zirconium alloys were to become unavailable, the Canadian nuclear power industry would collapse. As a contingency measure, Ontario Hydro maintains at least nine months' stocks of nuclear fuel bundles. Canada's vulnerability to short-term disruptions to supplies of nuclear fuel is diminished further by the availability of more expensive electricity from non-nuclear sources and, given time, from mothballed thermal plants. Zirconium minerals are present in many countries, notably Australia, the Republic of South Africa and the United States. Australia is Canada's principal source of zircon imports; South Africa is its sole source of baddeleyite. At this time, there are no shortages of either material. Canada has untapped zirconium resources in the Athabasca Oil Sands (zircon) and at Strange Lake along the ill-defined border between Quebec and Newfoundland (gittinsite). Adequate metal and alloy production facilities exist in France, Japan and the United States. No action by the federal government in regard to zirconium supplies is called for at this time

  17. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  18. Zircaloy oxidation and cladding deformation in PWR-specific CORA experiments

    International Nuclear Information System (INIS)

    Minato, K.; Hering, W.; Hagen, S.

    1991-07-01

    Out-of-pile bundle experiments (zircaloy 4) are performed in the CORA facility to investigate the behavior of PWR fuel elements during severe fuel damage (SFD) accidents. Within the international cooperation the most significant phenomena such as cladding deformation, oxidation (especially the zirconium/steam reaction), melt formation, melt release, and relocation which were found in all tests have been analyzed. (orig./MM) [de

  19. Zirconium - ab initio modelling of point defects diffusion

    International Nuclear Information System (INIS)

    Gasca, Petrica

    2010-01-01

    Zirconium is the main element of the cladding found in pressurized water reactors, under an alloy form. Under irradiation, the cladding elongate significantly, phenomena attributed to the vacancy dislocation loops growth in the basal planes of the hexagonal compact structure. The understanding of the atomic scale mechanisms originating this process motivated this work. Using the ab initio atomic modeling technique we studied the structure and mobility of point defects in Zirconium. This led us to find four interstitial point defects with formation energies in an interval of 0.11 eV. The migration paths study allowed the discovery of activation energies, used as entry parameters for a kinetic Monte Carlo code. This code was developed for calculating the diffusion coefficient of the interstitial point defect. Our results suggest a migration parallel to the basal plane twice as fast as one parallel to the c direction, with an activation energy of 0.08 eV, independent of the direction. The vacancy diffusion coefficient, estimated with a two-jump model, is also anisotropic, with a faster process in the basal planes than perpendicular to them. Hydrogen influence on the vacancy dislocation loops nucleation was also studied, due to recent experimental observations of cladding growth acceleration in the presence of this element [fr

  20. Effect of Electric Voltage and Current of X-ray Chamber on the Element inthe Zirconium Alloy Analysis X-ray by X-ray Fluorescence

    International Nuclear Information System (INIS)

    Yusuf-Nampira; Narko-Wibowo, L; Rosika-Krisnawati; Nudia-Barenzani

    2000-01-01

    The using of x-ray fluorescence in the chemical analysis depend heavilyon the parameters of x-ray chamber, for examples : electric voltage andelectric current. That parameter give effect in the result of determine ofSn, Cr, Fe and Ni in the zirconium alloy. 20 kV electric voltages are used onthe Mo x-ray chamber shall product x-ray of zirconium in the sample materialcan give effect in the stability of the analysis result (deviation more than5%). The result of analysis of elements in the zirconium alloy shall givedeviation less than 5% when using of electric voltage of the x-ray chamberless than 19 kV. The sensitivity of analysis can be reached by step upelectric current of x-ray chamber. (author)

  1. Stress concentration during pellet cladding interaction: Comparison of closed-form solutions with 2D(r,θ) finite element simulations

    International Nuclear Information System (INIS)

    Sercombe, Jérôme; Masson, Renaud; Helfer, Thomas

    2013-01-01

    Highlights: • This paper presents closed-formed solutions concerning pellet cladding interaction. • First, the opening of a radial crack in a pellet fragment is estimated. • Second, the stresses in the cladding in front of the pellet crack are calculated. • The closed-formed solutions are found in good agreement with 2D FE simulations. • They are then used in the fuel code ALCYONE to model PCI during power ramps. -- Abstract: This paper presents two closed-form solutions that can be used to enrich the mechanical description of fuel pellets and cladding behavior in standard one-dimensional based fuel performance codes. The first one is concerned with the estimation of the opening of a radial crack in a pellet fragment induced by the radial thermal gradient in the pellet and limited by the pellet-clad contact pressure. The second one describes the stress distribution in a cladding bore in front of an opening pellet crack. A linear angular variation of the pellet-clad contact pressure and a constant prescribed radial displacement are considered. The closed-form solutions are checked by comparison to independent finite element models of the pellet fragment and of the cladding. Their ability to describe non-axisymmetric displacement and stress fields during loading histories representative of base irradiation and power ramps is then demonstrated by cross-comparison with the 2D pellet fragment-cladding model of the multi-dimensional fuel performance code ALCYONE. The calculated radial crack opening profiles at different times and the hoop stress concentration in the cladding at the top of the ramp are found in good agreement with ALCYONE

  2. Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

    International Nuclear Information System (INIS)

    Scheglov, A.

    1994-01-01

    In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs

  3. Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability

    Energy Technology Data Exchange (ETDEWEB)

    Scheglov, A [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders (CCC) model as: axial asymmetry of fuel-cladding system (due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release), gaps between the pallets (and heat release peaking in fuel near the gap), chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets (characterized by a large central hole), temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs.

  4. Zirconium and cast zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Krone, K

    1977-04-01

    A survey is given on the occurence of zirconium, production of Zr sponge and semi-finished products, on physical and mechanical properties, production of Zr cast, composition of the commercial grades and reactor grades qualities, metal cutting, welding, corrosion behavior and use.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

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

    International Nuclear Information System (INIS)

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

    2001-02-01

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

  7. 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)

  8. Influence of alloying elements on the irradiation hardening and environmental sensitivity of zirconium alloys

    International Nuclear Information System (INIS)

    Pettersson, K.; Hallstadius, L.; Bergqvist, H.; Nylund, A.; Wikstroem, C.

    1992-01-01

    Ten different alloys of zirconium have been tested with regard to the effect of irradiation on their mechanical properties and their sensitivity to environmentally induced failure. Two different environments were used: iodine vapour and liquid cesium with an addition of 2% cadmium. The neutron dose was 10 21 n/cm 2 (E>1MeV) and the irradiation temperature was about 300 degrees C. All alloy additions increased the irradiation hardening. Especially notable was the large effect of titanium and tin on irradiation hardening. A limited amount of transmission electron microscopy was carried out in order to find an explanation to the effects. The testing in different environments showed that there is no clear correlation between environmental sensitivity and yield stress. For materials of similar yield stress an alloyed material tends to be more sensitive to environmental cracking than a material which only contains oxygen as an impurity. There also seems to be an effect of oxygen on the environmental cracking sensitivity. A material with 910 ppm oxygen was considerably more sensitive to cracking than a material with 470 ppm oxygen despite the fact that the yield stress values differed by only 90 MPa

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

    International Nuclear Information System (INIS)

    Silva, Roger F. da; Silva Júnior, Silvério F. da; Frade, Rangel T.; Rodrigues, Juliano S.

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  11. 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

  12. 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

  13. Electroless deposition process for zirconium and zirconium alloys

    Science.gov (United States)

    Donaghy, Robert E.; Sherman, Anna H.

    1981-01-01

    A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer.

  14. 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...

  15. 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...

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  17. Titanium(IV), zirconium, hafnium and thorium

    International Nuclear Information System (INIS)

    Brown, Paul L.; Ekberg, Christian

    2016-01-01

    Titanium can exist in solution in a number of oxidation states. The titanium(IV) exists in acidic solutions as the oxo-cation, TiO 2+ , rather than Ti 4+ . Zirconium is used in the ceramics industry and in nuclear industry as a cladding material in reactors where its reactivity towards hydrolysis reactions and precipitation of oxides may result in degradation of the cladding. In nature, hafnium is found together with zirconium and as a consequence of the contraction in ionic radii that occurs due to the 4f -electron shell, the ionic radius of hafnium is almost identical to that of zirconium. All isotopes of thorium are radioactive and, as a consequence of it being fertile, thorium is important in the nuclear fuel cycle. The polymeric hydrolysis species that have been reported for thorium are somewhat different to those identified for zirconium and hafnium, although thorium does form the Th 4 (OH) 8 8+ species.

  18. Investigation on fuel-cladding chemical interaction in metal fuel for FBR. Reaction of rare earth elements with Fe-Cr alloy

    International Nuclear Information System (INIS)

    Inagaki, Kenta; Ogata, Takanari

    2010-01-01

    Rare-earth fission product (FP) elements generated in the metal fuel interact with cladding alloy and result in the wastage of the cladding (Fuel-Cladding Chemical Interaction (FCCI)). To evaluate FCCI quantitatively, several influential factors must be considered. They are temperature, temperature gradient, time, composition of the cladding and the behavior of rare-earth FP. In this research, the temperature and time dependencies are investigated with tests in the simplified system. Fe-12wt%Cr was used as stimulant material of cladding and rare-earth alloy 13La -24Ce -12Pr -39Nd -12Sm (RE) as a rare-earth FP. A diffusion couple Fe-Cr/RE was made and annealed at 923K, 853K, 773K or 693K. The structures of reaction layers were analyzed with Electron Probe Micro Analyzer (EPMA) and the details of the structures were clarified. The width of the reaction layer in the Fe-Cr alloy grew in proportion to the square root of time. The reaction rate constants K=(square of the width of reaction layer / time) were evaluated. It was confirmed that the relation between K and the inverse of the temperature showed linearity above 773 K. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R.

    2015-01-01

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

  1. Degradation resistant fuel cladding materials and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Montes, J. [ENUSA, Madrid (Spain)

    1995-12-31

    GE has been producing the degradation resistant cladding (zirconium liner and zircaloy-2 surface larger) described here with the cooperation of its primary zirconium vendors since the beginning of 1994. Approximately 24 fuel reloads, or in excess of 250,000 fuel rods, have been produced using this material by GE. GE has also produced tubing for one reload of fuel that is currently being produced by its technology affiliate ENUSA. (orig./HP)

  2. Irradiation temperature memorization by retention of krypton-85. Application to the temperature determination for the internal cladding surface of fuel elements in PWR

    International Nuclear Information System (INIS)

    Fremiot, Claude

    1977-01-01

    The temperature of the inner surface of the cladding fuel elements, which can not be measured directly, can be determined after irradiation. During its stage within the reactor, the cladding is bombarded by krypton-85 fission product, which is trapped in the metallic lattice defects. The experience shows that the krypton release during postirradiation heating takes place at the irradiation temperature. This method was applied for PWR fuel element. A very simple model for retention and release of the krypton is proposed. The krypton trap-energy in zircaloy partakes in this model. This technique can be ordered amongst the Hot'Lab' control methods and expert appraisements. It is pointed out that the principal interest in that method is the fact that it does not need any fuel element instrumentation. At the present, this method is being used by CEA for routine-control. [fr

  3. 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.

  4. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

    1980-01-01

    Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

  5. Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

    International Nuclear Information System (INIS)

    Kulikov, I.S.; Tverkovkin, B.E.

    1984-01-01

    The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

  6. 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

  7. 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

  8. Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

    1992-01-01

    In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

  9. Localized deformation of zirconium-liner tube

    International Nuclear Information System (INIS)

    Nagase, Fumihisa; Uchida, Masaaki

    1988-03-01

    Zirconium-liner tube has come to be used in BWR. Zirconium liner mitigates the localized stress produced by the pellet-cladding interaction (PCI). In this study, simulating the ridging, stresses were applied to the inner surfaces of zirconium-liner tubes and Zircaloy-2 tubes, and, to investigate the mechanism and the extent of the effect, the behavior of zirconium liner was examined. As the result of examination, stress was concentrated especially at the edge of the deformed region, where zirconium liner was highly deformed. Even after high stress was applied, the deformation of Zircaloy part was small, since almost the concentrated stress was mitigated by the deformation of zirconium liner. In addition, stress and strain distributions in the cross section of specimen were calculated with a computer code FEMAXI-III. The results also showed that zirconium liner mitigated the localized stress in Zircaloy, although the affected zone was restricted to the region near the boundary between zirconium liner and Zircaloy. (author)

  10. Fuel Retention Improvement at High Temperatures in Tungsten-Uranium Dioxide Dispersion Fuel Elements by Plasma-Spray Cladding

    Science.gov (United States)

    Grisaffe, Salvatore J.; Caves, Robert M.

    1964-01-01

    An investigation was undertaken to determine the feasibility of depositing integrally bonded plasma-sprayed tungsten coatings onto 80-volume-percent tungsten - 20-volume-percent uranium dioxide composites. These composites were face clad with thin tungsten foil to inhibit uranium dioxide loss at elevated temperatures, but loss at the unclad edges was still significant. By preheating the composite substrates to approximately 3700 degrees F in a nitrogen environment, metallurgically bonded tungsten coatings could be obtained directly by plasma spraying. Furthermore, even though these coatings were thin and somewhat porous, they greatly inhibited the loss of uranium dioxide. For example, a specimen that was face clad but had no edge cladding lost 5.8 percent uranium dioxide after 2 hours at 4750 dgrees F in flowing hydrogen. A similar specimen with plasma-spray-coated edges, however, lost only 0.75 percent uranium dioxide under the same testing conditions.

  11. Zirconium and hafnium

    Science.gov (United States)

    Jones, James V.; Piatak, Nadine M.; Bedinger, George M.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Zirconium and hafnium are corrosion-resistant metals that are widely used in the chemical and nuclear industries. Most zirconium is consumed in the form of the main ore mineral zircon (ZrSiO4, or as zirconium oxide or other zirconium chemicals. Zirconium and hafnium are both refractory lithophile elements that have nearly identical charge, ionic radii, and ionic potentials. As a result, their geochemical behavior is generally similar. Both elements are classified as incompatible because they have physical and crystallochemical properties that exclude them from the crystal lattices of most rock-forming minerals. Zircon and another, less common, ore mineral, baddeleyite (ZrO2), form primarily as accessory minerals in igneous rocks. The presence and abundance of these ore minerals in igneous rocks are largely controlled by the element concentrations in the magma source and by the processes of melt generation and evolution. The world’s largest primary deposits of zirconium and hafnium are associated with alkaline igneous rocks, and, in one locality on the Kola Peninsula of Murmanskaya Oblast, Russia, baddeleyite is recovered as a byproduct of apatite and magnetite mining. Otherwise, there are few primary igneous deposits of zirconium- and hafnium-bearing minerals with economic value at present. The main ore deposits worldwide are heavy-mineral sands produced by the weathering and erosion of preexisting rocks and the concentration of zircon and other economically important heavy minerals, such as ilmenite and rutile (for titanium), chromite (for chromium), and monazite (for rare-earth elements) in sedimentary systems, particularly in coastal environments. In coastal deposits, heavy-mineral enrichment occurs where sediment is repeatedly reworked by wind, waves, currents, and tidal processes. The resulting heavy-mineral-sand deposits, called placers or paleoplacers, preferentially form at relatively low latitudes on passive continental margins and supply 100 percent of

  12. Production of nuclear grade zirconium: A review

    Energy Technology Data Exchange (ETDEWEB)

    Xu, L., E-mail: L.Xu-2@tudelft.nl [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110004 (China); Department of Materials Science and Engineering, Delft University of Technology, Delft 2628 CD (Netherlands); Xiao, Y. [Department of Metallurgical Engineering, Anhui University of Technology, Ma' anshan 243002 (China); Zr-Hf-Ti Metallurgie B.V., Den Haag 2582 SB (Netherlands); Sandwijk, A. van [Zr-Hf-Ti Metallurgie B.V., Den Haag 2582 SB (Netherlands); Xu, Q. [School of Materials Science and Metallurgy, Northeastern University, Shenyang 110004 (China); Yang, Y. [Department of Materials Science and Engineering, Delft University of Technology, Delft 2628 CD (Netherlands)

    2015-11-15

    Zirconium is an ideal material for nuclear reactors due to its low absorption cross-section for thermal neutrons, whereas the typically contained hafnium with strong neutron-absorption is very harmful for zirconium as a fuel cladding material. This paper provides an overview of the processes for nuclear grade zirconium production with emphasis on the methods of Zr–Hf separation. The separation processes are roughly classified into hydro- and pyrometallurgical routes. The known pyrometallurgical Zr–Hf separation methods are discussed based on the following reaction features: redox characteristics, volatility, electrochemical properties and molten salt–metal equilibrium. In the present paper, the available Zr–Hf separation technologies are compared. The advantages and disadvantages as well as future directions of research and development for nuclear grade zirconium production are discussed.

  13. Post-irradiation examination of the first SAP clad UO{sub 2} fuel elements irradiated in the X-7 organic loop

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, R. D.; Aspila, K.

    1962-02-15

    Seven fuel elements composing the first in-reactor test at Chalk River of SAP sheathing were irradiated in the X-7 organic loop. Activity, denoting a fuel failure, was detected in the loop coolant immediately after reactor start up; the fuel string was consequently removed from the loop nine hours later. Leak tests disclosed that five of the seven elements were defective. Inspection of the specimens showed essentially no change in element dimensions. Practically no organic fouling film was observed on the surface of the SAP cladding; organic coolant was found inside four of the defective elements. The appearance of the UO{sub 2} fuel was consistent with the irradiation time and the heat ratings achieved during the test. (author)

  14. Improvement of fuel-element reliability by insertion of UO2 microspheres in the gap between pellet and clad

    International Nuclear Information System (INIS)

    Mehedinteanu, S.; Glodeanu, F.; Dobos, I.

    1979-01-01

    With the accumulation of power reactor fuel operating experience, the study of the PCI phenomenon and the development of remedies have become important items in fuel research and development everywhere. The 'power-ramp' failure has drawn attention to the problem of obtaining high reliability from high burn-up fuel rods. Considerable attention has been paid to minimizing the cladding stresses imparted by fuel pellets during the power ramp. The paper describes a new concept of pellet-clad bonding by insertion of UO 2 microspheres in the gap. It is pointed out that the main advantages of this concept are: the low friction coefficient between pellet and clad; the accomodation of cracked pellet expansion by local microyielding of irradiation-embrittled clad; the reduced ridge height by use of undished pellets or other pellet shape; that the fine-sized UO 2 microspheres infiltrate around the pellets thus permitting the use of cracked or chipped pellets and also sintered pellets without the previously required grinding step needed for accurate sizing, etc. (author)

  15. First steps of integrated spatial modeling of titanium, zirconium, and rare earth element resources within the Coastal Plain sediments of the southeastern United States

    Science.gov (United States)

    Ellefsen, Karl J.; Van Gosen, Bradley S.; Fey, David L.; Budahn, James R.; Smith, Steven M.; Shah, Anjana K.

    2015-01-01

    The Coastal Plain of the southeastern United States has extensive, unconsolidated sedimentary deposits that are enriched in heavy minerals containing titanium, zirconium, and rare earth element resources. Areas favorable for exploration and development of these resources are being identified by geochemical data, which are supplemented with geological, geophysical, hydrological, and geographical data. The first steps of this analysis have been completed. The concentrations of lanthanum, yttrium, and titanium tend to decrease as distance from the Piedmont (which is the likely source of these resources) increases and are moderately correlated with airborne measurements of equivalent thorium concentration. The concentrations of lanthanum, yttrium, and titanium are relatively high in those watersheds that adjoin the Piedmont, south of the Cape Fear Arch. Although this relation suggests that the concentrations are related to the watersheds, it may be simply an independent regional trend. The concentration of zirconium is unrelated to the distance from the Piedmont, the equivalent thorium concentration, and the watershed. These findings establish a foundation for more sophisticated analyses using integrated spatial modeling.

  16. Advances in zirconium technology for nuclear reactor application

    International Nuclear Information System (INIS)

    Ganguly, C.

    2002-01-01

    Zirconium alloys are extensively used as a material for cladding nuclear fuels and for making core structurals of water-cooled nuclear power reactors all over the world for generation of nearly 16 percent of the worlds electricity. Only four countries in the world, namely France, USA, Russia and India, have large zirconium industry and capability to manufacture reactor grade zirconium sponge, a number of zirconium alloys and a wide variety of structural components for water cooled nuclear reactor. The present paper summarises the status of zirconium technology and highlights the achievement of Nuclear Fuel Complex during the last ten years in developing a wide variety of zirconium alloys and components for water-cooled nuclear power programme

  17. Annex 7 - Task 08/32 final report, Diffusion bond between the fuel and cladding of the RA reactor fuel element; Prilog 7 - Zavrsni izvestaj o podzadatku 08/32, Difuziona veza goriva i kosuljice gorivnog elementa reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Kustudic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Based on testing the diffusin bond between the fuel and the aluminium cladding of the RA reactor fuel element it was found that the diffusion bond exists; uranium is protected by a nickel layer to prevent the damaging diffusion between uranium and aluminium; the nickel layer is about 5 microns thick. Under assumption that the diffusion annealing was done after cladding the annealing temperature was about 500 deg C for about 1 hour.

  18. Experimental and thermodynamic study of the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems

    International Nuclear Information System (INIS)

    Jourdan, J.

    2009-11-01

    This work is a contribution to the development of innovative concepts for fuel cladding in pressurized water nuclear reactors. This concept implies the insertion of rare earth (erbium and gadolinium) in the zirconium fuel cladding. The determination of phase equilibria in the systems is essential prior to the implementation of such a promising solution. This study consisted in an experimental determination of the erbium-zirconium phase diagram. For this, we used many different techniques in order to obtain diagram data such as solubility limits, solidus, liquidus or invariant temperatures. These data allowed us to present a new diagram, very different from the previous one available in the literature. We also assessed the diagram using the CALPHAD approach. In the gadolinium-zirconium system, we determined experimentally the solubility limits. Those limits had never been determined before, and the values we obtained showed a very good agreement with the experimental and assessed versions of the diagram. Because these alloys are subjected to oxygen diffusion throughout their life, we focused our attention on the erbium-oxygen-zirconium and gadolinium-oxygen-zirconium systems. The first system has been investigated experimentally. The alloys fabrication has been performed using powder metallurgy. In order to obtain pure raw materials, we fabricated powder from erbium and zirconium bulk metals using hydrogen absorption/desorption. The characterisation of the ternary pellets allowed the determination of two ternary isothermal sections at 800 and 1100 C. For the gadolinium-oxygen-zirconium system, we calculated the phase equilibria at temperatures ranging from 800 to 1100 C, using a homemade database compiled from literature assessments of the oxygen-zirconium, gadolinium-zirconium and gadolinia-zirconia systems. Finally, we determined the mechanical properties, in connexion with the microstructure, of industrial quality alloys in order to identify the influence of

  19. ICP-AES determination of rare earths in zirconium with prior chemical separation of the matrix

    International Nuclear Information System (INIS)

    Rajeswari, B.; Dhawale, B.A.; Page, A.G.; Sastry, M.D.

    2002-01-01

    Zirconium being one of the most important material in nuclear industry used as a fuel cladding in reactors and an additive in advanced fuels necessitates its characterization for trace metallic contents. Zirconium, as refractory in nature as the rare earth elements, has a complex spectrum comprising of several emission lines. Rare earths, which are high neutron absorbers have to be analysed at very low limits. Hence, to achieve the desired limits, the major matrix has to be separated prior to rare earth determination. The present paper describes the method developed for the separation of rare earths from zirconium by solvent extraction using Trioctyl Phosphine Oxide (TOPO) as the extractant followed by their determination using Inductively Coupled Plasma - Atomic Emission Spectrometric (ICP-AES) method. Initially, radiochemical studies were carried out using known amounts of gamma active tracers of 141 Ce, 152-154 Eu, 153 Gd and 95 Zr for optimisation of extraction conditions using Tl- activated NaI detector. The optimum conditions at 0.5 M TOPO/xylene in 6 M HCl so as to achieve a quantitative recovery of rare earth analytes alongwith a near total extraction of zirconium in the organic phase, was further extended to carry out the studies using ICP-AES method. The recovery of rare earths was found to be quantitative within experimental error with a precision better than 10% RSD. (author)

  20. Method of separating hafnium from zirconium

    International Nuclear Information System (INIS)

    Megy, J.A.

    1980-01-01

    English. A new anhydrous method was developed for separating zirconium and hafnium, which gives higher separation factors and is more economical than previous methods. A molten phase, comprising a solution of unseparated zirconium and hafnium and a solvent metal, is first prepared. The molten metal phase is contacted with a fused salt phase which includes a zirconium salt. Zirconium and hafnium separation is effected by mutual displacement with hafnium being transported from the molten metal phase to the fused salt phase, while zirconium is transported from the fused salt phase to the molten metal phase. The solvent metal is less electropositive than zirconium. Zinc was chosen as the solvent metal, from a group which also included cadmium, lead, bismuth, copper, and tin. The fused salt phase cations are more electropositive than zirconium and were selected from a group comprising the alkali elements, the alkaline earth elements, the rare earth elements, and aluminum. A portion of the zirconium in the molten metal phase was oxidized by injecting an oxidizing agent, chlorine, to form zirconium tetrachlorid

  1. A standardized procedure for eddy-current testing of stainless steel, thin-walled nuclear fuel element cladding tubes

    International Nuclear Information System (INIS)

    Barat, P.; Raj, B.; Bhattacharya, D.K.

    1982-01-01

    Thin-walled nuclear fuel cladding tubes made of AISI 316 stainless steel have been examined by eddy-current testing. Standardization of the procedures has required investigations on optimizing the test frequency, finding a method to locate a defect with respect to the probe reference end, and the use of standard defects and sequential metallography of natural defects detected by eddy-current testing, to understand the influence of the nature of defects on the impedance output signals. Test frequency and method of locating the defect were optimized by the use of standard defects made by machining in reference cladding tubes. Subsequent metallography of natural defects detected by eddy-current testing revealed mainly clusters of inclusions but also other types of defects. The effect of the distribution of inclusions along the length of the tube on the impedance output is discussed. (author)

  2. Standardized procedure for eddy-current testing of stainless steel, thin-walled nuclear fuel element cladding tubes

    Energy Technology Data Exchange (ETDEWEB)

    Barat, P; Raj, B; Bhattacharya, D K [Reactor Research Centre, Kalpakkam (India)

    1982-10-01

    Thin-walled nuclear fuel cladding tubes made of AISI 316 stainless steel have been examined by eddy-current testing. Standardization of the procedures has required investigations on optimizing the test frequency, finding a method to locate a defect with respect to the probe reference end, and the use of standard defects and sequential metallography of natural defects detected by eddy-current testing, to understand the influence of the nature of defects on the impedance output signals. Test frequency and method of locating the defect were optimized by the use of standard defects made by machining in reference cladding tubes. Subsequent metallography of natural defects detected by eddy-current testing revealed mainly clusters of inclusions but also other types of defects. The effect of the distribution of inclusions along the length of the tube on the impedance output is discussed.

  3. Study of the uranium-zirconium diffusion; Etude de la diffusion uranium-zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Adda, Y; Mairy, C; Bouchet, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1957-07-01

    The intermetallic diffusion of uranium fuel and zirconium used as cladding is studied. Intermetallic diffusion can occur during the cladding of uranium rods and uranium can penetrate the zirconium cladding. Different parameters are involved in this mechanism as structure and mechanical properties of the diffusion area as well as presence of impurities in the metal. The uses of different analysis techniques (micrography, Castaing electronic microprobe, microhardness and autoradiography) have permitted to determine with great accuracy the diffusion coefficient in gamma phase (body centered cubic system) and the results have given important information on the intermetallic diffusion mechanisms. The existence of the Kirkendall effect in the U-Zr diffusion is also an argument in favor of the generality of the diffusion mechanism by vacancies in body centered cubic system. (M.P.)

  4. Study on dynamic measurement of fuel pellet length during loading into cladding tube

    International Nuclear Information System (INIS)

    Zhang Kai

    1993-09-01

    Various methods are presented for measuring the pellet length in the cladding tube (zirconium tube) during the loading process of the preparation of single rod of nuclear fuel assembly. These methods are used in former Soviet Union, west European countries and China in the manufacturing of nuclear power plant element. Different methods of dynamic measurement by using mechanics, optics and electricity and their special features are analysed and discussed. The structure and measuring principle of a developed measuring device,and its measuring precision and system deviation are also introduced. Finally, the length of loaded pellets is checked with analog pellets. The results are as expected and show that the method and principle used in the measuring device are feasible. It is an ideal and advanced method for the pellet loading of single cladding tube. The principle mentioned above can also be used in other industries

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. Some recent trends in the use of zirconium alloys for nuclear service

    International Nuclear Information System (INIS)

    Balaramamoorthy, K.

    1992-01-01

    Without any exception nuclear power reactors particularly the water cooled ones, operating in the World use natural or slightly enriched uranium oxide fuel pellets with zirconium alloy cladding. While the zirconium alloys have proven to be successful in their designed usage, a desire for longer lifetimes of core components and increased duty cycle puts more demand on materials performance. This demand has led to more in depth studies of phenomena associated with zirconium alloy corrosion mechanism, fine tuning of the zirconium alloy composition, development of fabrication techniques and to the evaluation of newer zirconium alloys for critical applications. (author). 5 refs., 32 figs

  11. Determination of zirconium by fluoride ion selective electrode

    International Nuclear Information System (INIS)

    Mahanty, B.N.; Sonar, V.R.; Gaikwad, R.; Raul, S.; Das, D.K.; Prakash, A.; Afzal, Md.; Panakkal, J.P.

    2010-01-01

    Full text: Zirconium is used in a wide range of applications including nuclear clad, catalytic converters, surgical appliances, metallurgical furnaces, superconductors, ceramics, lamp filaments, anti corrosive alloys and photographical purposes. Irradiation testing of U-Zr and U-Pu-Zr fuel pins has also demonstrated their feasibility as fuel in liquid metal reactors. Different methods that are employed for the determination of zirconium are spectrophotometry, potentiometry, neutron activation analysis and mass spectrometry. Ion-selective electrode (ISE), selective to zirconium ion has been studied for the direct potentiometric measurements of zirconium ions in various samples. In the present work, an indirect method has been employed for the determination of zirconium in zirconium nitrate sample using fluoride ion selective electrode. This method is based on the addition of known excess amount of fluoride ion to react with the zirconium ion to produce zirconium tetra fluoride at about pH 2-3, followed by determination of residual fluoride ion selective electrode. The residual fluoride ion concentrations were determined from the electrode potential data using calibration plot. Subsequently, zirconium ion concentrations were determined from the concentration of consumed fluoride ions. A precision of about 2% (RSD) with the mean recovery of more than 94% has been achieved for the determination of zirconium at the concentration of 4.40 X 10 -3 moles lit -1

  12. Preparation and characterization of zirconium phosphate ion exchanger samples with respect to the separation of highly active actinoid elements

    International Nuclear Information System (INIS)

    Treplan, J.

    1972-01-01

    Inorganic ion exchangers are of growing interest in connection with separation processes of α-radiators of high specific activity, or with high gamma doses, because they have a considerably higher radiation resistance at their disposal compared to the commonly used organic ion exchangers. In opposition to their use, however, are the worse properties regarding capacity, chemical resistivity, exchange rate and reproducibility of the ion exchange bed. In the present work, an attempt has been made to influence the properties of a typical representative of this group, zirconium phosphate (ZP), by systematic changing of the preparation parameters in such a manner that a sufficient capacity is obtained regarding tri-valent ions. In addition, information is to be gathered in order to clarify the connection between exchanger property and structure of the ZP. (orig./LH) [de

  13. Unirradiated UO2 in irradiated zirconium alloy sheathing

    International Nuclear Information System (INIS)

    MacDonald, R.D.; Hardy, D.G.; Hunt, C.E.L.; Scoberg, J.A.

    1979-07-01

    Zircaloy-clad UO 2 fuel elements have defected in power reactors when element power outputs were raised significantly after a long irradiation at low power. We have irradiated fuel elements fabricated from fresh UO 2 pellets and zirconium alloy sheaths previously irradiated without fuel. This gave a fuel element with radiation-damaged low-ductility sheathing but with no fission products in the fuel. The elements were power boosted in-reactor to linear power outputs up to 84 kW/m for two five-day periods. No elements defected despite sheath strains of 0.82 percent at circumferential ridge postions. Half of these elements were subsequently soaked at low power to build up the fission product inventory in the fuel and then power boosted to 63 kW/m for a third time. Two elements defected on this final boost. We conclude that these defects were caused by fission product induced stress-corrosion cracking and that this mechanism plays an importent role in power reactor fuel defects. (auth)

  14. Identification and characterization of a new Zirconium hydride

    International Nuclear Information System (INIS)

    Zhao, Z.

    2007-01-01

    In order to control the integrity of the fuel clad, alloy of zirconium, it is necessary to predict the behavior of zirconium hydrides in the environment (temperature, stress...), at a microscopic scale. A characterization study by TEM of hydrides has been realized. It shows little hydrides about 500 nm, in hydride Zircaloy 4. Then a more detailed study identified a new hydride phase presented in this paper. (A.L.B.)

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

    OpenAIRE

    Bo Cheng; Young-Jin Kim; Peter Chou

    2016-01-01

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

  16. Amine extraction of lead(II) and zirconium(IV) with succinate media

    International Nuclear Information System (INIS)

    Mahamuni, S.V.; Mane, C.P.; Sargar, B.M.; Rajmane, M.M.; Anuse, M.A.

    2004-01-01

    Lead is an important constituent of various alloys, which are in increasing demand in manufacture of batteries and nuclear shielding while the use of zirconium in nuclear power plants as entirely cladding uranium fuel is most important. This study was carried out to optimize the extraction conditions for Pb(II) and zirconium(IV)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

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

    International Nuclear Information System (INIS)

    Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie; Abe, Hiroaki

    2017-01-01

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

  19. Beryllium and zirconium

    International Nuclear Information System (INIS)

    Salesse, Marc

    1959-01-01

    Pure beryllium and zirconium, both isolated at about the same date but more than a century ago remained practically unused for eighty years. Fifteen years ago they were released from this state of inactivity by atomic energy, which made them into current metal a with an annual production which runs into tens of tons for the one and thousands for the other. The reasons for this promotion promise well for the future of the two metals, which moreover will probably find additional uses in other branches of industry. The attraction of beryllium and zirconium for atomic energy is easily explained. The curve of figure 1 gives the price per gram of uranium-235 as a function of enrichment: this price increases by about a factor of 3 on passing from natural uranium (0, 7 percent 235 U) to almost pure uranium-235. Because of their tow capture cross-section beryllium and zirconium make it possible, or at least easier, to use natural uranium and they thus enjoy an advantage the extent of which must be calculated for each reactor or fuel element project, but which is generally considerable. It will be seen later that this advantage should be based on figures which are even more favourable that would appear from the simple ratio 3 of the price of pure uranium- 235 contained in natural uranium. Reprint of a paper published in 'Industries Atomiques' - n. 1-2, 1959

  20. Oxidation of zirconium alloys in steam: influence of tetragonal zirconia on oxide growth mechanism

    International Nuclear Information System (INIS)

    Godlewski, J.

    1990-07-01

    The oxidation of zirconium alloys in presence of steam, presents after a 'parabolic' growth law, an acceleration of the oxidation velocity. This phenomenon limits the use of zirconium alloys as nuclear fuel cladding element. In order to determine the physico-chemical process leading to this kinetic transition, two approaches have been carried out: the first one has consisted to determine the composition of the oxide layer and its evolution with the oxidation time; and the second one to determine the oxygen diffusion coefficients in the oxide layers of pre- and post-transition as well as their evolution with the oxidation time. The composition of the oxide layers has been determined by two analyses techniques: the X-ray diffraction and the laser Raman spectroscopy. This last method has allowed to confirm the presence of tetragonal zirconium oxide in the oxide layers. Analyses carried out by laser Raman spectroscopy on oxides oblique cuttings have revealed that the tetragonal zirconium oxide is transformed in monoclinic phase during the kinetic transition. A quantitative approach has allowed to corroborate the results obtained by these two techniques. In order to determine the oxygen diffusion coefficients in the oxides layers, two diffusion treatments have been carried out: 1)under low pressure with D 2 18 O 2 ) under high pressure in an autoclave with H 2 18 O. The oxygen 18 concentration profiles have been obtained by two analyses techniques: the nuclear microprobe and the secondary ions emission spectroscopy. The obtained profiles show that the mass transport is made by the volume and particularly by the grain boundaries. The corresponding diffusion coefficients have been calculated with the WHIPPLE and LE CLAIRE solution. The presence of tetragonal zirconium oxide, its relation with the kinetic transition, and the evolution of the diffusion coefficients with the oxidation time, are discussed in terms of internal stresses in the oxide layer and of the oxide layer

  1. Anisotropy of mechanical properties of zirconium and zirconium alloys

    International Nuclear Information System (INIS)

    Medrano, R.E.

    1975-01-01

    In studies of technological applications of zirconium to fuel elements of nuclear reactor, it was found that the use of plasticity equations for isotropic materials is not in agreement with experimental results, because of the strong anisotropy of zirconium. The present review describes recent progress on the knowledge of the influence of anisotropy on mechanical properties, after Douglass' review in 1971. The review was written to be selfconsistent, changing drastically the presentation of some of the referenced papers. It is also suggested some particular experiments to improve developments in this area

  2. Zirconium determination in refractories (gravimetric method)

    International Nuclear Information System (INIS)

    Capiotto, N.; Narahashi, Y.; Perish, P.G.; Souza, J.R. de

    1991-01-01

    A gravimetric method for zirconium determination in refractories is described. X-ray fluorescence analysis is also employed in this experiment and considerations about interfering elements are presented. (M.V.M.)

  3. ;Study of secondary hydriding at high temperature in zirconium based nuclear fuel cladding tubes by coupling information from neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and laser induced breakdown spectroscopy microprobe

    Science.gov (United States)

    Brachet, Jean-Christophe; Hamon, Didier; Le Saux, Matthieu; Vandenberghe, Valérie; Toffolon-Masclet, Caroline; Rouesne, Elodie; Urvoy, Stéphane; Béchade, Jean-Luc; Raepsaet, Caroline; Lacour, Jean-Luc; Bayon, Guy; Ott, Frédéric

    2017-05-01

    This paper gives an overview of a multi-scale experimental study of the secondary hydriding phenomena that can occur in nuclear fuel cladding materials exposed to steam at high temperature (HT) after having burst (loss-of-coolant accident conditions). By coupling information from several facilities, including neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and micro laser induced breakdown spectroscopy, it was possible to map quantitatively, at different scales, the distribution of oxygen and hydrogen within M5™ clad segments having experienced ballooning and burst at HT followed by steam oxidation at 1100 and 1200 °C and final direct water quenching down to room temperature. The results were very reproducible and it was confirmed that internal oxidation and secondary hydriding at HT of a cladding after burst can lead to strong axial and azimuthal gradients of hydrogen and oxygen concentrations, reaching 3000-4000 wt ppm and 1.0-1.2 wt% respectively within the β phase layer for the investigated conditions. Consistent with thermodynamic and kinetics considerations, oxygen diffusion into the prior-β layer was enhanced in the regions highly enriched in hydrogen, where the α(O) phase layer is thinner and the prior-β layer thicker. Finally the induced post-quenching hardening of the prior-β layer was mainly related to the local oxygen enrichment. Hardening directly induced by hydrogen was much less significant.

  4. Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

    International Nuclear Information System (INIS)

    Zmitko, M.

    1999-01-01

    In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

  5. In reactor performance of defected zircaloy-clad U{sub 3}Si fuel elements in pressurized and boiling water coolants

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Allison, G M; Ambler, J F.R.; Chalder, G H; Lipsett, J J

    1968-05-15

    The results of two in-reactor defect tests of Zircaloy-clad U{sub 3}Si are reported. In the first test, a previously irradiated element ({approx}5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at {approx}270{sup o}C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U{sub 3}Si at 300{sup o}C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  6. Influence of protecting gel film on oxidation of zirconium alloys

    Czech Academy of Sciences Publication Activity Database

    Frank, H.; Weishauptová, Zuzana; Vrtílková, V.

    2007-01-01

    Roč. 360, č. 3 (2007), s. 282-292 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : fuel cladding * corrosion * Zirconium oxide Subject RIV: JF - Nuclear Energetics Impact factor: 1.643, year: 2007

  7. 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)

  8. FUNDAMENTAL MECHANISMS OF CORROSION OF ADVANCED LIGHT WATER REACTOR FUEL CLADDING ALLOYS AT HIGH BURNUP

    International Nuclear Information System (INIS)

    Lott, Randy G.

    2003-01-01

    OAK (B204) The corrosion behavior of nuclear fuel cladding is a key factor limiting the performance of nuclear fuel elements, improved cladding alloys, which resist corrosion and radiation damage, will facilitate higher burnup core designs. The objective of this project is to understand the mechanisms by which alloy composition, heat treatment and microstructure affect corrosion rate. This knowledge can be used to predict the behavior of existing alloys outside the current experience base (for example, at high burn-up) and predict the effects of changes in operation conditions on zirconium alloy behavior. Zirconium alloys corrode by the formation f a highly adherent protective oxide layer. The working hypothesis of this project is that alloy composition, microstructure and heat treatment affect corrosion rates through their effect on the protective oxide structure and ion transport properties. The experimental task in this project is to identify these differences and understand how they affect corrosion behavior. To do this, several microstructural examination techniques including transmission electron microscope (TEM), electrochemical impedance spectroscopy (EIS) and a selection of fluorescence and diffraction techniques using synchrotron radiation at the Advanced Photon Source (APS) were employed

  9. Quantification of hydrogen distribution with the nuclear microprobe of the Pierre Sue Laboratory in the thickness of the PWR fuel cladding in zirconium alloy; Quantification de la repartition de l'hydrogene a la microsonde nucleaire du Laboratoire Pierre Sue dans l'epaisseur du tube de gainage du combustible des REP en alliage de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Raepsaet, C. [Laboratoire Pierre Sue (DSM/DRECAM/LPS) - CEA Saclay, 91 - Gif-sur-Yvette (France); Bossis, Ph. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SEMI/LM2E), 91 - Gif sur Yvette (France); Hamon, D.; Bechade, J.L.; Brachet, J.C. [CEA Saclay, Dept. des Materiaux pour le Nucleaire (DEN/DANS/DMN/SRMA/LA2M), 91 - Gif sur Yvette (France)

    2007-07-01

    In a first part of this study, are detailed the general principles of the specific technique ERDA (Elastic Recoil Detection Analysis) used in the Pierre Sue Laboratory. Then, the contribution of this technique is illustrated with two studies examples on the behaviour of PWR nuclear fuel cladding. (O.M.)

  10. 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.

  11. “Study of secondary hydriding at high temperature in zirconium based nuclear fuel cladding tubes by coupling information from neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and laser induced breakdown spectroscopy microprobe

    Energy Technology Data Exchange (ETDEWEB)

    Brachet, Jean-Christophe, E-mail: jean-christophe.brachet@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Hamon, Didier; Le Saux, Matthieu [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Vandenberghe, Valérie [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); DEN-Service d’Etudes Mécaniques et Thermiques (SEMT), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Toffolon-Masclet, Caroline; Rouesne, Elodie; Urvoy, Stéphane [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Béchade, Jean-Luc [DEN-Service de Recherches Métallurgiques Appliquées (SRMA), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); DEN-Service de Recherches de Métallurgie Physique (SRMP), CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Raepsaet, Caroline [LEEL, CEA, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); NIMBE, CEA, CNRS, Université Paris-Saclay, 91191 Gif-sur-Yvette Cedex (France); and others

    2017-05-15

    This paper gives an overview of a multi-scale experimental study of the secondary hydriding phenomena that can occur in nuclear fuel cladding materials exposed to steam at high temperature (HT) after having burst (loss-of-coolant accident conditions). By coupling information from several facilities, including neutron radiography/tomography, electron probe micro analysis, micro elastic recoil detection analysis and micro laser induced breakdown spectroscopy, it was possible to map quantitatively, at different scales, the distribution of oxygen and hydrogen within M5™ clad segments having experienced ballooning and burst at HT followed by steam oxidation at 1100 and 1200 °C and final direct water quenching down to room temperature. The results were very reproducible and it was confirmed that internal oxidation and secondary hydriding at HT of a cladding after burst can lead to strong axial and azimuthal gradients of hydrogen and oxygen concentrations, reaching 3000–4000 wt ppm and 1.0–1.2 wt% respectively within the β phase layer for the investigated conditions. Consistent with thermodynamic and kinetics considerations, oxygen diffusion into the prior-β layer was enhanced in the regions highly enriched in hydrogen, where the α(O) phase layer is thinner and the prior-β layer thicker. Finally the induced post-quenching hardening of the prior-β layer was mainly related to the local oxygen enrichment. Hardening directly induced by hydrogen was much less significant. - Highlights: •More than 50% of the gaseous hydrogen produced by the inner clad oxidation absorbed and trapped into prior-β layer. •High hydrogen and oxygen local concentrations, up to 3000–4000 wt. ppm and 1.0–1.2 wt.% respectively, within the β phase. •Enhanced oxygen diffusion into hydrogen enriched prior-β layer, with locally thinner α(O) and thicker prior-β layers. •Post-quenching hardening of the prior-β structure mainly related to the (local) oxygen concentration.

  12. 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.

  13. Deformation mechanisms and irradiation effects in zirconium alloys. A multi-scale study

    International Nuclear Information System (INIS)

    Onimus, Fabien

    2015-01-01

    Zirconium alloys have been used for more than 30 years in the nuclear industry as structural materials for the fuel assemblies of pressurized water reactors. In particular, the cladding tube, made of zirconium alloys, constitutes the first barrier against the dissemination of radioactive elements. It is therefore essential to have a good understanding and prediction of the mechanical behavior of these materials in various conditions. The work presented in this dissertation deals with an experimental study and numerical simulations, at several length scales, of the deformation mechanisms and the mechanical behavior of zirconium alloys before irradiation, but also after irradiation and under irradiation. The mechanical behavior of zirconium single crystal has been determined, during an original study, using tensile test specimens containing large grains. Based on this study, crystal plasticity constitutive laws have been proposed. A polycrystalline model has also been developed to simulate the behavior of unirradiated zirconium alloys. A thorough Transmission Electron Microscopy (TEM) study has been able to clarify the deformation mechanisms of zirconium alloys occurring after irradiation. The clearing of loops by gliding dislocations leading to the dislocation channeling mechanism has been studied in details. This phenomenon has also been simulated using a dislocation dynamics code. The macroscopic consequences of this process have also been analyzed. A polycrystalline model taking into account the specificity of this mechanism has eventually been proposed. This approach has then been extended to the post-irradiation creep behavior. The recovery of radiation defects during creep tests has been characterized by TEM and modeled using cluster dynamics method. Deformation modes during creep have also been studied and a simple model for the creep behavior has eventually been proposed. Finally, the mechanism responsible for the acceleration of irradiation growth that

  14. 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

  15. 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.

  16. Zirconium ignition in exposed fuel channel

    Energy Technology Data Exchange (ETDEWEB)

    Elias, E., E-mail: merezra@technion.ac.il; Hasan, D.; Nekhamkin, Y.

    2015-05-15

    Highlights: • We demonstrate the idea of runaway zirconium–steam reactions in severe accidents in today's LWRs. • We predict the thermal-hydraulics conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core. • The Semenov theory of metal combustion is extended to define a criterion for runaway oxidation reaction in fuel cladding. - Abstract: A theoretical model based on simultaneous solution of the heat and mass transfer equations is developed for predicting the rate of thermo-chemical reaction between zirconium cladding and a hot steam environment. Ignition conditions relevant to cladding oxidation in an exposed fuel channel of a partially uncovered core are predicted based on the theory of metal combustion. A range of decay power, convective heat transfer coefficients, and initial temperatures leading to uncontrolled runaway cladding oxidation is identified. The model could be readily integrated as part of a fuel channel analysis code for predicting possible outcomes of different accident mitigation procedures in light water nuclear reactors under LOCA conditions.

  17. Super ODS steel R and D for fuel cladding of next generation nuclear systems. 2) Effect of minor alloying elements

    International Nuclear Information System (INIS)

    Ohnuki, S.; Hashimoto, N.; Ukai, S.; Kimura, A.; Inoue, Masaki; Kaito, Takeji; Fujisawa, T.; Okuda, T.; Abe, F.

    2009-01-01

    For development of advanced ferritic ODS steels including high concentration of Cr and Al, the effect of minor alloying elements on fine dispersion of oxide particle was investigated. Microstructural analysis for Fe-16Cr-4Al-mY 2 O 3 -nZr or mHf due to TEM indicated that 0.3Zr or 0.6Hf are the optimum concentration. The mechanism of nano-sized oxide formation was also discussed. (author)

  18. Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

    Science.gov (United States)

    Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

    2017-12-01

    High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

  19. Modelling of zirconium alloys corrosion in LWRs

    International Nuclear Information System (INIS)

    Kritskij, V.G.; Berezina, I.G.; Kritskij, A.V.; Stjagkin, P.S.

    1999-01-01

    Chemical parameters, that exerted effect on Zr+1%Nb alloy corrosion and deserved consideration during reactor operation, were defined and a model was developed to describe the influence of physical and chemical parameters on zirconium alloys corrosion in nuclear power plants. The model is based on the correlation between the zirconium oxide solubility in high-temperature water under the influence of the chemical parameters and the measured values of fuel cladding corrosion under LWR conditions. The intensity of fuel cladding corrosion in the primary circuits depends on the coolant water quality, growth of iron oxide deposits and vaporization portion. Mathematically, the oxidation rate can be expressed as a sum of heat and radiation components. The temperature dependence on the oxidation rate can be described by the Arrenius equation. The radiation component of Zr uniform corrosion equation is a function of several factors such as neutron fluency, the temperature the metallurgical composition and et. We assume that the main factor is the changing of water chemistry and the H 2 O 2 concentration play the determinative role. Probably, the influence of H 2 O 2 is based on the formation of unstable compound ZrO 3 ·nH 2 O and Zr(OH) 4 with high solubility. The validity of the used formulae was confirmed by corrosion measurements on WWER and RBMK fuel cladding. The model can be applied for calculating the reliability of nuclear fuel operation. (author)

  20. Microstructures and properties of ceramic particle-reinforced metal matrix composite layers produced by laser cladding

    Science.gov (United States)

    Zhang, Qingmao; He, Jingjiang; Liu, Wenjin; Zhong, Minlin

    2005-01-01

    Different weight ratio of titanium, zirconium, WC and Fe-based alloy powders were mixed, and cladded onto a medium carbon steel substrate using a 3kW continuous wave CO2 laser, aiming at producing Ceramic particles- reinforced metal matrix composites (MMCs) layers. The microstructures of the layers are typical hypoeutectic, and the major phases are Ni3Si2, TiSi2, Fe3C, FeNi, MC, Fe7Mo3, Fe3B, γ(residual austenite) and M(martensite). The microstructure morphologies of MMCs layers are dendrites/cells. The MC-type reinforcements are in situ synthesis Carbides which main compositions consist of transition elements Zr, Ti, W. The MC-type particles distributed within dendrite and interdendritic regions with different volume fractions for single and overlapping clad layers. The MMCs layers are dense and free of cracks with a good metallurgical bonding between the layer and substrate. The addition ratio of WC in the mixtures has the remarkable effect on the microhardness of clad layers.

  1. Cladding temperature measurement by thermocouples at preirradiated LWR fuel rod samples

    International Nuclear Information System (INIS)

    Leiling, W.

    1981-12-01

    This report describes the technique to measure cladding temperatures of test fuel rod samples, applied during the in-pile tests on fuel rod failure in the steam loop of the FR2 reactor. NiCr/Ni thermocouples with stainless steel and Inconel sheaths, respectively,of 1 mm diameter were resistance spot weld to the outside of the fuel rod cladding. For the pre-irradiated test specimens, welding had to be done under hot-cell conditions, i.e. under remote handling. In order to prevent the formation of eutectics between zirconium and the chemical elements of the thermocouple sheath at elevated temperatures, the thermocouples were covered with a platinum jacket of 1.4 mm outside diameter swaged onto the sheath in the area of the measuring junction. This thermocouple design has worked satisfactorily in the in-pile experiments performed in a steam atmosphere. Even in the heatup phase, in which cladding temperatures up to 1050 0 C were reached, only very few failures occured. This good performance is to a great part due to a careful control and a thorough inspection of the thermocouples. (orig.) [de

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

    International Nuclear Information System (INIS)

    Babiche, Ivan

    2013-01-01

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

  3. 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

  4. Joint titrimetric determination of zirconium and hafnium

    International Nuclear Information System (INIS)

    Vazquez, Cristina; Botbol, Moises; Bianco de Salas, G.N.; Cornell de Casas, M.I.

    1980-01-01

    A method for the joint titrimetric determination of zirconium and hafnium, which are elements of similar chemical behaviour, is described. The disodic salt of the ethylendiaminetetracetic acid (EDTA) is used for titration, while xilenol orange serves as final point indicator. Prior to titration it is important to evaporate with sulfuric acid, the solution resulting from the zirconium depolymerization process, to adjust the acidity and to eliminate any interferences. The method, that allows the quick and precise determination of zirconium and hafnium in quantities comprised between 0.01 and mg, was applied to the analysis of raw materials and of intermediate and final products in the fabrication of zirconium sponge and zircaloy. (M.E.L.) [es

  5. Evaluation of a Zirconium Recycle Scrubber System

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from a synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.

  6. Anticorrosion ion implantation of fragments of zirconium fuel can specimens

    International Nuclear Information System (INIS)

    Kalin, B.A.; Osipov, V.V.; Volkov, N.V.; Khernov, V.Yu.

    2001-01-01

    Aimed at the study of specific features of oxide film formation in the initial stage of Eh110 and Eh635 alloy fuel can oxidation the modification of tubular specimen surfaces is performed using an ion mixing technique, and the structure of oxide films produced in a steam-water environment is investigated. Using the method of vacuum vapor deposition the outer surface of specimens is coated with alloying element films irradiated by a polyenergetic Ar + ion beam with a 10 keV mean energy up to radiation doses of (7-10) x 10 17 ion/cm 2 . Monatomic (Al, Fe, Cu, Cr, Mo, Sn) or diatomic (Al-Fe, Al-Mo, Al-Sn, Fe-Cu, Fe-Mo, Fe-Sn, Cr-Mo, Cr-Sn) implantation into a zirconium cladding occurs under irradiation effect. The positive influence of combined intrusion of Al and other elements is revealed. The presence of Al atoms enhances the oxide film structure. The least ZeO 2 film thickness is observed when alloying with molybdenum, Al-Fe, Al-Mo and Al-Sn [ru

  7. Study of the processes for of remelting zirconium alloys in an electric arc furnace

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Luiz A.T.; Rossi, Jesualdo L.; Costa, Guilherme R.; Martinez, Luis G.; Sato, Ivone M., E-mail: luiz.atp@uol.com.br, E-mail: jelrossi@ipen.br, E-mail: guilhermeramoscosta@gmail.com, E-mail: lgallego@ipen.br, E-mail: imsato@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Zirconium alloy tubes are used as cladding for fuel elements of PWR nuclear reactors, which contains the UO{sub 2} pellets. In the manufacture of these fuel element parts, machining chips from the nuclear grade zirconium alloys are generated. Hence, these machining chips cannot be discarded, as ordinary metallic waste. Thus, the recycling of this material is a strategic aspect for the nuclear technology, both for economic and environmental issues. The main reason is that nuclear grade alloys have very high cost, are not commercially produced in Brazil and has to be imported for the manufacture of the nuclear fuels. This work discusses a method to melt and recycle Zircaloy chips, using an electric-arc furnace to obtain small laboratory ingots. The chemical composition of the ingots was determined using X-ray fluorescence spectroscopy and was compared to the specifications of nuclear grade Zircaloy and to the chemical composition of the received machining chips. The ingots were annealed in high vacuum, as well as were hot rolled in a mill. The microstructures were characterized by optical microscopy. The hardness was evaluated using the Rockwell B scale hardness. The results showed that the compositions of the recycled Zircaloy comply with the chemical specifications and a suitable microstructure has been obtained for nuclear use. (author)

  8. Influence de l’irradiation et de la radiolyse sur la vitesse et les mécanismes de corrosion des alliages de zirconium

    OpenAIRE

    Verlet , Romain

    2015-01-01

    The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO2 pe...

  9. Method of reducing zirconium

    International Nuclear Information System (INIS)

    Megy, J.A.

    1980-01-01

    A method was developed for making nuclear-grade zirconium from a zirconium compound, which ismore economical than previous methods since it uses aluminum as the reductant metal rather than the more expensive magnesium. A fused salt phase containing the zirconium compound to be reduced is first prepared. The fused salt phase is then contacted with a molten metal phase which contains aluminum and zinc. The reduction is effected by mutual displacment. Aluminum is transported from the molten metal phase to the fused salt phase, replacing zirconium in the salt. Zirconium is transported from the fused salt phase to the molten metal phase. The fused salt phase and the molten metal phase are then separated, and the solvent metal and zirconium are separated by distillation or other means. (DN)

  10. Waterside corrosion of zirconium alloys in nuclear power plants

    International Nuclear Information System (INIS)

    1998-01-01

    Technically the study of corrosion of zirconium alloys in nuclear power reactors is a very active field and both experimental work and understanding of the mechanisms involved are going through rapid changes. As a result, the lifetime of any publication in this area is short. Because of this it has been decided to revise IAEA-TECDOC-684 - Corrosion of Zirconium Alloys in Nuclear Power Plants - published in 1993. This updated, revised and enlarged version includes major changes to incorporate some of the comments received about the first version. Since this review deals exclusively with the corrosion of zirconium and zirconium based alloys in water, and another separate publication is planned to deal with the fuel-side corrosion of zirconium based fuel cladding alloys, i.e. stress corrosion cracking, it was decided to change the original title to Waterside Corrosion of Zirconium Alloys in Nuclear Power Plants. The rapid changes in the field have again necessitated a cut-off date for incorporating new data. This edition incorporates data up to the end of 1995; including results presented at the 11 International Symposium on Zirconium in the Nuclear Industry held in Garmisch-Partenkirchen, Germany, in September 1995. The revised format of the review now includes: Introductory chapters on basic zirconium metallurgy and oxidation theory; A revised chapter discussing the present extent of our knowledge of the corrosion mechanism based on laboratory experiments; a separate and revised chapter discussing hydrogen uptake; a completely reorganized chapter summarizing the phenomenological observations of zirconium alloy corrosion in reactors; a new chapter on modelling in-reactor corrosion; a revised chapter devoted exclusively to the manner in which irradiation might influence the corrosion process; finally, a summary of our present understanding of the corrosion mechanisms operating in reactor

  11. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys; Approche experimentale et modelisation micromecanique du comportement des alliages de zirconium irradies

    Energy Technology Data Exchange (ETDEWEB)

    Onimus, F

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  12. 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.

  13. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

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

  14. PLUTONIUM-ZIRCONIUM ALLOYS

    Science.gov (United States)

    Schonfeld, F.W.; Waber, J.T.

    1960-08-30

    A series of nuclear reactor fuel alloys consisting of from about 5 to about 50 at.% zirconium (or higher zirconium alloys such as Zircaloy), balance plutonium, and having the structural composition of a plutonium are described. Zirconium is a satisfactory diluent because it alloys readily with plutonium and has desirable nuclear properties. Additional advantages are corrosion resistance, excellent fabrication propenties, an isotropie structure, and initial softness.

  15. The development of zirconium alloy and its manufacturing

    International Nuclear Information System (INIS)

    Yuan Gaihuan; Yue Qiang

    2015-01-01

    Nuclear power which acts as one of low-carbon energy resources is the most realistic in large-scale application. It is also the preferred choice for many countries to develop energy resources and optimize its structure. Zirconium alloy is a key structural material for nuclear power plant fuel assemblies and cladding tubes of zirconium alloy are often referred as the first safeguard to nuclear power safety. With the development of nuclear power, three kinds of zirconium alloys Zr-Sn, Zr-Nb, Zr-Sn-Nb and with the representative products of Zr-4, M5, Zirlo respectively are developed and widely applied. Because of its severe operating environment and influence to nuclear safety, the requirements to zirconium alloys for physical and chemical properties, nuclear capability, tolerance and surface quality are very strict. The in-depth research and its manufacture capability become one of the main barriers for many countries who are developing the nuclear energy. In recent years, a stated-owned company, State Nuclear Bao Ti Zirconium Industry Company ('SNZ' for short) as well as National R and D Center for Nuclear Grade Zirconium material, is founded to meet the requirement of the rapid development of China's nuclear power industry. SNZ is dedicated for the fabrication and the research of nuclear grade zirconium products. After the successful completion of technology transfer of manufacturing for production chain and fully grasped of the manufacturing technology for the nuclear grade zirconium sponge through zirconium alloy tube, rod and strip products. National R and D Center for Nuclear Grade Zirconium material is cooperating with universities, nuclear energy research and design institutes and the owners of nuclear power plant to develop new zirconium alloy of self-owned brand. Through the selection of components, in-process testing and product inspection, four kinds of new zirconium alloys owns better performance than currently commercialized M5, Zirlo etc

  16. Study of corrosion kinetics of fuel element tubes from calcium-thermal zirconium alloy Zr1Nb in water at 350 degree C and in vapour at 400 and 500 degree C

    International Nuclear Information System (INIS)

    Petel'guzov, I.A.

    2002-01-01

    In the report brought results of corrosion process studies in water medium of pipe samples for fuel element shells from Zr1Nb alloy (earlier KTZ-110),made from the calcium-thermal zirconium alloys developed in the Ukraine of technology and,for the comparison,samples of pipes from the staff alloy E110, applicable in fuel elements acting reactors of type WWER. Tests were conducted under the working temperature of fuel shells in the reactor (350 degree C) in during of 14000 hours and under increased temperatures (400 degree C) within a time acordinly 4000 hours. Samples from the alloy Zr1Nb had more high contents of oxygen (before 0,12%...0,16%), than staff alloy Eh110 (0,08%O). Studies have shown sufficiently high corrosion stability of experimental alloy Zr1Nb, close to stability of alloy E110.Discovered signs of corrosion 'breakway' or 'transition' on kinetic corrosion curves of Zr1Nb alloys and E110 alloy, characterisating zircaloy type of alloy. Considered mechanism of influence of oxygen on the corrosion process of zirconium alloys with the additive a niobium

  17. Sensitivity analysis on the zirconium ignition in a postulated SFP loss of coolant accident

    International Nuclear Information System (INIS)

    Park, Sanggil; Lee, Jaeyoung; Kim, Sun-ki; Chun, Tae-hyun; Bang, Je-geon

    2016-01-01

    From both SFP complete LOCA experiments, it was observed that zirconium alloy cladding temperature was abruptly increased at a certain point and the cladding was almost fully oxidized. To capture this phenomenon, the concept of air oxidation breakaway model was adopted in MELCOR code. This paper examines this air oxidation breakaway model by comparing the SFP project test data and MELCOR code calculation results by using this model. The air oxidation model parameters are slightly altered to see their sensitivities on the occurrence of the zirconium ignition. Through such sensitivity analysis, limitations of the air oxidation breakaway model are revealed in comparison to the actual zirconium ignition phenomenon during air ingress scenarios. In addition, ways to overcome the identified limitations of the air oxidation model are recommended to estimate better the zirconium ignition phenomenon in SFP sequences. In this paper, the zirconium ignition phenomenon was reviewed and the model to capture this phenomenon was investigated. The model is the air oxidation breakaway model in MELCOR code, and its sensitivity of the model parameters on the time to ignition was studied. From the sensitivity analysis, the slight change of model parameters induce the large variation of the time to ignition. The model itself includes its weakness to fully represent both the air oxidation breakaway phenomenon and the followed zirconium ignition behavior. Furthermore, this model considers no effect of N2 on the cladding degradation and its promoted exothermic heat release

  18. Sensitivity analysis on the zirconium ignition in a postulated SFP loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sanggil; Lee, Jaeyoung [Handong Global Univ., Pohang (Korea, Republic of); Kim, Sun-ki; Chun, Tae-hyun; Bang, Je-geon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    From both SFP complete LOCA experiments, it was observed that zirconium alloy cladding temperature was abruptly increased at a certain point and the cladding was almost fully oxidized. To capture this phenomenon, the concept of air oxidation breakaway model was adopted in MELCOR code. This paper examines this air oxidation breakaway model by comparing the SFP project test data and MELCOR code calculation results by using this model. The air oxidation model parameters are slightly altered to see their sensitivities on the occurrence of the zirconium ignition. Through such sensitivity analysis, limitations of the air oxidation breakaway model are revealed in comparison to the actual zirconium ignition phenomenon during air ingress scenarios. In addition, ways to overcome the identified limitations of the air oxidation model are recommended to estimate better the zirconium ignition phenomenon in SFP sequences. In this paper, the zirconium ignition phenomenon was reviewed and the model to capture this phenomenon was investigated. The model is the air oxidation breakaway model in MELCOR code, and its sensitivity of the model parameters on the time to ignition was studied. From the sensitivity analysis, the slight change of model parameters induce the large variation of the time to ignition. The model itself includes its weakness to fully represent both the air oxidation breakaway phenomenon and the followed zirconium ignition behavior. Furthermore, this model considers no effect of N2 on the cladding degradation and its promoted exothermic heat release.

  19. Status and task of the study on the hydrogen embrittlement of zirconium alloys

    International Nuclear Information System (INIS)

    Nagase, Fumihisa; Furuta, Teruo; Seino, Shun; Komatsu, Kazushi.

    1995-08-01

    As the burnup of the LWR fuel is extended, waterside corrosion and hydrogen pickup increase in the Zircaloy cladding. Hydrogen embrittlement of Zircaloy is one of the main factors which may limit the life of the fuel rod. This report presents a review on the hydrogen embrittlement of zirconium and its alloys including the irradiated materials. Research tasks for the reduction of ductility in the high burnup fuel cladding are also discussed. Many fundamental investigations have been performed on the hydrogen embrittlement of zirconium alloys. However, the embrittlement mechanism of the high burnup fuel cladding is complicated. Especially, a coupled effect of hydrides and radiation defects are expected to be pronounced with neutron dose increase. In order to evaluate the reduction of ductility of the higher burnup fuel cladding properly, it is necessary to investigate the coupled effect of these two factors by systematic examinations. (author) 64 refs

  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. 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

  2. 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

  3. 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)

  4. 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)

  5. 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)

  6. Choices of canisters and elements for the first fuel and canister sludge shipment from K East Basin

    International Nuclear Information System (INIS)

    Makenas, B.J.

    1996-01-01

    The K East Basin contains open-top canisters with up to fourteen N Reactor fuel assemblies distributed between the two barrels of each canister. Each fuel assembly generally consists of inner and outer concentric elements fabricated from uranium metal with zirconium alloy cladding. The canisters also contain varying amounts of accumulated sludge. Retrieval of sample fuel elements and associated sludge for examination is scheduled to occur in the near future. The purpose of this document is to specify particular canisters and elements of interest as candidate sources of fuel and sludge to be shipped to laboratories

  7. Thermofluency in zirconium alloys

    International Nuclear Information System (INIS)

    Orozco M, E.A.

    1976-01-01

    A summary is presented about the theoretical and experimental results obtained at present in thermofluency under radiation in zirconium alloys. The phenomenon of thermofluency is presented in a general form, underlining the thermofluency at high temperature because this phenomenon is similar to the thermofluency under radiation, which ocurrs in zirconium alloys into the operating reactor. (author)

  8. Modeling of Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

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

    1989-07-01

    Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs

  9. A microstuctural study on accelerated zirconium alloy oxidation

    International Nuclear Information System (INIS)

    Sohn, Seung Bum; Oh, Seung Jun; Jang, Jung Nam; Kim, Yong Soo; Jung, Yong Hwan; Baek, Jong Hyuk; Park, Jung Yong

    2005-01-01

    It has been reported that the effect of thermal redistribution of hydrides across the zirconium metaloxide interface, coupled with thermal feedback on the metal-oxide interface, is a dominating factor in the accelerated oxidation in zirconium alloys cladding PWR fuel. Basically this influence determines characteristic of oxide layer. Influence estimation for corrosion oxide layer due to hydrogen / hydride carried out because of investigation on the kinetic on accelerated oxidation due to hydride precipitation was preceded. Generally, it is known that ZrO 2 tetragonal layer structures play an important role as a barrier layer. So analysing the ZrO 2 monoclinic and tetragonal structure distribution is our main aim. Especially, this study focused on the hydride effects. In other words, the difference of crystal structure distribution between pre-hydrided and without hydrided specimen is just expected results. Experimental results of microstructure at zirconium metal-oxide interface through TEM and EBSD analysis was confirmed

  10. Contribution to the study of zirconium self-diffusion in zirconium carbide

    International Nuclear Information System (INIS)

    An, Chul

    1972-01-01

    The objective of this research thesis is to determine experimental conditions allowing the measurement of the self-diffusion coefficient of zirconium in zirconium carbide. The author reports the development of a method of preparation of zirconium carbide samples. He reports the use of ion implantation as technique to obtain a radio-tracer coating. The obtained results give evidence of the impossibility to use sintered samples with small grains because of the demonstrated importance of intergranular diffusion. The self-diffusion coefficient is obtained in the case of zirconium carbide with grains having a diameter of few millimetres. The presence of 95 Nb from the disintegration of 95 Zr indicates that these both metallic elements have very close diffusion coefficients at 2.600 C [fr

  11. 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.

  12. Coupling between Experimental Measurements and Finite Element Calculations for identification of crystallographic constitutive law. Application to zirconium alloys; Methode de couplage entre experimentations et simulations numeriques en vue de l'identification de lois de comportement intracristallin. Application aux alliages de zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Dexet, M

    2006-10-15

    This thesis presents a methodology for multi scale coupling between the morphology and texture of a microstructure as has been characterised experimentally, and the results of mechanical strain field analysis. This methodology is based on a coupling between experimental characterisation of the microstructure, ex-situ mechanical tests, local strain field measurements performed at the grain scale, and finite element simulations. Then, a definition of a cost function is proposed in order to optimise the parameters of the crystallographic constitutive law. This method is applied to the studies of zirconium alloys in order to improve the understanding of their mechanical behaviour in relation with their microstructures, which is a key requirement for their use in the nuclear industries. This work was funded by the joint research program SMIRN between EDF, CEA and CNRS. (author)

  13. Experimental approach and micro-mechanical modeling of the mechanical behavior of irradiated zirconium alloys

    International Nuclear Information System (INIS)

    Onimus, F.

    2003-12-01

    Zirconium alloys cladding tubes containing nuclear fuel of the Pressurized Water Reactors constitute the first safety barrier against the dissemination of radioactive elements. Thus, it is essential to predict the mechanical behavior of the material in-reactor conditions. This study aims, on the one hand, to identify and characterize the mechanisms of the plastic deformation of irradiated zirconium alloys and, on the other hand, to propose a micro-mechanical modeling based on these mechanisms. The experimental analysis shows that, for the irradiated material, the plastic deformation occurs by dislocation channeling. For transverse tensile test and internal pressure test this channeling occurs in the basal planes. However, for axial tensile test, the study revealed that the plastic deformation also occurs by channeling but in the prismatic and pyramidal planes. In addition, the study of the macroscopic mechanical behavior, compared to the deformation mechanisms observed by TEM, suggested that the internal stress is higher in the case of irradiated material than in the case of non-irradiated material, because of the very heterogeneous character of the plastic deformation. This analysis led to a coherent interpretation of the mechanical behavior of irradiated materials, in terms of deformation mechanisms. The mechanical behavior of irradiated materials was finally modeled by applying homogenization methods for heterogeneous materials. This model is able to reproduce adequately the mechanical behavior of the irradiated material, in agreement with the TEM observations. (author)

  14. 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.

  15. A spanish mineral of zirconium and hafnium. Separation of the two elements by liquid-liquid extraction, using tributyl phosphate as chelating agent

    International Nuclear Information System (INIS)

    Ruiz Sanchez, F.; Cruz Castillo, F. de la; Fernandez Cellini, R.

    1962-01-01

    The zirconium and Hafnium oxides are obtained from a Spanish mineral of zircon with an average contest of 55% in ZrO 2 -HfO 2 . An alkaline fusion to open the mineral, followed by a purification by crystallization as (Zr O-Hf O)Cl 2 H 2 O or as (Zr-Hf) (SO 4 ) 2 . 4H 2 O, is used. A discussion of the best experimental conditions for opening the mineral and of the purification method is made. (Author) 45 refs

  16. Nuclear fuel element

    International Nuclear Information System (INIS)

    Penrose, R.T.; Thompson, J.R.

    1976-01-01

    A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

  17. Determination of impurities in zirconium by emission spectrograph method

    International Nuclear Information System (INIS)

    Simbolon, S.; Masduki, B.; Aryadi

    2000-01-01

    Analysis of B, Cd, Si and Cr elements in zirconium oxide was carried out. Zirconium oxide was made by precipitating zirconium solution with oxalic acid and calcination was at temperature 900 oC for four hours. Silver chloride compound as much as 10% was used as a distillation carrier and 7 step filtration was used to reduce the impurities element spectra having high density. It was found that B concentration is between 3.80 and 7.44 ppm, Cd less then 0.5 ppm, Si between 74.38-150.33 ppm and Cr between 19.90-45.76 ppm. (author)

  18. Optimized manufacture of nuclear fuel cladding tubes by FEA of hot extrusion and cold pilgering processes

    Science.gov (United States)

    Gaillac, Alexis; Ly, Céline

    2018-05-01

    Within the forming route of Zirconium alloy cladding tubes, hot extrusion is used to deform the forged billets into tube hollows, which are then cold rolled to produce the final tubes with the suitable properties for in-reactor use. The hot extrusion goals are to give the appropriate geometry for cold pilgering, without creating surface defects and microstructural heterogeneities which are detrimental for subsequent rolling. In order to ensure a good quality of the tube hollows, hot extrusion parameters have to be carefully chosen. For this purpose, finite element models are used in addition to experimental tests. These models can take into account the thermo-mechanical coupling conditions obtained in the tube and the tools during extrusion, and provide a good prediction of the extrusion load and the thermo-mechanical history of the extruded product. This last result can be used to calculate the fragmentation of the microstructure in the die and the meta-dynamic recrystallization after extrusion. To further optimize the manufacturing route, a numerical model of the cold pilgering process is also applied, taking into account the complex geometry of the tools and the pseudo-steady state rolling sequence of this incremental forming process. The strain and stress history of the tube during rolling can then be used to assess the damage risk thanks to the use of ductile damage models. Once validated vs. experimental data, both numerical models were used to optimize the manufacturing route and the quality of zirconium cladding tubes. This goal was achieved by selecting hot extrusion parameters giving better recrystallized microstructure that improves the subsequent formability. Cold pilgering parameters were also optimized in order to reduce the potential ductile damage in the cold rolled tubes.

  19. Zirconium (IV) complexes with some polymethylenediimines | Na ...

    African Journals Online (AJOL)

    The syntheses of zirconium (IV) complexes have been carried out by the reaction of oxozirconium (IV) chloride with the appropriate diimines (Schiff bases). The complexes were isolated as yellow solids which are stable to heat. The complexes were found to be insoluble in most solvents. The infrared spectra, elemental ...

  20. Determination of microquantities of zirconium and thorium in uranium dioxide

    International Nuclear Information System (INIS)

    Weber de D'Alessio, Ana; Zucal, Raquel.

    1975-07-01

    A method for the determination of 10 to 50 ppm of zirconium and thorium in uranium IV oxide of nuclear purity is established. Zirconium and thorium are retained in a strong cation-exchange resin Dowex 50 WX8 in 1 M HCl. Zirconium is eluted with 0,5% oxalic acid solution and thorium with 4% ammonium oxalate. The colorimetric determination of zirconium with xilenol orange is done in perchloric acid after destructtion of oxalic acid and thorium is determined with arsenazo III in 5 M HCl. 10 μg of each element were determined with a standard deviation of 2,1% for thorium and 3,4% for zirconium. (author) [es

  1. Fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1976-01-01

    A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

  2. 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

  3. Evaluation of long-term creep behaviour on K-cladding tubes

    International Nuclear Information System (INIS)

    Bang, J. G.; Jeong, Y. H.; Jeong, Y. H.

    2003-01-01

    KAERI has developed new zirconium alloys for high burnup fuel cladding. To evaluate the performance of these alloys, various out-pile tests are conducting. At high burnup, the creep resistance as well as corrosion resistance is one of the major factors determining nuclear fuel performance. Long-term creep test was performed at 350 .deg. C and 400 .deg. C and 100, 120, 135, and 150 MPa of applied hoop stress to evaluate the creep properties. The creep resistance was strongly affected by the final heat treatment conditions, while there was no effect of intermediate heat treatment. The creep strain of the recrystallized alloys is higher than that of the stress-relieved alloys by a factor of 3. The alloying elements also influenced the creep behaviour. Increase of Sn content enhanced the creep resistance, while Nb decreased the creep resistance. As a result of texture analysis, basal pole was directed to normal direction, while prism pole was to rolling direction. The development of the deformation texture and the ammealing texture showed almost similar process to Zircaloy cladding

  4. SEPARATING HAFNIUM FROM ZIRCONIUM

    Science.gov (United States)

    Lister, B.A.J.; Duncan, J.F.

    1956-08-21

    A dilute aqueous solution of zirconyl chloride which is 1N to 2N in HCl is passed through a column of a cation exchange resin in acid form thereby absorbing both zirconium and associated hafnium impurity in the mesin. The cation exchange material with the absorbate is then eluted with aqueous sulfuric acid of a O.8N to 1.2N strength. The first portion of the eluate contains the zirconium substantially free of hafnium.

  5. 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

  6. Gel structure of the corrosion layer on cladding pipes of nuclear fuels

    Czech Academy of Sciences Publication Activity Database

    Medek, Jiří; Weishauptová, Zuzana

    2009-01-01

    Roč. 393, č. 2 (2009), s. 306-310 ISSN 0022-3115 R&D Projects: GA ČR GA106/04/0043 Institutional research plan: CEZ:AV0Z30460519 Keywords : cladding pipes of nuclear fuels * corrosion layer * zirconium alloys Subject RIV: JF - Nuclear Energetics Impact factor: 1.933, year: 2009

  7. In situ Investigation of Oxide Films on Zirconium Alloy in PWR Primary Water Chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Taeho; Choi, Kyoung Joon; Yoo, Seung Chang; Kim, Ji Hyun [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2015-05-15

    Zirconium alloys are used as fuel cladding materials in nuclear power reactors, because these materials have a very low thermal neutron capture cross section as well as desirable mechanical properties. However, the Fukushima accident shows that the oxidation behavior of zirconium alloy is an important issue because the zirconium alloy functions as a shield of nuclear material (i.e., uranium, fission gas), and the degradation on zirconium cladding directly causes severe accident on nuclear power plant. Therefore, to ensure the safety of nuclear power reactors, the performance and sustainability of nuclear fuel should be understood. Currently, the water-metal interface is regarded as the rate-controlling site governing the rapid oxidation transition in high-burn-up fuels. Zirconium oxide is formed at the water-metal interface, and its structure and phase play an important role in determining its mechanical properties. In the early stage of the oxidation process, zirconium oxide with both tetragonal and monoclinic phases is formed. With an increase in the oxidation time to 150 h, the unstable tetragonal phase disappears and the monoclinic phase is dominant and possibly because of the stress relaxation according to previous and present results.

  8. 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.

  9. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  12. 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)

  13. 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.

  14. 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

  15. Zirconium isotope separation process

    International Nuclear Information System (INIS)

    Peterson, S.H.; Lahoda, E.J.

    1988-01-01

    A process is described for reducing the amount of zirconium 91 isotope in zirconium comprising: forming a first solution of (a) a first solvent, (b) a scavenger, and (c) a zirconium compound which is soluble in the first solvent and reacts with the scavenger when exposed to light of a wavelength of 220 to 600 nm; irradiating the first solution with light at the wavelength for a time sufficient to photoreact a disproportionate amount of the zirconium compound containing the zirconium 91 isotope with the scavenger to form a reaction product in the first solution; contacting the first solution, while effecting the irradiation, with a second solvent which is immiscible with the first solvent, which the second solvent is a preferential solvent for the reaction product relative to the first solvent, such that at least a portion of the reaction product is transferred to the second solvent to form a second solution; and separating the second solution from the first solution after the contacting

  16. PROCESS FOR DISSOLVING BINARY URANIUM-ZIRCONIUM OR ZIRCONIUM-BASE ALLOYS

    Science.gov (United States)

    Jonke, A.A.; Barghusen, J.J.; Levitz, N.M.

    1962-08-14

    A process of dissolving uranium-- zirconium and zircaloy alloys, e.g. jackets of fuel elements, with an anhydrous hydrogen fluoride containing from 10 to 32% by weight of hydrogen chloride at between 400 and 450 deg C., preferably while in contact with a fluidized inert powder, such as calcium fluoride is described. (AEC)

  17. Design of an integrated system to recycle Zircaloy cladding using a hydride–milling–dehydride process

    Energy Technology Data Exchange (ETDEWEB)

    Kelley, Randy, E-mail: rkelley@pitt.edu [Mechanical Engineering Department, 236 Engineering and Science Building, University of Pittsburgh – Johnstown, Johnstown, PA 15904 (United States); McDeavitt, Sean [Texas A and M University, Department of Nuclear Engineering, 327 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States)

    2013-10-15

    Highlights: • Dehydriding zirconium hydride was studied at relatively low temperatures (<800 °C). • High vacuum pressures decrease dehydriding temperatures. • Specialized equipment was designed, built and demonstrated to process zirconium. • The process hydrided–milled–dehydrided zirconium metal to a fine metal powder. • Two powder samples were analyzed and proved the operation of the machine. -- Abstract: A hydride–dehydride process was evaluated to recover a portion of spent nuclear fuel cladding; a zirconium alloy (Zircaloy), as a metal powder that may be used for advanced nuclear fuel applications. The investigation was part of a broader study that sought to determine the viability of recovering components of used nuclear fuel to for a metal matrix cermet for transuranic burning. The zirconium powder process begins with the conversion of Zircaloy cladding hulls into a brittle zirconium hydride, which is easily pulverized into a powder. The dehydriding process removes hydrogen by heating the powder in a vacuum, resulting in a zirconium metal powder. In support of this, a specialized piece of equipment was designed to demonstrate the entire zirconium conversion process to transform Zircaloy tubes into metal powder without intermediate handling. This was accomplished by building a milling system that rotates inside of controlled atmosphere chamber with an internal heater. The hydriding process was accomplished using an argon–5% hydrogen atmosphere at 500 °C. The process variables for the dehydriding process were determined using a thermogavimetric analysis (TGA) method. It was determined that a rough vacuum (∼0.001 bar) and 800 °C were sufficient to decompose the zirconium hydride. Zirconium metal powder was created using different milling times: 45 min (coarse powder) and 12 h (fine powder). X-ray diffraction (XRD) analysis indicated that the process produced a zirconium metal. Additionally, visual observations of the samples silvery

  18. Metallurgical and mechanical behaviours of PWR fuel cladding tube oxidised at high temperature; Comportements metallurqigue et mecanique des materiaux de gainage du combustible REP oxydes a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Stern, A

    2007-12-15

    Zirconium alloys are used as cladding materials in Pressurized Water Reactors (PWR). As they are submitted to very extreme conditions, it is necessary to check their behaviour and especially to make sure they meet the safety criteria. They are therefore studied under typical in service-loadings but also under accidental loadings. In one of these accidental scenarios, called Loss of Coolant Accident (LOCA) the cladding temperature may increase above 800 C, in a steam environment, and decrease before a final quench of the cladding. During this temperature transient, the cladding is heavily oxidised, and the metallurgical changes lead to a decrease of the post quench mechanical properties. It is then necessary to correlate this drop in residual ductility to the metallurgical evolutions. This is the problem we want to address in this study: the oxidation of PWR cladding materials at high temperature in a steam environment and its consequences on post quench mechanical properties. As oxygen goes massively into the metallic part - a zirconia layer grows at the same time - during the high temperature oxidation, the claddings tubes microstructure shows three different phases that are the outer oxide layer (zirconia) and the inner metallic phases ({alpha}(O) and 'ex {beta}') - with various mechanical properties. In order to reproduce the behaviour of this multilayered material, the first part of this study consisted in creating samples with different - but homogeneous in thickness - oxygen contents, similar to those observed in the different phases of the real cladding. The study was especially focused on the {beta}-->{alpha} phase transformation upon cooling and on the resulting microstructures. A mechanism was proposed to describe this phase transformation. For instance, we conclude that for our oxygen enriched samples, the phase transformation kinetics upon cooling are ruled by the oxygen partitioning between the two allotropic phases. Then, these materials

  19. Purification of zirconium concentrates

    International Nuclear Information System (INIS)

    Brown, A.E.P.

    1976-01-01

    A commercial grade ZrO 2 and an ammonium uranate (yellow cake) are obtained from the caldasito ore processing. This ore is found in the Pocos de Caldas Plateau, State of Minas Gerais, Brazil. Caldasito is an uranigerous zirconium ore, a mixture of zircon and baddeleyite and contains 60% ZrO 2 and 0,3% U 3 O 8 . The chemical opening of the ore was made by alkaline fusion with NaOH at controlled temperature. The zirconium-uranium separation took place by a continuous liquid-liquid extraction in TBP-varsol-HNO 3 -H 2 O system. The raffinate containing zirconium + impurities (aluminium, iron and titanium) was purified by an ion exchange operation using a strong cationic resin [pt

  20. Synthesis of Zirconium Lower Chlorides

    International Nuclear Information System (INIS)

    Gaviria, Juan P.

    2002-01-01

    This research is accurately related to the Halox concept of research reactor spent fuel element treatment.The aim of this project is to work the conditioning through selected chlorination of the element that make the spent fuel element. This research studied the physical chemistry conditions which produce formation of the lower zirconium chlorides through the reaction between metallic Zr and gaseous ZrCl 4 in a silica reactor.This work focused special attention in the analysis and confrontation of the published results among the different authors in order to reveal coincidences and contradictions.Experimental section consisted in a set of synthesis with different reaction conditions and reactor design. After reaction were analyzed the products on Zr shavings and the deposit growth on wall reactor.The products were strongly dependent of reactor design. It was observed that as the distance between Zr and wall reactor increased greater was tendency to lower chlorides formation.In reactors with small distance the reaction follows other way without formation of lower chlorides.Analysis on deposit growth on reactor showed that may be formed to a mixture of Si x Zr y intermetallics and zirconium oxides.Presence of oxygen in Zr and Zr-Si compounds on wall reactor reveals that there is an interaction between quartz and reactants.This interaction is in gaseous phase because contamination is observed in experiences where Zr was not in contact with reactor.Finally, it was made a global analysis of all experiences and a possible mechanism that interprets reaction ways is proposed

  1. Highly corrosion resistant zirconium based alloy for reactor structural material

    International Nuclear Information System (INIS)

    Ito, Yoichi.

    1996-01-01

    The alloy of the present invention is a zirconium based alloy comprising tin (Sn), chromium (Cr), nickel (Ni) and iron (Fe) in zirconium (Zr). The amount of silicon (Si) as an impurity is not more than 60ppm. It is preferred that Sn is from 0.9 to 1.5wt%, that of Cr is from 0.05 to 0.15wt%, and (Fe + Ni) is from 0.17 to 0.5wt%. If not less than 0.12wt% of Fe is added, resistance against nodular corrosion is improved. The upper limit of Fe is preferably 0.40wt% from a view point of uniform suppression for the corrosion. The nodular corrosion can be suppressed by reducing the amount of Si-rich deposition product in the zirconium based alloy. Accordingly, a highly corrosion resistant zirconium based alloy improved for the corrosion resistance of zircaloy-2 and usable for a fuel cladding tube of a BWR type reactor can be obtained. (I.N.)

  2. Solvent extraction of zirconium

    International Nuclear Information System (INIS)

    Kim, S.S.; Yoon, J.H.

    1981-01-01

    The extraction of zirconium(VI) from an aqueous solution of constant ionic strength with versatic acid-10 dissolved in benzen was studied as a function of pH and the concentration of zirconium(VI) and organic acid. The effects of sulphate and chlorine ions on the extraction of the zirconium(VI) were briefly examined. It was revealed that (ZrOR 2 .2RH) is the predominant species of extracted zirconium(VI) in the versatic acid-10. The chemical equation and the apparent equilibrium constants thereof have been determined as follows. (ZrOsup(2+))aq+ 2(R 2 H 2 )sub(org) = (ZrOR 2 .2RH)sub(org)+2(H + )aq Ksub(Zr) = (ZrOR 2 .2RH)sub(org)(H + ) 2 /(ZrOsup(2+))sub(aq)(R 2 H 2 )sup(2)sub(org) = 3.3 x 10 -7 . The synergistic effects of TBP and D2EHPA were also studied. In the mixed solvent with 0.1M TBP, the synergistic effect was observed, while the mixed solvent with D2EHPA showed the antisynergistic effect. (Author)

  3. Titanium and zirconium alloys

    International Nuclear Information System (INIS)

    Pinard Legry, G.

    1994-01-01

    Titanium and zirconium pure and base alloys are protected by an oxide film with anionic vacancies which gives a very good resistance to corrosion in oxidizing medium, in some ph ranges. Results of pitting and crevice corrosion are given for Cl - , Br - , I - ions concentration with temperature and ph dependence, also with oxygenated ions effect. (A.B.). 32 refs., 6 figs., 3 tabs

  4. Modelling zirconium hydrides using the special quasirandom structure approach

    KAUST Repository

    Wang, Hao; Chroneos, Alexander I.; Jiang, Chao; Schwingenschlö gl, Udo

    2013-01-01

    The study of the structure and properties of zirconium hydrides is important for understanding the embrittlement of zirconium alloys used as cladding in light water nuclear reactors. Simulation of the defect processes is complicated due to the random distribution of the hydrogen atoms. We propose the use of the special quasirandom structure approach as a computationally efficient way to describe this random distribution. We have generated six special quasirandom structure cells based on face centered cubic and face centered tetragonal unit cells to describe ZrH2-x (x = 0.25-0.5). Using density functional theory calculations we investigate the mechanical properties, stability, and electronic structure of the alloys. © the Owner Societies 2013.

  5. MAX Phase Modified SiC Composites for Ceramic-Metal Hybrid Cladding Tubes

    International Nuclear Information System (INIS)

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

    2015-01-01

    A metal-ceramic hybrid cladding consists of an inner zirconium tube, and an outer SiC fiber-matrix SiC ceramic composite with surface coating as shown in Fig. 1 (left-hand side). The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. In addition, the outermost layer prevents the dissolution of SiC during normal operation. On the other hand, a ceramic-metal hybrid cladding consists of an outer zirconium tube, and an inner SiC ceramic composite as shown in Fig. 1 (right-hand side). The outer zirconium protects the fuel rod from a corrosion during reactor operation, as in the present fuel claddings. The inner SiC composite, additionally, is designed to resist the severe oxidation under a postulated accident condition of a high-temperature steam environment. Reaction-bonded SiC was fabricated by modifying the matrix as the MAX phase. The formation of Ti 3 SiC 2 was investigated depending on the compositions of the preform and melt. In most cases, TiSi 2 was the preferential phase because of its lowest melting point in the Ti-Si-C system. The evidence of Ti 3 SiC 2 was the connection with the pressurizing

  6. Quercetin as colorimetric reagent for determination of zirconium

    Science.gov (United States)

    Grimaldi, F.S.; White, C.E.

    1953-01-01

    Methods described in the literature for the determination of zirconium are generally designed for relatively large amounts of this element. A good procedure using colorimetric reagent for the determination of trace amounts is desirable. Quercetin has been found to yield a sensitive color reaction with zirconium suitable for the determination of from 0.1 to 50?? of zirconium dioxide. The procedure developed involves the separation of zirconium from interfering elements by precipitation with p-dimethylaminoazophenylarsonic acid prior to its estimation with quercetin. The quercetin reaction is carried out in 0.5N hydrochloric acid solution. Under the operating conditions it is indicated that quercetin forms a 2 to 1 complex with zirconium; however, a 2 to 1 and a 1 to 1 complex can coexist under special conditions. Approximate values for the equilibrium constants of the complexes are K1 = 0.33 ?? 10-5 and K2 = 1.3 ?? 10-9. Seven Bureau of Standards samples of glass sands and refractories were analyzed with excellent results. The method described should find considerable application in the analysis of minerals and other materials for macro as well as micro amounts of zirconium.

  7. International strategic minerals inventory summary report; zirconium

    Science.gov (United States)

    Towner, R.R.

    1992-01-01

    Zircon, a zirconium silicate, is currently the most important commercial zirconium-bearing mineral. Baddeleyite, a natural form of zirconia, is less important but has some specific end uses. Both zircon and baddeleyite occur in hard-rock and placer deposits, but at present all zircon production is from placer deposits. Most baddeleyite production is from hard-rock deposits, principally as a byproduct of copper and phosphate-rock mining. World zirconium resources in identified, economically exploitable deposits are about 46 times current production rates. Of these resources, some 71 percent are in South Africa, Australia, and the United States. The principal end uses of zirconium minerals are in ceramic applications and as refractories, abrasives, and mold linings in foundries. A minor amount, mainly of zircon, is used for the production of hafnium-free zirconium metal, which is used principally for sheathing fuel elements in nuclear reactors and in the chemical-processing industry, aerospace engineering, and electronics. Australia and South Africa are the largest zircon producers and account for more than 70 percent of world output; the United States and the Soviet Union account for another 20 percent. South Africa accounts for almost all the world's production of baddeleyite, which is about 2 percent of world production of contained zirconia. Australia and South Africa are the largest exporters of zircon. Unless major new deposits are developed in countries that have not traditionally produced zircon, the pattern of world production is unlikely to change by 2020. The proportions, however, of production that come from existing producing countries may change somewhat.

  8. 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.)

  9. 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.

  10. Prospects for zirconium structural alloys at high temperatures

    International Nuclear Information System (INIS)

    Thomas, W.R.

    1969-05-01

    Improved station efficiencies and lower capital costs provide incentives for the development of zirconium alloys for pressure tubes which can operate at temperatures above 450 o C. The experience of the Ti industry indicates that a complex alloy containing solution hardeners of Sn or Al and precipitation hardeners of Mo and Nb and perhaps Si will be required. The thermal neutron cross-section of the alloy will be about 10% higher than Zircaloy-2 and because of its poor corrosion resistance will require cladding with a corrosion resistant alloy such as Zr-Cr. Results to date indicate that such a pressure tube is feasible. (author)

  11. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

    Yeo, D.

    1976-01-01

    A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

  12. 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

  13. The addition zirconium effect on the solubility and activity of sulfur in liquid iron

    International Nuclear Information System (INIS)

    Burylev, B.P.; Mojsov, L.P.

    1994-01-01

    Critical analysis of reference data on thermodynamic properties of zirconium sulfides is conducted for evaluation of zirconium desulfonation ability in liquid steel. Sulfur solubility dependence on zirconium concentration in liquid iron is presented. Curves of sulfur solubility in liquid iron in the presence of other elements, including titanium, manganese, vanadium and chromium are presented for comparison. It is shown that equilibrium concentration of sulfur is much lower than standard sulfur concentrations in steel, therefore zirconium appears to be the best desulfonator among the metals considered

  14. In-situ synthesized Ni–Zr intermetallic/ceramic reinforced composite coatings on zirconium substrate by high power diode laser

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Kun; Li, Yajiang, E-mail: yajli@sdu.edu.cn; Wang, Juan; Ma, Qunshuang

    2015-03-05

    Highlights: • In-situ synthesized Ni–Zr intermetallics/ceramic reinforced composite coatings. • Si enrichment and Ni replacing site of Si both resulted in forming Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4.} • Microstructure and forming of ZrB{sub 2} depended on affinity of elements and Si/B ratio. - Abstract: Ni–Zr intermetallic/ceramic reinforced composite coatings were in-situ synthesized by laser cladding series of Ni–Cr–B–Si powders on zirconium substrate. Microstructure, phase constituents and microhardness of coatings were investigated by means of optical microscope (OM), scanning electron microscope (SEM), energy dispersive spectrometer (EDS), X-ray diffraction (XRD) and microsclemeter. Results indicated that coatings with metallurgical bonding to substrate consisted of cellular NiZr matrix and massive reinforcements including NiZr{sub 2}, Zr{sub 5}(Si{sub x}Ni{sub 1−x}){sub 4} and ZrB{sub 2}. Morphologies of reinforcements were mainly dominated by temperature gradient and cooling rate from surface to bottom of the coating produced by same powder. In different coatings, microstructure and forming of ZrB{sub 2} mainly depended on affinity of elements and Si/B ratio in different powders. In addition, the mean microhardness of coatings up to 1200–1300 HV{sub 0.2} is nearly 7 times higher than that of R60702 zirconium substrate.

  15. Study of gap conductance model for thermo mechanical fully coupled finite element model

    International Nuclear Information System (INIS)

    Kim, Hyo Cha; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

    2012-01-01

    A light water reactor (LWR) fuel rod consists of zirconium alloy cladding and uranium dioxide pellets, with a slight gap between them. Therefore, the mechanical integrity of zirconium alloy cladding is the most critical issue, as it is an important barrier for fission products released into the environment. To evaluate the stress and strain of the cladding during operation, fuel performance codes with a one-dimensional (1D) approach have been reported since the 1970s. However, it is difficult for a 1D model to simulate the stress and strain of the cladding accurately owing to a lack of degree of freedom. A LWR fuel performance code should include thermo-mechanical coupled model owing to the existence of the fuel-cladding gap. Generally, the gap that is filled with helium gas results in temperature drop along radius direction. The gap conductance that determines temperature gradient within the gap is very sensitive to gap thickness. For instance, once the gap size increases up to several microns in certain region, difference of surface temperatures increases up to 100 Kelvin. Therefore, iterative thermo-mechanical coupled analysis is required to solve temperature distribution throughout pellet and cladding. Consequently, the Finite Element (FE) module, which can simulate a higher degree of freedom numerically, is an indispensable requirement to understand the thermomechanical behavior of cladding. FRAPCON-3, which is reliable performance code, has iterative loop for thermo-mechanical coupled calculation to solve 1D gap conductance model. In FEMAXI-III, 1D thermal analysis module and FE module for stress-strain analysis were separated. 1D thermal module includes iterative analysis between them. DIONISIO code focused on thermal contact model as function of surface roughness and contact pressure when the gap is closed. In previous works, gap conductance model has been developed only for 1D model or hybrid model (1D and FE). To simulate temperature, stress and strain

  16. On the stabilization of niobium(V) solutions by zirconium(IV) and hafnium(IV)

    DEFF Research Database (Denmark)

    Sørensen, E.; Bjerre, A.B.

    1992-01-01

    Niobium cannot be separated from zirconium or hafnium when these elements occur together in solution with common anions such as chloride and sulphate. This is ascribed to the co-polymerization of niobium(V) and the hydrolysed ionic species of zirconium(IV) and hafnium(IV) to form colloidal...

  17. Modelling of stress corrosion cracking in zirconium alloys

    International Nuclear Information System (INIS)

    Fandeur, O.; Rouillon, L.; Pilvin, P.; Jacques, P.; Rebeyrolle, V.

    2001-01-01

    During normal and incidental operating conditions, PWR power plants must comply with the first safety requirement, which is to ensure that the cladding wall is sound. Indeed some severe power transients potentially induce Stress Corrosion Cracking (SCC) of the zirconium alloy clad, due to strong Pellet Cladding Interaction (PCI). Since, at present, the prevention of this risk has some consequences on the French reactors manoeuvrability, a better understanding and forecast of the clad damage related to SCC/PCI is needed. With this aim, power ramp tests are performed in experimental reactors to assess the fuel rod behaviour and evaluate PCI failure risks. To study in detail SCC mechanisms, additional laboratory experiments are carried out on non-irradiated and irradiated cladding tubes. Numerical simulations of these tests have been developed aiming, on the one hand, to evaluate mechanical state variables and, on the other hand, to study consistent mechanical parameters for describing stress corrosion clad failure. The main result of this simulation is the determination of the validity ranges of the stress intensity factor, which is frequently used to model SCC. This parameter appears to be valid only at the onset of crack growth, when crack length remains short. In addition, the role of plastic strain rate and plastic strain as controlling parameters of the SCC process has been analysed in detail using the above mechanical description of the crack tip mechanical fields. Finally, the numerical determination of the first-order parameter(s) in the crack propagation rate law is completed by the development of laboratory tests focused on these parameters. These tests aim to support experimentally the results of the FE simulation. (author)

  18. Plasma arc melting of zirconium

    International Nuclear Information System (INIS)

    Tubesing, P.K.; Korzekwa, D.R.; Dunn, P.S.

    1997-01-01

    Zirconium, like some other refractory metals, has an undesirable sensitivity to interstitials such as oxygen. Traditionally, zirconium is processed by electron beam melting to maintain minimum interstitial contamination. Electron beam melted zirconium, however, does not respond positively to mechanical processing due to its large grain size. The authors undertook a study to determine if plasma arc melting (PAM) technology could be utilized to maintain low interstitial concentrations and improve the response of zirconium to subsequent mechanical processing. The PAM process enabled them to control and maintain low interstitial levels of oxygen and carbon, produce a more favorable grain structure, and with supplementary off-gassing, improve the response to mechanical forming

  19. Zirconium microstructures: uncharted possibilities

    International Nuclear Information System (INIS)

    Samajdar, I.; Kumar, Gulshan; Singh, Jaiveer; Lodh, Arijit; Srivastava, D.; Tewari, R.; Dey, G.K.; Saibaba, N.

    2015-01-01

    The 'conventional' Zirconium microstructures can be significantly extended with information on: (i) microtexture, (ii) residual stresses and (iii) local mechanical properties. Though these involve different tools, but a consolidated microstructure can be crated. This is the theme of this presentation. Examples of this consolidated picture will be made from deformation twinning, recovery-recrystallization, burst ductility and orientation versus solid solution hardening. (author)

  20. Zirconium elasticity modules

    International Nuclear Information System (INIS)

    Vavra, G.

    1978-01-01

    Considered are the limit and the intermediate values of the Young modulus E, modulus of shear G and of linear modulus of compression K obtainable at various temperatures (4.2 to 1133 K) for single crystals of α-zirconium. Determined and presented are the corrected isotropic elasticity characteristics of E, G, K over the above range of temperatures of textured and non-textured α-Zr

  1. 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.

  2. 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.

  3. Recovery of zirconium from pickling solution, regeneration and its reuse

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharjee, D. [Nuclear Fuel Complex, Hyderabad 500062 (India); Mandal, D., E-mail: dmandal10@gmail.com [Alkali Material & Metal Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Visweswara Rao, R.V.R.L.; Sairam, S.; Thakur, S. [Nuclear Fuel Complex, Hyderabad 500062 (India)

    2017-05-15

    Graphical abstract: The following compares the performance of fresh pickling solution (PS) and regenerated and used pickling solution (UPS). - Highlights: • Pickling of zircaloy tubes and appendages is carried out to remove oxide layer. • The pickling solution become saturated with zirconium due to reuse. • As NaNO{sub 3} concentration increases, conc. of Zr in pickling solution decreases. • Experimental results shows that, used pickling solution can be regenerated. • Regenerated solution may be reused by adding makeup quantities of HF-HNO{sub 3}. - Abstract: The pressurized heavy water reactors use natural uranium oxide (UO{sub 2}) as fuel and uses cladding material made up of zircaloy, an alloy of zirconium. Pickling of zircaloy tubes and appendages viz., spacer and bearing pads is carried out to remove the oxide layer and surface contaminants, if present. Pickling solution, after use for many cycles i.e., used pickling solution (UPS) is sold out to vendors, basically for its zirconium value. UPS, containing a relatively small concentration of hydrofluoric acid. After repeated use, pickling solution become saturated with zirconium fluoride complex and is treated by adding sodium nitrate to precipitate sodium hexafluro-zirconate. The remaining solution can be recycled after suitable makeup for further pickling use. The revenue lost by selling UPS is very high compared to its zirconium value, which causes monetary loss to the processing unit. Experiments were conducted to regenerate and reuse UPS which will save a good amount of revenue and also protect the environment. Experimental details and results are discussed in this paper.

  4. 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.

  5. Experimental creep behaviour determination of cladding tube materials under multi-axial loadings

    International Nuclear Information System (INIS)

    Grosjean, Catherine; Poquillon, Dominique; Salabura, Jean-Claude; Cloue, Jean-Marc

    2009-01-01

    Cladding tubes are structural parts of nuclear plants, submitted to complex thermomechanical loadings. Thus, it is necessary to know and predict their behaviour to preserve their integrity and to enhance their lifetime. Therefore, a new experimental device has been developed to control the load path under multi-axial load conditions. The apparatus is designed to determine the thermomechanical behaviour of zirconium alloys used for cladding tubes. First results are presented. Creep tests with different biaxial loadings were performed. Results are analysed in terms of thermal expansion and of creep strain. The anisotropy of the material is revealed and iso-creep strain curves are given.

  6. Interrelationship between structure and corrosion behaviour of zirconium

    Energy Technology Data Exchange (ETDEWEB)

    Guenther, T [Bayer A.G., Leverkusen (Germany, F.R.)

    1979-05-01

    Due to plant failures caused by the break-down of zirconium grade 702 subjected to sulphuric acid the structure and corrosion behaviour of welded and as delivered specimens were tested for various heat treatments. It was shown by structure investigations and electron microprobe analysis that the corrosion behaviour of zirconium (in boiling 65 pct sulphuric acid) is strongly infuenced by the structure, which in its turn is dependent on the grade of purity and the prehistory of the material. Type, amount, and distribution of residual elements or precipitations caused by them are responsible for the corrosion resistance. This is valid particularly for the element iron. The plant failures mentioned here coincided with the examination results. Measures to improve the chemical resistance of pure zirconium subjected to extremely aggressive media were derived.

  7. 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.

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

    Directory of Open Access Journals (Sweden)

    Ki-Nam Jang

    2017-12-01

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

  9. Annex 3 - Testing the microstructure of the fuel element with metal uranium with aluminium cladding; Prilog 3 - Ispitivanje mikrostrukture gorivnog elementa na bazi metalnog urana sa aluminijumskom kosuljicom

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, S; Momcilovic, I [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    Based on the uranium microstructure in the fuel element, i.e. grain size, presence of other phases, porosity and the position of metal connections, some data could be obtained about the mechanical and thermal treatment of the uranium during fuel element fabrication. The mentioned characteristics could indicate the behaviour of the fuel under reactor operating conditions taking into account its stable dimensions and roughness of its surfaces. Na osnovu mikrostrukture urana u gorivnom elementu, odnosno velicine i oblika njegovih kristalnih zrna, prisustva drugih faza, pozoziteta kao i rasporeda metalnih ukljucaka, mogu se dobiti izvesni podaci o tome koje je postupke mehanickog i termickog tretiranja pretrpeo uran u pripremi za gorivni elemenat. Prethodne karakteristike takodje mogu da daju predstavu o tome kako ce se gorivo ponasati u uslovima rada reaktora, s obzirom na njegovu dimenzionu stabilnost i ogrubljivanje njegove povrsine (author)

  10. Process for purifying zirconium sponge

    International Nuclear Information System (INIS)

    Abodishish, H.A.M.; Kimball, L.S.

    1992-01-01

    This patent describes a Kroll reduction process wherein a zirconium sponge contaminated with unreacted magnesium and by-product magnesium chloride is produced as a regulus, a process for purifying the zirconium sponge. It comprises: distilling magnesium and magnesium chloride from: a regulus containing a zirconium sponge and magnesium and magnesium chloride at a temperature above about 800 degrees C and at an absolute pressure less than about 10 mmHg in a distillation vessel to purify the zirconium sponge; condensing the magnesium and the magnesium chloride distilled from the zirconium sponge in a condenser; and then backfilling the vessel containing the zirconium sponge and the condenser containing the magnesium and the magnesium chloride with a gas; recirculating the gas between the vessel and the condenser to cool the zirconium sponge from above about 800 degrees C to below about 300 degrees C; and cooling the recirculating gas in the condenser containing the condensed magnesium and the condensed magnesium chloride as the gas cools the zirconium sponge to below about 300 degrees C

  11. 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.

  12. 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

  13. 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.

  14. Fluorometric determination of zirconium in minerals

    Science.gov (United States)

    Alford, W.C.; Shapiro, L.; White, C.E.

    1951-01-01

    The increasing use of zirconium in alloys and in the ceramics industry has created renewed interest in methods for its determination. It is a common constituent of many minerals, but is usually present in very small amounts. Published methods tend to be tedious, time-consuming, and uncertain as to accuracy. A new fluorometric procedure, which overcomes these objections to a large extent, is based on the blue fluorescence given by zirconium and flavonol in sulfuric acid solution. Hafnium is the only element that interferes. The sample is fused with borax glass and sodium carbonate and extracted with water. The residue is dissolved in sulfuric acid, made alkaline with sodium hydroxide to separate aluminum, and filtered. The precipitate is dissolved in sulfuric acid and electrolysed in a Melaven cell to remove iron. Flavonol is then added and the fluorescence intensity is measured with a photo-fluorometer. Analysis of seven standard mineral samples shows excellent results. The method is especially useful for minerals containing less than 0.25% zirconium oxide.

  15. Study of zirconium-addition binary systems

    International Nuclear Information System (INIS)

    Wozniakova, B.; Kuchar, L.

    1975-01-01

    The curves are given of the solid and the liquid of binary zirconium-addition systems. Most additions reduce the melting temperature of zirconium. The only known additions to increase the melting temperature are nitrogen, oxygen and hafnium. Also given are the transformation curves of the systems and the elements are given which reduce or raise the temperature of α-β transformation. From the Mendeleev table into which are plotted the curves of the solid and the liquid of binary systems it is possible to predict the properties of unknown binary systems. For the calculations of the curves of the solid and the liquid, 1860 degC was taken as the temperature of zirconium melting. For the calculations of transformation curves, 865 degC was taken as the temperature of α-β transformation. The equations are given of the curves of the solid and the liquid and of the transformation curves of some Zr-addition systems. Also given are the calculated equilibrium distribution coefficients and the equilibrium distribution coefficients of the transformation of additions in Zr and their limit values for temperatures approximating the melting point or the temperature of the transformation of pure Zr, and the values pertaining to eutectic and peritectic or eutectoid and peritectoid temperatures. (J.B.)

  16. Neutronographic Texture Analysis of Zirconium Based Alloys

    International Nuclear Information System (INIS)

    Kruz'elová, M; Vratislav, S; Kalvoda, L; Dlouhá, M

    2012-01-01

    Neutron diffraction is a very powerful tool in texture analysis of zirconium based alloys used in nuclear technique. Textures of five samples (two rolled sheets and three tubes) were investigated by using basal pole figures, inversion pole figures, and ODF distribution function. The texture measurement was performed at diffractometer KSN2 on the Laboratory of Neutron Diffraction, Department of Solid State Engineering, Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. Procedures for studying textures with thermal neutrons and procedures for obtaining texture parameters (direct and inverse pole figures, three dimensional orientation distribution function) are also described. Observed data were processed by software packages HEXAL and GSAS. Our results can be summarized as follows: i) All samples of zirconium alloys show the distribution of middle area into two maxima in basal pole figures. This is caused by alloying elements. A characteristic split of the basal pole maxima tilted from the normal direction toward the transverse direction can be observed for all samples, ii) Sheet samples prefer orientation of planes (100) and (110) perpendicular to rolling direction and orientation of planes (002) perpendicular to normal direction, iii) Basal planes of tubes are oriented parallel to tube axis, meanwhile (100) planes are oriented perpendicular to tube axis. Level of resulting texture and maxima position is different for tubes and for sheets. The obtained results are characteristic for zirconium based alloys.

  17. Nuclear fuel element

    International Nuclear Information System (INIS)

    Mogard, J.H.

    1977-01-01

    A nuclear fuel element is disclosed for use in power producing nuclear reactors, comprising a plurality of axially aligned ceramic cylindrical fuel bodies of the sintered type, and a cladding tube of metal or metal alloys, wherein said cladding tube on its cylindrical inner surface is provided with a plurality of slightly protruding spacing elements distributed over said inner surface

  18. Tasks related to increase of RA reactor exploitation and experimental potential, 02. Verification of the system for detecting failures of the RA reactor fuel element cladding; Radovi na povecanju eksploatacionih i eksperimentalnih mogucnosti reaktora RA, 02. Provera sistema za detekciju pucanja kosuljice gorivnog elementa reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    For the purpose of this task it was necessary to analyze the time dependent distribution of fission products in the fuel element and leaks through the cracks in the cladding; to calculate the quantity of solid fission products, volatile fission products and fission gases for the RA reactor; to collect the data for estimating the activity of the short-living isotopes created by neutron irradiation of D{sub 2}O; analyze the number of delayed neutrons in D{sub 2}O. Experiment were needed to estimate the distribution of Xe and Kr in the heavy water in the reactor channel and analyze the activity of D{sub 2}O and helium based on reactor operation data. For the purpose of verifying the efficiency and safety of the existing system for detecting the cracks of the fuel element cladding is presented in this report together with the review of similar systems at a number of reactors in the world.

  19. ZIRCONIUM PHOSPHATE ADSORPTION METHOD

    Science.gov (United States)

    Russell, E.R.; Adamson, A.S.; Schubert, J.; Boyd, G.E.

    1958-11-01

    A method is presented for separating plutonium values from fission product values in aqueous acidic solution. This is accomplished by flowing the solutlon containing such values through a bed of zirconium orthophosphate. Any fission products adsorbed can subsequently be eluted by washing the column with a solution of 2N HNO/sub 3/ and O.lN H/sub 3/PO/sub 4/. Plutonium values may subsequently be desorbed by contacting the column with a solution of 7N HNO/sub 3/ .

  20. Laves intermetallics in stainless steel-zirconium alloys

    International Nuclear Information System (INIS)

    Abraham, D.P.; McDeavitt, S.M.; Richardson, J.W. Jr.

    1997-01-01

    Laves intermetallics have a significant effect on properties of metal waste forms being developed at Argonne National Laboratory. These waste forms are stainless steel-zirconium alloys that will contain radioactive metal isotopes isolated from spent nuclear fuel by electrometallurgical treatment. The baseline waste form composition for stainless steel-clad fuels is stainless steel-15 wt.% zirconium (SS-15Zr). This article presents results of neutron diffraction measurements, heat-treatment studies and mechanical testing on SS-15Zr alloys. The Laves intermetallics in these alloys, labeled Zr(Fe,Cr,Ni) 2+x , have both C36 and C15 crystal structures. A fraction of these intermetallics transform into (Fe,Cr,Ni) 23 Zr 6 during high-temperature annealing; the authors have proposed a mechanism for this transformation. The SS-15Zr alloys show virtually no elongation in uniaxial tension, but exhibit good strength and ductility in compression tests. This article also presents neutron diffraction and microstructural data for a stainless steel-42 wt.% zirconium (SS-42Zr) alloy

  1. Newly developed active braze powders based on commercial nickel brazes using zirconium as active element for joining ceramic to metal; Entwicklung von neuen Aktivlotpulvern auf Basis kommerzieller Nickellote mit Zirkon als aktivelement zum Fuegen von Keramik-Metall-Verbunden

    Energy Technology Data Exchange (ETDEWEB)

    Bobzin, K.; Schlaefer, T.; Kopp, N.; Schlegel, A. [Institut fuer Oberflaechentechnik der RWTH Aachen (Germany)

    2010-06-15

    The increased requirements of highly stressed components, concerning the resistance to thermal-induced stresses, oxidation, corrosion, hardness as well as wear resistance make high-performance technical ceramics ideally suited for such applications. On the other hand they exhibit properties like high brittleness, partly low thermal shock resistance, low workability and consequential limitations in the engineering design. Hybrid material concepts, as combination of high-performance technical ceramics and metallic engineering materials, can offer interesting technological solutions, if suitable and joining technologies are available. Active brazing, which is a very flexible joining technology in respect of the material selection, arises for the development of new and innovative applications, such as high-temperature fuel cells. Currently silver/copper, copper and silver active brazing filler metals are already used in the industry and are characterised by a decrease of their mechanical strength at approx. 500 C. Referring to this, gold and palladium active brazing filler metals show better features, but because of their high price, they are seldom used. The aim of the reported investigations is the development of active brazing filler metals with reasonable raw materials costs for working temperatures above 500 C and moreover to be used in hydrocarbonated environments with better corrosion-resistance than silver/copper, copper and silver active brazing filler metals. Experimental brazing filler metals with zirconium as surface-active element has been manufactured on the basis of nickel brazing filler metals NI 102, NI 105 and NI 107. The modification of each nickel brazing filler metal was carried out on the one hand by powder metallurgy, whereby zirconium hydride has been mixed or mechanically alloyed. On the other hand the nickel brazing filler metals have been alloyed with zirconium by melting metallurgy. The content of active metal varied between 2 weight-% and

  2. Evaluation of a Ductility after High Temperature Oxidation with the Three-Point Bend Test in Zirconium Alloys

    International Nuclear Information System (INIS)

    Jung, Yang Il; Park, Sang Yoon; Park, Jeong Yong; Jeong, Yong Hwan

    2010-01-01

    In a light water reactor, the fuel cladding play an important role of preventing leakage of radioactive materials into the coolant, and thus the mechanical integrity of the cladding should be guaranteed under the conditions of normal and transient operation. In the case of a loss of coolant accident (LOCA), the cladding is subjected to a high temperature oxidation which is finally quenched because of an emergency coolant reflooding into the core. In this situation, the current LOCA criteria consist of five separate requirements: i) peak cladding temperature, ii) maximum cladding oxidation, iii) maximum hydrogen generation, iv) coolable geometry, and v) long-term cooling. The claddings lose their ductility due to the microstructural phase transformation from beta to martensite alpha-prime. and hydrogen up-take after LOCA. Since the reduction in ductility can induce embrittlement of claddings, post-quench ductility is one of the major concerns in transient operation circumstances. For the analysis, usually ring compression test are performed on ring samples cut from the tube to examine the oxidized cladding ductility. However, the test would not be applicable to the platelet samples which are general form of a specimen for developing alloys. As a high burn-up fuel cladding materials, Zircaloys are being replaced by modern zirconium alloys such as ZIRLO, and M5. Korea has also developed a new fuel cladding material HANA (high performance alloy for nuclear application) by the Korea Atomic Energy Research Institute. Because of the different composition of the newer claddings in comparison with the conventional Zircaloy-4, the high temperature oxidation behavior and the ductility after the oxidation would be different, and the properties should be evaluated how much the newer claddings were improved

  3. 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.

  4. 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)

  5. 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

  6. Plasma spheroidization and cladding of powders

    Energy Technology Data Exchange (ETDEWEB)

    Petrunichev, V.A.; Averin, V.V.; Sorokin, L.M.; Koroleva, E.B.

    1987-02-01

    With reference to experimental results for nickel and chromium alloys, it is shown that complex alloy powders can be spheroidized in plasma discharges using an argon plasma with hydrogen. The spheroidizing process is accompanied by the reduction of surface oxides, with uniform element distribution within the particles; the granulometric composition of the particles is preserved. It is also shown that plasma technology can be used for producing metal-clad oxide and carbide powders, which improve the performance of cermets and coatings.

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

    Science.gov (United States)

    Rebak, Raul B.

    2017-12-01

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

  8. Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

    Science.gov (United States)

    Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

    2018-02-01

    The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

  9. Stainless steel-zirconium alloy waste forms

    International Nuclear Information System (INIS)

    McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.

    1996-01-01

    An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ''noble'' nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation

  10. Modification in band gap of zirconium complexes

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Mayank, E-mail: mayank30134@gmail.com; Singh, J.; Chouhan, S. [Department of Physics, ISLE, IPS Academy, Indore (M.P.) (India); Mishra, A. [School of Physics, Devi Ahilya Vishwavidyalaya, Indore (M.P.) (India); Shrivastava, B. D. [Govt. P. G. College, Biora (M.P.) (India)

    2016-05-06

    The optical properties of zirconium complexes with amino acid based Schiff bases are reported here. The zirconium complexes show interesting stereo chemical features, which are applicable in organometallic and organic synthesis as well as in catalysis. The band gaps of both Schiff bases and zirconium complexes were obtained by UV-Visible spectroscopy. It was found that the band gap of zirconium complexes has been modified after adding zirconium compound to the Schiff bases.

  11. 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

  12. Zirconium for nitric acid solutions

    International Nuclear Information System (INIS)

    Yau, T.L.

    1984-01-01

    The excellent corrosion resistance of zirconium in nitric acid has been known for over 30 years. Recently, there is an increasing interest in using zirconium for nitric acid services. Therefore, an extensive research effort has been carried out to achieve a better understanding of the corrosion properties of zirconium in nitric acid. Particular attention is paid to the effect of concentration, temperature, structure, solution impurities, and stress. Immersion, autoclave, U-bend, and constant strain-rate tests were used in this study. Results of this study indicate that the corrosion resistance of zirconium in nitric acid is little affected by changes in temperature and concentration, and the presence of common impurities such as seawater, sodium chloride, ferric chloride, iron, and stainless steel. Moreover, the presence of seawater, sodium chloride, ferric chloride, and stainless steel has little effect on the stress corrosion craking (SCC) susceptibility of zirconium in 70% nitric acid at room temperatures. However, zirconium could be attacked by fluoride-containing nitric acid and the vapors of chloride-containing nitric acid. Also, high sustained tensile stresses should be avoided when zirconium is used to handle 70% nitric acid at elevated temperatures or > 70% nitric acid

  13. 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.

  14. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  15. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  16. Characterization of zirconium alloy oxidation films by alternating current impedance

    International Nuclear Information System (INIS)

    Rosecrans, P.M.

    1984-01-01

    Kinetics of zirconium alloy oxidation are highly nonlinear. The results of electrochemical measurements and electron microscopy support the existence of porosity in oxide films formed on zirconium alloys in high temperature aqueous environments. Analytical treatment is presented relating oxidation kinetics to the thickness and distribution of nonporous elements within the oxide. This analysis illustrates that both the level and distribution of porosity within the oxide factor into oxidation kinetics. The barrier layer model can provide a basis for predicting the effect of environmental changes on oxidation rate. In addition, it demonstrates the need for further research into porosity generation mechanisms in oxide films

  17. Recrystallization resistance in aluminum alloys containing zirconium

    International Nuclear Information System (INIS)

    Ranganathan, K.

    1991-01-01

    Zirconium forms a fine dispersion of the metastable β' (Al 3 Zr) phase that controls recrystallization by retarding the motion of high-angle boundaries. The primary material chosen for this research was aluminum alloy 7150 containing zinc, magnesium, and copper as the major solute elements and zirconium as the dispersoid-forming element. The size, distribution, and the volume fraction of β' was controlled by varying the alloy composition and preheat practices. Preheated ingots were subjected to a specific sequence of hot-rolling operations to evaluate the resistance to recrystallization of the different microstructures. Optical and transmission electron microscopy (TEM) techniques were used to investigate the influence of dispersoid morphology resulting from the thermal treatments and deformation processing on the recrystallization behavior of the alloy. Studies were conducted to determine the influence of the individual solute elements present in 7150 on the precipitation of β' and consequently on the recrystallization behavior of the material. These studies were done on compositional variants of commercial 7150

  18. A half-century of changes in zirconium alloys

    International Nuclear Information System (INIS)

    Mardon, J.P.; Barberis, P.; Hoffmann, P.B.

    2008-01-01

    This article presents the history of zirconium alloys for PWR and BWR technologies. For more than 20 years zirconium alloys have evolved to cope with demands of the reactor operators concerning the burn-up extension and new safety margins. The poor properties of Zircaloy-1 concerning corrosion have led researchers to add elements like iron by developing Zircaloy-3A and Zircaloy-3C, and resulting in Zircaloy-4 with tin addition (from 1.30% to 1.50%). Zircaloy-4 is now outdated for PWR and new zirconium alloys with niobium are used (M5, ZIRLO...) they present a better resistance to corrosion, to hydridation, to creep and they are less prone to dimensional changes under irradiation. (A.C.)

  19. 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.

  20. Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

    International Nuclear Information System (INIS)

    Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

    2008-01-01

    As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

  1. SEPARATION OF HAFNIUM FROM ZIRCONIUM

    Science.gov (United States)

    Overholser, L.B.; Barton, C.J. Sr.; Ramsey, J.W.

    1960-05-31

    The separation of hafnium impurities from zirconium can be accomplished by means of organic solvent extraction. The hafnium-containing zirconium feed material is dissolved in an aqueous chloride solution and the resulting solution is contacted with an organic hexone phase, with at least one of the phases containing thiocyanate. The hafnium is extracted into the organic phase while zirconium remains in the aqueous phase. Further recovery of zirconium is effected by stripping the onganic phase with a hydrochloric acid solution and commingling the resulting strip solution with the aqueous feed solution. Hexone is recovered and recycled by means of scrubbing the onganic phase with a sulfuric acid solution to remove the hafnium, and thiocyanate is recovered and recycled by means of neutralizing the effluent streams to obtain ammonium thiocyanate.

  2. Zirconium nitride hard coatings

    International Nuclear Information System (INIS)

    Roman, Daiane; Amorim, Cintia Lugnani Gomes de; Soares, Gabriel Vieira; Figueroa, Carlos Alejandro; Baumvol, Israel Jacob Rabin; Basso, Rodrigo Leonardo de Oliveira

    2010-01-01

    Zirconium nitride (ZrN) nanometric films were deposited onto different substrates, in order to study the surface crystalline microstructure and also to investigate the electrochemical behavior to obtain a better composition that minimizes corrosion reactions. The coatings were produced by physical vapor deposition (PVD). The influence of the nitrogen partial pressure, deposition time and temperature over the surface properties was studied. Rutherford backscattering spectrometry (RBS), X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD), scanning electron microscopy (SEM) and corrosion experiments were performed to characterize the ZrN hard coatings. The ZrN films properties and microstructure changes according to the deposition parameters. The corrosion resistance increases with temperature used in the films deposition. Corrosion tests show that ZrN coating deposited by PVD onto titanium substrate can improve the corrosion resistance. (author)

  3. Zirconium-based alloys, nuclear fuel rods and nuclear reactors including such alloys, and related methods

    Science.gov (United States)

    Mariani, Robert Dominick

    2014-09-09

    Zirconium-based metal alloy compositions comprise zirconium, a first additive in which the permeability of hydrogen decreases with increasing temperatures at least over a temperature range extending from 350.degree. C. to 750.degree. C., and a second additive having a solubility in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. At least one of a solubility of the first additive in the second additive over the temperature range extending from 350.degree. C. to 750.degree. C. and a solubility of the second additive in the first additive over the temperature range extending from 350.degree. C. to 750.degree. C. is higher than the solubility of the second additive in zirconium over the temperature range extending from 350.degree. C. to 750.degree. C. Nuclear fuel rods include a cladding material comprising such metal alloy compositions, and nuclear reactors include such fuel rods. Methods are used to fabricate such zirconium-based metal alloy compositions.

  4. Corrosion performance of optimised and advanced fuel rod cladding in PWRs at high burnups

    International Nuclear Information System (INIS)

    Jourdain, P.; Hallstadius, L.; Pati, S.R.; Smith, G.P.; Garde, A.M.

    1997-01-01

    The corrosion behaviour both in-pile and out-of-pile for a number of cladding alloys developed by ABB to meet the current and future needs for fuel rod cladding with improved corrosion resistance is presented. The cladding materials include: 1) Zircaloy-4 (OPTIN) with optimised composition and processing and Zircaloy-2 optimised for Pressurised Water Reactors (PWR), (Zircaloy-2P), and 2) several alternative zirconium-based alloys with compositions outside the composition range for Zircaloys. The data presented originate from fuel rods irradiated in six PWRs to burnups up to about 66 MWd/kgU and from tests conducted in 360 o water autoclave. Also included are in-pile fuel rod growth measurements on some of the alloys. (UK)

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

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [ORNL

    2014-03-28

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

  6. 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

  7. Activation analysis in zirconium and alloys for nuclear application

    International Nuclear Information System (INIS)

    Cohen, I.M.; Mila, M.I.; Gomez, C.D.

    1981-01-01

    A study has been performed with the purpose to ascertain the possibilities of using neutron activation analysis in non-destructive determination of several elements present in zirconium and its alloys. Those elements must be limited within acceptable top levels, in accordance to standards for nuclear applications. The experimental techniques used are described and the results obtained are discussed, showing that the method is adequate for determining Cl, Co, Hf, Mn, and W, but not Ni and U. (M.E.L.) [es

  8. Corrosion mechanisms of zirconium alloys - study of the initial oxidation kinetics and of the mechanical behaviour of the metal/oxide system

    International Nuclear Information System (INIS)

    Parise, M.

    1996-12-01

    Nuclear fuel claddings are made of zirconium alloys. The conditions of use lead the cladding oxidize outside. The so-formed layers behaves like a thermal barrier and prevents from using oxidized claddings with an oxide thickness larger than 100 μm. The oxidation kinetic is approximately cubic for oxide thicknesses smaller than about 2μm, linear beyond. A kinetic model has been proposed which estimates the post-transition growth rate from the kinetic parameters of the pre-transition state and morphological features of post-transition layers. This work aims at providing the necessary elements to validate this model and studying the layers around the kinetic transition, in order to determine whether the oxidation mechanisms before and after the transition are similar. Thicknesses of the 50 - 500 nm range of the oxide layers are measured by an optical method; pre-transition kinetics are thus precisely determined. The effect of the composition, the thermal treatment and the presence of oxygen in solid solution is studied. The morphological and crystallographic study of the layers show that they exhibit a lot of similarities before and after the kinetic transition. The results concerning the kinetic aspects and the morphology of the post-transition layers point out that the proposed model leads to realistic post-transition growth rates. Furthermore, the kinetic transition corresponds to the appearance of cracks in the oxide layer. The mechanical behaviour of the metal/oxide system has been modelled at different scales. When the specific behaviours of the metal and the oxide are taken into account together with the interface geometry, radial stresses appear, which are high enough to locally open cracks. The appearance and localization of cracks depend on both the interface geometry and the stress distribution in the metal/oxide system. (author)

  9. Irradiation effects of the zirconium oxidation and the uranium diffusion in zirconia; Effets d'irradiation sur l'oxydation du zirconium et la diffusion de l'uranium dans la zircone

    Energy Technology Data Exchange (ETDEWEB)

    Bererd, N

    2003-07-01

    The context of this study is the direct storage of spent fuel assemblies after operation in reactor. In order to obtain data on the capacities of the can as the uranium diffusion barrier, a fundamental study has been carried out for modelling the internal cladding surface under and without irradiation. The behaviour of zirconium in reactor conditions has at first been studied. A thin uranium target enriched with fissile isotope has been put on a zirconium sample, the set being irradiated by a thermal neutrons flux leading to the fission of the deposited uranium. The energetic history of the formed fission products has revealed two steps: 1)the zirconium oxidation and 2)the diffusion of uranium in the zirconia formed at 480 degrees C. A diffusion coefficient under irradiation has been measured. Its value is 10{sup -15} cm{sup 2}.s{sup -1}. In order to be able to reveal clearly the effect of the irradiation by the fission products on the zirconium oxidation, measurements of thermal oxidation and under {sup 129}Xe irradiation have been carried out. They have shown that the oxidation is strongly accelerated by the irradiation and that the temperature is negligible until 480 degrees C. On the other hand, the thermal diffusion of the uranium in zirconium and in zirconia has been studied by coupling ion implantation and Rutherford backscattering spectroscopy. This study shows that the uranium diffuses in zirconium and is trapped in zirconia in a UO{sub 3} shape. (O.M.)

  10. 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

  11. 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.

  12. Precipitation of γ-zirconium hydride in zirconium

    International Nuclear Information System (INIS)

    Carpenter, G.J.C.

    1978-01-01

    A mechanism for the precipitation of γ-zirconium hydride in zirconium is presented which does not require the diffusion of zirconium. The transformation is completed by shears caused by 1/3 (10 anti 10) Shockley partial dislocations on alternate zirconium basal planes, either by homogeneous nucleation or at lattice imperfections. Homogeneous nucleation is considered least likely in view of the large nucleation barrier involved. Hydrides may form at dislocations by the generation of partials by means of either a pole or ratchet mechanism. The former requires dislocations with a component of Burgers vector along the c-axis, but contrast experiments show that these are not normally observed in annealed zirconium. It is therefore most likely that intragranular hydrides form at the regular 1/3 (11 anti 20) dislocations, possibly by means of a ratchet mechanism. Contrast experiments in the electron microscope show that the precipitates have a shear character consistent with the mechanism suggested. The possibility that the shear dislocations associated with the hydrides are emissary dislocations is considered and a model suggested in which this function is satisfied together with the partial relief of misfit stresses. The large shear strains associated with the precipitation mechanism may play an important role in the preferential orientation of hydrides under stress

  13. Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Korkut, Turgay; Kara, Ayhan; Korkut, Hatun [Sinop Univ. (Turkey). Dept. of Nuclear Energy Engineering

    2016-12-15

    Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.

  14. Fine-grained zirconium-base material

    Science.gov (United States)

    Van Houten, G.R.

    1974-01-01

    A method is described for making zirconium with inhibited grain growth characteristics, by the process of vacuum melting the zirconium, adding 0.3 to 0.5% carbon, stirring, homogenizing, and cooling. (Official Gazette)

  15. 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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  17. On elastic structural elements for nuclear reactors

    International Nuclear Information System (INIS)

    Povolo, F.

    1978-03-01

    The in-pile stress-relaxation behaviour of materials usually employed for the elastic structural elements, in nuclear reactors, is critically reviewed and the results are compared with those obtained in commercial zirconium alloys irradiated under similar conditions. Finally, it is shown that, under certain conditions, some zirconium alloys may be used as an alternative material for these structural elements. (orig.) [de

  18. Problems of zirconium metal production in Czechoslovakia

    International Nuclear Information System (INIS)

    Vareka, J.; Vaclavik, E.

    1975-01-01

    The problems are summed up of the production and quality control of zirconium sponge. A survey is given of industrial applications of zirconium in form of pure metal or alloys in nuclear power production, ferrous and non-ferrous metallurgy, chemical engineering and electrical engineering. A survey is also presented of the manufacture of zirconium metal in advanced capitalist countries. (J.B.)

  19. 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)

  20. Atomistic modeling of zirconium hydride precipitation: methodology for deriving a tight-binding potential

    International Nuclear Information System (INIS)

    Dufresne, Alice

    2014-01-01

    The zirconium-hydrogen system is of nuclear safety interest, as the hydride precipitation leads to the cladding embrittlement, which is made of zirconium-based alloys. The cladding is the first safety barrier confining the radioactive products: its integrity shall be kept during the entire fuel-assemblies life, in reactor, including accidental situation, and post-operation (transport and storage). Many uncertainties remain regarding the hydrides precipitation kinetics and the local stress impact on their precipitation. The atomic scale modeling of this system would bring clarifications on the relevant mechanisms. The usual atomistic modeling methods are based on thermo-statistic approaches, whose precision and reliability depend on the interatomic potential used. However, there was no potential allowing a rigorous study of the Zr-H system. The present work has indeed addressed this issue: a new tight-binding potential for zirconium hydrides modeling is now available. Moreover, this thesis provides a detailed manual for deriving such potentials accounting for spd hybridization, and fitted here on DFT results. This guidebook has be written in light of modeling a pure transition metal followed by a metal-covalent coupling (metallic carbides, nitrides and silicides). (author)

  1. Waterside corrosion of zirconium alloys in nuclear power plants

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan; Baek, B. J.; Park, S. Y. and others

    1999-08-01

    The overview of corrosion and hydriding behaviors of Zr-based alloy under the conditions of the in-reactor service and in the absence of irradiation is introduced in this report. The metallurgical characteristics of Zr-based alloys and the thermo-mechanical treatments on the microstructures and the textures in the manufacturing process for fuel cladding are also introduced. The factors affecting the corrosion of Zr alloy in reactor are summarized. And the corrosion mechanism and hydrogen up-take are discussed based on the laboratory and in-reactor results. The phenomenological observations of zirconium alloy corrosion in reactors are summarized and the models of in-reactor corrosion are exclusively discussed. Finally, the effects of irradiation on the corrosion process in Zr alloy were investigated mainly based on the literature data. (author). 538 refs., 26 tabs., 105 figs

  2. Investigation of chlorination of zirconium and hafnium and their compounds in discharge from hollow cathode

    International Nuclear Information System (INIS)

    Ioffe, R.B.; Korovin, Yu.I.

    1978-01-01

    The possibility is investigated of chlorinating various zirconium and hafnium compounds (metal, oxide, carbide) in a hot discharge from a hollow cathode with various chlorinating reagents: copper monochloride, nickel chloride, magnesium chloride, for the purpose of accelerating their entrance into the excitation zone. It has been shown thermodynamically and experimentally that chlorination of metal zirconium and hafnium and their carbides with copper monochloride in hot hollow cathode conditions provides a sharp increase in the intensity of the lines of these elements

  3. Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

    International Nuclear Information System (INIS)

    Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

    1995-01-01

    The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

  4. PROCESS OF DISSOLVING ZIRCONIUM ALLOYS

    Science.gov (United States)

    Shor, R.S.; Vogler, S.

    1958-01-21

    A process is described for dissolving binary zirconium-uranium alloys where the uranium content is about 2%. In prior dissolution procedures for these alloys, an oxidizing agent was added to prevent the precipitation of uranium tetrafluoride. In the present method complete dissolution is accomplished without the use of the oxidizing agent by using only the stoichiometric amount or slight excess of HF required by the zirconium. The concentration of the acid may range from 2M to 10M and the dissolution is advatageously carried out at a temperature of 80 deg C.

  5. 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)

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

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi

    2001-03-01

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

  7. Environmentally-induced cracking of zirconium alloys - a review

    International Nuclear Information System (INIS)

    Cox, B.

    1990-01-01

    The general field of environmentally-induced cracking of zirconium alloys has been reviewed and the phenomena that are observed and the progress in understanding the mechanisms are summarized. The details of the industrially important pellet-clad interaction failures of nuclear reactor fuel have been left for a companion review, and only observations on the mechanism are summarized briefly here. It is concluded that in the zirconium alloy system, by virtue of the physical peculiarities of the system, it is easier to reach unambiguous conclusions about the environmental cracking mechanisms that are operating than with other systems. Thus, chemical dissolution in either liquid or vapour phase is thought to be the principal mechanism for intergranular cracking, while adsorption-induced embrittlement is thought to be the most common transgranular quasi-cleavage process. Hydrogen embrittlement in this system can be identified because it requires precipitated hydride that gives characteristic fractography when cracked. Only in a few instances does stress-corrosion cracking appear to proceed by a hydride cracking mechanism. (orig.)

  8. Process for etching zirconium metallic objects

    International Nuclear Information System (INIS)

    Panson, A.J.

    1988-01-01

    In a process for etching of zirconium metallic articles formed from zirconium or a zirconium alloy, wherein the zirconium metallic article is contacted with an aqueous hydrofluoric acid-nitric acid etching bath having an initial ratio of hydrofluoric acid to nitric acid and an initial concentration of hydrofluoric and nitric acids, the improvement, is described comprising: after etching of zirconium metallic articles in the bath for a period of time such that the etching rate has diminished from an initial rate to a lesser rate, adding hydrofluoric acid and nitric acid to the exhausted bath to adjust the concentration and ratio of hydrofluoric acid to nitric acid therein to a value substantially that of the initial concentration and ratio and thereby regenerate the etching solution without removal of dissolved zirconium therefrom; and etching further zirconium metallic articles in the regenerated etching bath

  9. Influence of Zircaloy cladding composition on hydride formation during aqueous hydrogen charging

    Energy Technology Data Exchange (ETDEWEB)

    Rajasekhara, S. [Intel Corporation, 2501 NW 229th Av., Hillsboro, OR 97124 (United States); Kotula, P.G.; Enos, D.G.; Doyle, B.L. [Sandia National Laboratories, Albuquerque, NM, 87185 (United States); Clark, B.G., E-mail: blyclar@sandia.gov [Sandia National Laboratories, Albuquerque, NM, 87185 (United States)

    2017-06-15

    Although hydrogen uptake in Zirconium (Zr) based claddings has been a topic of many studies, hydrogen uptake as a function of alloy composition has received little attention. In this work, commercial Zr-based cladding alloys (Zircaloy-2, Zircaloy-4 and ZIRLO™), differing in composition but with similar initial textures, grain sizes, and surface roughness, were aqueously charged with hydrogen for 100, 300, and 1000 s at nominally 90 °C to produce hydride layers of varying thicknesses. Transmission electron microscope characterization following aqueous charging showed hydride phase and orientation relationship were identical in all three alloys. However, elastic recoil detection measurements confirmed that surface hydride layers in Zircaloy-2 and Zircaloy-4 were an order of magnitude thicker relative to ZIRLO™. - Highlights: •Aqueous charging was performed to produce a layer of zirconium hydride for three different Zr-alloy claddings. •Hydride thicknesses were analyzed by elastic recoil detection and transmission electron microscopy. •Zircaloy-2 and Zircaloy-4 formed thicker hydride layers than ZIRLO™ for the same charging durations.

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

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed. - Highlights: •Complementary fission gas release events are reported for U-Mo fuel with and without cladding. •Exothermic reaction between Zr diffusion layer and cladding influences fission gas release. •Mechanisms responsible for fission gas release are similar, but with varying timing and magnitude. •Behavior of samples is similar after 800 °C signaling the onset of superlattice destabilization.

  11. Influence of processing variables and alloy chemistry on the corrosion behavior of ZIRLO nuclear fuel cladding

    International Nuclear Information System (INIS)

    Comstock, R.J.; Sabol, G.P.; Schoenberger, G.

    1996-01-01

    Variations in the thermal heat treatments used during the fabrication of ZIRLO (Zr-1Nb-1Sn-0.1Fe) fuel clad tubing and in ZIRLO alloy chemistry were explored to develop a further understanding of the relationship between processing, microstructure, and cladding corrosion performance. Heat treatment variables included intermediate tube annealing temperatures as well as a beta-phase heat treatment during the latter stages of the tube reduction schedule. Chemistry variables included deviations in niobium and tin content from the nominal composition. The effects of both heat treatment and chemistry on corrosion behavior were assessed by autoclave tests in both pure and lithiated water and high-temperature steam. Analytical electron microscopy demonstrated that the best out-reactor corrosion performance is obtained for microstructures containing a fine distribution of beta-niobium and Zr-Nb-Fe particles. Deviations from this microstructure, such as the presence of beta-zirconium phase, tend to degrade corrosion resistance. ZIRLO fuel cladding was irradiated in four commercial reactors. In all cases, the microstructure in the cladding included beta-niobium and Zr-Nb-Fe particles. ZIRLO fuel cladding processed with a late-stage beta heat treatment to further refine the second-phase particle size exhibited in-reactor corrosion behavior that was similar to reference ZIRLO cladding. Variations of the in-reactor corrosion behavior of ZIRLO were correlated to tin content, with higher oxide thickness observed in the ZIRLO cladding containing higher tin. The results of these studies indicate that optimum corrosion performance of ZIRLO is achieved by maintaining a uniform distribution of fine second-phase particles and controlled levels of tin

  12. Performance of U-Pu-Zr fuel cast into zirconium molds

    International Nuclear Information System (INIS)

    Crawford, D.C.; Lahm, C.E.; Tsai, H.

    1992-01-01

    Current fabrication techniques for the integral fast reactor (IFR) fuel utilize injection casting into quartz molds after reprocessing in the IFR fuel cycle facility. The quartz molds are destroyed during the fuel demolding process, and the quartz residue must therefore be treated as contaminated waste. Alternatively, if the fuel can be cast into molds that remain as part of the fuel slugs (i.e., if the fuel can be left inside the molds for irradiation), then the quartz mold contribution to the waste stream can be eliminated. This possibility is being addresssed in an ongoing effort to evaluate the irradiation performance of fuel cast into zirconium sheaths rather than quartz molds. Zirconium was chosen as the sheath material because it is the component of the U-Pu-Zr fuel alloy that raises the alloy solidus temperatures and provides resistance to fuel-cladding chemical interaction (FCCI)

  13. New zirconium alloys for nuclear application; Novas ligas de zirconio para aplicacao nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Lobo, R.M.; Andrade, A.H.P., E-mail: rmlobo@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2010-07-01

    Zirconium alloys are widely used in the nuclear industry, mainly in fuel cladding tubes and structural components for PWR plants. The service life of these components, which operate under high temperatures conditions ({approx} 300 deg C), has led to developing new alloys with the aim to improve the mechanical properties, corrosion resistance and irradiation damage. The variation in the composition of the alloy produces second phase particles which alter the materials properties according to their size and distribution, is essential therefore, knowledge their characteristics. Analysis of second phase particles in zirconium alloys are carried out by scanning electron microscopy, transmission electron microscopy and image analysis. This study used the zircaloy-4 to illustrate the characterization of these alloys through the study of second phase particles. (author)

  14. Atomistic studies of cation transport in tetragonal ZrO2 during zirconium corrosion

    International Nuclear Information System (INIS)

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

    2015-01-01

    Zirconium alloys are the major fuel cladding materials in current reactors. The water-side corrosion is a significant degradation mechanism of these alloys. During corrosion, the transport of oxidizing species in zirconium dioxide (ZrO 2 ) determines the corrosion kinetics. Previously, it has been argued that the outward diffusion of cations is important for forming protective oxides. In this work, the migration of Zr defects in tetragonal ZrO 2 is studied with temperature accelerated dynamics and molecular dynamics simulations. The results show that Zr interstitials have anisotropic diffusion and migrate preferentially along the [001] or c direction in tetragonal ZrO 2 . The compressive stresses can increase the Zr interstitial migration barrier significantly. The migration of Zr interstitials at a grain boundary is much slower than in a bulk oxide. The implications of these atomistic simulation results in the Zr corrosion are discussed. (authors)

  15. 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)

  16. Rapid screening of nuclear grade zirconium silicate without separation of hafnium from the bulk matrix

    International Nuclear Information System (INIS)

    Venkatesh, Manisha; Sharma, P.K.; Avhad, D.K.; Basu, H.; Singhal, R.K.; Reddy, A.V.R.

    2014-01-01

    Zirconium silicate, also zirconium orthosilicate, (ZrSiO 4 ) is a chemical compound, and occurs in nature as zircon, a silicate mineral. The concentration of Hafnium in nuclear grade Zirconium must be less than 0.2% w/w of Zr. In view of this it must be accurately chemically characterized before issuing a certification for export under non nuclear category. As the chemistry of Zr and Hf is similar, it is difficult to separate Hf by direct wet chemical method. During this work, concentration of Hf in zirconium silicate was measured by Field Portable X-ray Fluorescence (FPXRF) and results obtained were validated by using detailed chemical method. FPXRF spectrometry has become a common analytical technique for on-site screening and fast turnaround analysis of contaminant elements in environmental samples

  17. Experimental study of the zirconium alloy oxidation under high pressure of steam and modelling of the mechanisms

    International Nuclear Information System (INIS)

    Dali, Yacoub

    2007-01-01

    The corrosion of the cladding materials used for the fuel rods is one of the limiting factor of their lifetime in light water reactors. In this field, the aim of the nuclear industry is today to increase the time and the number of cycles and to submit the claddings in zirconium alloys to higher corrosive conditions. In this way, new alloys devoted to replace the standard Zircaloy-4, for instance Nb containing alloys, have been recently developed and licensed and show better corrosion resistance. A better understanding of the corrosion mechanisms of the zirconium alloys is necessary to predict the corrosion behaviour of these materials. In this work, the oxidation rate of model alloys of two metallurgic families has been studied in steam in a pressure range between 100 milli-bars and 100 bars. The Zircaloy type alloys contain as alloying elements oxygen and/or tin and/or iron and chromium. For the Zr-Nb family, three niobium contents have been studied, respectively 0.2, 0.4 and 1 weight percent of niobium. Our objectives were to understand the variations of the reactivity between the low pressure and the high pressure range, in quantifying the dependency of the corrosion rate with the steam pressure and the alloying element concentrations. The segregation process of the niobium at the surface has also been studied on the Zr-Nb alloys. During this work, a magnetic suspension thermo-balance has been developed and used to follow in-situ the corrosion rate at high pressure of water vapour. The oxide layers have been characterized by many techniques, macro and micro-photo-electrochemistry, XRD, FEG-SEM, XPS, HR-TEM and SIMS. For the Zircaloy type alloys, we have confirmed the major role of the intermetallic precipitates Zr(Fe,Cr) 2 on the corrosion resistance. Unlike the standard Zircaloy-4, for which the oxidation rate does not depend on the pressure of the water vapour and is thus limited by the vacancy diffusion in the oxide layer, we have shown that the rate of the

  18. Solute redistribution studies in oxidised zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Khera, S K; Kale, G B; Gadiyar, H S [Bhabha Atomic Research Centre, Bombay (India). Metallurgy Div.

    1977-01-01

    Electron microprobe studies on solute distribution in oxide layers and in the regions near oxide metal interface have been carried out in the case of zircaloy-2 and zirconium binary alloys containing niobium, tin, iron, copper, chromium and nickel and oxidised in steam at 550 deg C. In the case of alloys having higher oxidation rates, the oxide of solute element was found to dissolve in ZrO/sub 2/ without any composition variation. However, for solute addition with limited solubility like Cr, Cu and Fe, solute enrichment at metal/oxide interface and depletion of the same matrix has been observed. The intensity profiles for nickel distribution were also found to be identical to Fe or Cr distribution. The mode of solute distribution has been discussed in relation to oxidation behaviour of these alloys.

  19. Hydriding failure in water reactor fuel elements

    International Nuclear Information System (INIS)

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  20. Oxide properties of autoclaved zircaloy cladding tubes investigated by the photoelectric polarization method

    International Nuclear Information System (INIS)

    Nystrand, A.C.

    2000-06-01

    Corrosion of zirconium alloys is an important lifetime limiting factor for the nuclear reactor fuel. The corrosion resistance of a metal is highly dependent on the ability of the surface metal oxide to transport electrons and ions, which is related to the stoichiometry of the oxide and the oxide defect concentration. The Photoelectric Polarization Method (PEP) is a structure sensitive method which earlier has been investigated as a possible method to study the defect structure in zirconium oxides. The purpose of the following work is, by using more optimized experimental equipment, to verify if the PEP method is a suitable method to study the defect structure in zirconium oxides and to predict the corrosion resistance for different zirconium alloys. The conclusions from the experiments are as follows: - The modifications of the experimental setup by means of a new source of light (deuterium lamp) and a new oscilloscope with an amplifier gave distinct Vpep signals. - The photoresponse is negative for all types of cladding and under all kind of oxidation regimes and hence the oxide is a n-type semiconductor with deficiency of oxygen. - The method needs to be verified by testing semiconductors with a known defect concentration

  1. Synthesis, Characterization and Antimicrobial Activity of Zirconium (IV) Complexes

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Shobhana; Jain, Asha; Saxena, Sanjiv [Univ. of Rajasthan, Jaipur (India)

    2012-08-15

    Heteroleptic complexes of zirconium (IV) derived from bulky Schiff base ligands containing a sulphur atom and oximes of heterocyclic β-diketones of the general formula ZrLL' (where L'H{sub 2}=RCNH(C{sub 6}H{sub 4})SC : C(OH)N(C{sub 6}H{sub 5})N : CCH{sub 3}, R=-C{sub 6}H{sub 5}, -C{sub 6}H{sub 4}Cl(p) and L'H{sub 2}=R'C : (NOH)C : C(OH)N(C{sub 6}H{sub 5})N : CCH{sub 3}, R' = -CH{sub 2}CH{sub 3}, -C{sub 6}H{sub 5}, -C{sub 6}H{sub 4}Cl (p) were prepared by the reactions of zirconium tetrachloride with disodium salts of Schiff bases (L Na{sub 2}) and oximes of heterocyclic β-diketones (L' Na{sub 2}) in 1:1:1 molar ratio in dry refluxing THF. The structures of these monomeric zirconium (IV) complexes were elucidated with the help of elemental analysis, molecular weight measurements, spectroscopic (IR, NMR and mass) studies. A distorted trigonal bipyramidal geometry may be suggested for these heteroleptic zirconium (IV) complexes. The ligands (bulky Schiff base ligands containing a sulphur atom and oximes of heterocyclic β-diketones) and their heteroleptic complexes of zirconium (IV) were screened against A. flavus, P. aeruginesa and E. coli.

  2. Reuse of spent fuel cladding Zr by molten salt toward advanced recycle society

    International Nuclear Information System (INIS)

    Amano, Osamu; Kobayashi, Hiroaki; Suzuki, Kazunori; Yasuike, Y.; Sato, Nobuaki

    2003-01-01

    Cladding tubes of zircaloy 95% generated from reprocessing process for spent nuclear fuels are to be chopped in about 3 cm length, compacted and solidified with cements. This paper reports the summary of investigation of the present possible techniques for zirconium recovery: (1) electrolysis of molten salts (Zr-chlorides and/or fluorides) and (2) separation as volatile zirconium chlorides (ZrCl 4 ) (chloride volatility process) followed by reaction with metallic magnesium at 900degC to produce sponged Zr (Kroll method). The feasibility are discussed from the point of view of reduction of secondary radioactive wastes, accumulation of such nuclides as Co-60 and Ni-63 in electrolytic basin, radioactivity estimation in the products, and also problems of cleaning and reducing chemicals. (S. Ohno)

  3. 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)

  4. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-01-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

  5. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-12-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

  6. 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

  7. 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

  8. 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)

  9. 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.

  10. Neutron activation of chlorine in zirconium and zirconium alloys use of the matrix as comparator

    International Nuclear Information System (INIS)

    Cohen, I.M.; Gomez, C.D.; Mila, M.I.

    1981-01-01

    A procedure is described for neutron activation analysis of chlorine in zirconium and zirconium alloys. Calculation of chlorine concentration is performed relative to zirconium concentration in the matrix in order to minimize effects of differences in irradiation and counting geometry. Principles of the method and the results obtained are discussed. (author)

  11. Stone cladding engineering

    National Research Council Canada - National Science Library

    Camposinhos, Rui de Sousa

    2014-01-01

    .... Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM...

  12. 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

  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. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

    International Nuclear Information System (INIS)

    Wang, Jun; Mccabe, Mckinleigh; Wu, Lei; Dong, Xiaomeng; Wang, Xianmao; Haskin, Troy Christopher; Corradini, Michael L.

    2017-01-01

    Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

  15. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jun, E-mail: jwang564@wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Mccabe, Mckinleigh [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wu, Lei [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Dong, Xiaomeng [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wang, Xianmao [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Haskin, Troy Christopher [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Corradini, Michael L., E-mail: corradini@engr.wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States)

    2017-03-15

    Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

  16. Spectrophotometric titration of zirconium in siliceous materials

    International Nuclear Information System (INIS)

    Sugawara, K.F.; Su, Y.-S.; Strzegowski, W.R.

    1978-01-01

    An accurate and selective complexometric titration procedure based upon a spectrophotometrically detected end-point has been developed for the determination of zirconium in glasses, glass-ceramics and refractories. A p-bromomandelic acid separation step for zirconium imparts excellent selectivity to the procedure. The method is particularly important for the 1 to 5% concentration range where a simple, accurate and selective method for the determination of zirconium has been lacking. (author)

  17. Voltammetric determination of zirconium using azo compounds

    International Nuclear Information System (INIS)

    Orshulyak, O.O.; Levitskaya, G.D.

    2008-01-01

    The optimum conditions for zirconium complexation with azo compounds are found. The applicability of Eriochrome Red B, Calcon, and Calcion to the voltammetric determination of zirconium, total Zr(IV) and Hf(IV), and Zr(IV) in the presence of Zn(II), Cu(II), Cd(II), Ni(II), or Ti(IV) is demonstrated. The developed procedures are used to determine zirconium in a terbium alloy and in an alloy for airplane wheel drums [ru

  18. Applications for zirconium and columbium alloys

    International Nuclear Information System (INIS)

    Condliff, A.F.

    1986-01-01

    Currently, zirconium and columbium are used in a wide range of applications, overlapping only in the field of corrosion control. As a construction material, zirconium is primarily used by the nuclear power industry. The use of zirconium in the chemical processing industry (CPI) is, however, increasing steadily. Columbian alloys are primarily applied as superconducting alloys for research particle accelerators and fusion generators as well as in magnetic resonance imaging for medical diagnosis

  19. Zirconium Phosphate Supported MOF Nanoplatelets.

    Science.gov (United States)

    Kan, Yuwei; Clearfield, Abraham

    2016-06-06

    We report a rare example of the preparation of HKUST-1 metal-organic framework nanoplatelets through a step-by-step seeding procedure. Sodium ion exchanged zirconium phosphate, NaZrP, nanoplatelets were judiciously selected as support for layer-by-layer (LBL) assembly of Cu(II) and benzene-1,3,5-tricarboxylic acid (H3BTC) linkers. The first layer of Cu(II) is attached to the surface of zirconium phosphate through covalent interaction. The successive LBL growth of HKUST-1 film is then realized by soaking the NaZrP nanoplatelets in ethanolic solutions of cupric acetate and H3BTC, respectively. The amount of assembled HKUST-1 can be readily controlled by varying the number of growth cycles, which was characterized by powder X-ray diffraction and gas adsorption analyses. The successful construction of HKUST-1 on NaZrP was also supported by its catalytic performance for the oxidation of cyclohexene.

  20. 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])

  1. 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)

  2. 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)

  3. Microhardness and microplasticity of zirconium nitride

    International Nuclear Information System (INIS)

    Neshpor, V.S.; Eron'yan, M.A.; Petrov, A.N.; Kravchik, A.E.

    1978-01-01

    To experimentally check the concentration dependence of microhardness of 4 group nitrides, microhardness of zirconium nitride compact samples was measured. The samples were obtained either by bulk saturation of zirconium iodide plates or by chemical precipitation from gas. As nitrogen content decreased within the limits of homogeneity of zirconium nitride samples where the concentration of admixed oxygen was low, the microhardness grew from 1500+-100 kg/mm 2 for ZrNsub(1.0) to 27000+-100 kg/mm 2 for ZrNsub(0.78). Microplasticity of zirconium nitride (resistance to fracture) decreased, as the concentration of nitrogen vacancies was growing

  4. Production kinetics of zirconium tetrachloride

    International Nuclear Information System (INIS)

    Sudjoko, D.; Masduki, B.; Sunardjo; Sulistyo, B.

    1996-01-01

    This research was intended to study the kinetics of zirconium tetrachloride production. The process was carried out in semi continuous reactor, equipped with heater, temperature controller, sublimator and scrubber. The variables investigated were time, temperature and the pellet forming pressure. Within the range of variables studied, the expression of the process in the chemical reaction controller region and diffusion controller region were both presented. (author)

  5. Microstructural aspects of the oxidation of zirconium alloys

    International Nuclear Information System (INIS)

    Proff, Ch.

    2011-01-01

    This thesis is focused on the microstructural characterisation of precipitates in the oxide of binary zirconium alloys (1 wt.% Fe, Cr or Ni or 0.6 wt.% Nb) under different oxidation conditions at 415 C. The samples were oxidised in autoclave in air and steam and in an environmental scanning electron microscope in water vapour. The microstructural evolution of the precipitates during oxidation was characterised using electron microscopy. The findings from the analysis are the following: -Two types of oxidation behaviour are observed for precipitates. -Pilling Bedworth ratio of precipitates is higher than that of the zirconium matrix. -Formation of pure iron oxide crystals on the surface for iron bearing precipitates close to or at the surface. From these observations it is concluded that the precipitate oxidation behaviour can be correlated to precipitate composition and oxidation tendency of the elements in the precipitates. Iron exhibits clearly different behaviour. (author)

  6. Characteristics and properties of cladding tubes for VVER-1000 higher Uranium content fuel rods

    International Nuclear Information System (INIS)

    Peregud, M.; Markelov, A.; Novikov, V.; Gusev, A.; Konkov, V.; Pimenov, Y.; Agapitov, V.; Shtutsa, M.

    2009-01-01

    To improve the fuel cycle economics and to further increase the VVER fuel usability the work programme is under way to design novel improved fuel, fuel rods and fuel assemblies. Longer FA operation time that is needed to increase the fuel burnup and the related design developments of novel fuel assemblies resulted not only in changing types and sizes of Zirconium items and fuel assembly components but also altered the requirements placed on their technical characteristics. To use fuel rods having a larger charge of fuel, to improve their behaviour in LOCA, to reduce fuel rod damage ability during assembling the work was carried out to perfect the characteristics of both the cladding (reduced wall thickness and more rigid tolerances for geometry) and its material. To meet the more rigid requirements for the geometry dimensions of cladding tubes an improved process flow sheet has been designed and employed for their fabrication and also the finishing treatment of tube surfaces has been improved. The higher and stable properties of the cladding materials were managed through using the special purity in terms of Hafnium Zirconium (not higher than 100 ppm Hf) as a base of the E110 alloy and maintaining within the valid specifications for the alloy the optimized contents of Oxygen and Iron at the levels of (600 - 990) ppm and (250 - 700) ppm, respectively. The work was under way in 2004 - 2008 years; during this period the technology and materials science solutions were mastered that were phased-in introduced into the production of the cladding tubes for the fuels loaded into the of the Kalinin NPP Unit 1

  7. Hydrogen content in titanium and a titanium–zirconium alloy after acid etching

    Energy Technology Data Exchange (ETDEWEB)

    Frank, Matthias J.; Walter, Martin S. [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Lyngstadaas, S. Petter [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway); Wintermantel, Erich [Institute of Medical and Polymer Engineering, Chair of Medical Engineering, Technische Universität München, Boltzmannstrasse 15, 85748 Garching (Germany); Haugen, Håvard J., E-mail: h.j.haugen@odont.uio.no [Department of Biomaterials, Institute for Clinical Dentistry, University of Oslo, P.O. Box 1109, Blindern, NO-0317 Oslo (Norway)

    2013-04-01

    Dental implant alloys made from titanium and zirconium are known for their high mechanical strength, fracture toughness and corrosion resistance in comparison with commercially pure titanium. The aim of the study was to investigate possible differences in the surface chemistry and/or surface topography of titanium and titanium–zirconium surfaces after sand blasting and acid etching. The two surfaces were compared by X-ray photoelectron spectroscopy, secondary ion mass spectroscopy, scanning electron microscopy and profilometry. The 1.9 times greater surface hydrogen concentration of titanium zirconium compared to titanium was found to be the major difference between the two materials. Zirconium appeared to enhance hydride formation on titanium alloys when etched in acid. Surface topography revealed significant differences on the micro and nanoscale. Surface roughness was increased significantly (p < 0.01) on the titanium–zirconium alloy. High-resolution images showed nanostructures only present on titanium zirconium. - Highlights: ► TiZr alloy showed increased hydrogen levels over Ti. ► The alloying element Zr appeared to catalyze hydrogen absorption in Ti. ► Surface roughness was significantly increased for the TiZr alloy over Ti. ► TiZr alloy revealed nanostructures not observed for Ti.

  8. 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)

  9. 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

  10. Investigation and recovery of unrecovered fuel pellets and cladding tube pieces

    International Nuclear Information System (INIS)

    Kobayashi, Keiji

    1980-01-01

    The total weight of the fuel pellets lost due to break was about 1206 g, and cladding tube pieces were about 217 g. Among these, the pellets of about 527 g and the cladding tube pieces of about 152 g were recovered when broken fuel rods were discovered. It is not desirable to leave these broken pieces as unrecovered in view of safety and the management of nuclear fuel materials. Kansai Electric Power Co., Inc., investigated the position and the amount of these pellets and cladding tube pieces for about a year, and recovered a part of them. The results were written in two reports. The objects of the investigation and recovery, and the method of recovery are explained. The UO 2 and zirconium recovered were 58.52 g and 369.58 g, respectively. The solid pellets were recovered from the reactor, fuel assemblies, a spent fuel pit and canals, and the content in sludge was recovered from other installations. The amounts of unrecovered pellets and cladding tube pieces in primary cooling water, coolant filters, sealing water filters, primary cooling pipes, waste resins and fuel assemblies were estimated. The problems concerning the recovery and estimation are pointed out. The results of estimating the amount of uranium in coolant filters and sealing water filters are useful to know the time of the occurrence of accident. (Kako, I.)

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

    International Nuclear Information System (INIS)

    Stout, R.B.

    1989-10-01

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

  12. FOIL ELEMENT FOR NUCLEAR REACTOR

    Science.gov (United States)

    Noland, R.A.; Walker, D.E.; Spinrad, B.I.

    1963-07-16

    A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

  13. 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)

  14. Improved nuclear fuel element

    International Nuclear Information System (INIS)

    Klepfer, H.H.

    1974-01-01

    A nuclear fuel element is described which comprises: 1) an elongated clad container, 2) a layer of high lubricity material being disposed in and adjacent to the clad container, 3) a low neutron capture cross section metal liner being disposed in the clad container and adjacent to the layer, 4) a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, 5) an enclosure integrally secured and sealed at each end of the container, and a nuclear fuel material retaining means positioned in the cavity. (author)

  15. 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)

  16. Nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  17. Nuclear fuel element

    International Nuclear Information System (INIS)

    Thompson, J.R.; Rowland, T.C.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  18. 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

  19. Biomechanical testing of zirconium dioxide osteosynthesis system for Le Fort I advancement osteotomy fixation.

    Science.gov (United States)

    Hingsammer, Lukas; Grillenberger, Markus; Schagerl, Martin; Malek, Michael; Hunger, Stefan

    2018-01-01

    The following work is the first evaluating the applicability of 3D printed zirconium dioxide ceramic miniplates and screws to stabilize maxillary segments following a Le-Fort I advancement surgery. Conventionally used titanium and individual fabricated zirconium dioxide miniplates were biomechanically tested and compared under an occlusal load of 120N and 500N using 3D finite element analysis. The overall model consisted of 295,477 elements. Under an occlusal load of 500N a safety factor before plastic deformation respectively crack of 2.13 for zirconium dioxide and 4.51 for titanium miniplates has been calculated. From a biomechanical point of view 3D printed ZrO 2 mini-plates and screws are suggested to constitute an appropriate patient specific and metal-free solution for maxillary stabilization after Le Fort I osteotomy. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Analysis of hydrogen in zirconium metallic

    International Nuclear Information System (INIS)

    Rodrigues, A.N.; Vega Bustillos, J.O.W.

    1991-02-01

    Determination of hydrogen in zirconium metallic have been performed using the hot vacuum extraction system and the gas chromatographic technique. The zirconium metallic samples were hydrieded by electrolitic technique at difference temperatures and times, then the samples were annealing at vacuum and eatching by fluoridric acid solution. The details of the hydrieded process, analytical technique and the data obtained are discussed. (author)

  1. Expanded heat treatment to form residual compressive hoop stress on inner surface of zirconium alloy tubing

    International Nuclear Information System (INIS)

    Megata, Masao

    1997-01-01

    A specific heat treatment process that introduces hoop stress has been developed. This technique can produce zirconium alloy tubing with a residual compressive hoop stress near the inner surface by taking advantage of the mechanical anisotropy in hexagonal close-packed zirconium crystal. Since a crystal having its basal pole parallel to the tangential direction of the tubing is easier to exhibit plastic elongation under the hoop stress than that having its basal pole parallel to the radial direction, the plastic and elastic elongation can coexist under a certain set of temperature and hoop stress conditions. The mechanical anisotropy plays a role to extend the coexistent stress range. Thus, residual compressive hoop stress is formed at the inner surface where more plastic elongation occurs during the heat treatment. This process is referred to as expanded heat treatment. Since this is a fundamental crystallographic principle, it has various applications. The application to improve PCI/SCC (pellet cladding interaction/stress corrosion cracking) properties of water reactor fuel cladding is promising. Excellent results were obtained with laboratory-scale heat treatment and an out-reactor iodine SCC test. These results included an extension of the time to SCC failure. (author)

  2. Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

    Energy Technology Data Exchange (ETDEWEB)

    Michael V. Glazoff

    2013-08-01

    Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel. This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

  3. Autoclave Testing on Zirconium Alloy Materials

    International Nuclear Information System (INIS)

    Hoffmann, Petra-Britt; Sell, Hans-Juergen; Garzarolli, Friedrich

    2012-09-01

    The corrosion of Zirconium components like fuel rod claddings and spacer grids is limiting lifetime and duty of these components. In Pressurized and Boiling Water Reactors (PWR and BWR), different corrosion phenomena are of interest. Although in-pile experience is the final proof for a material development, significant experience was gained by autoclave tests, trying to simulate in-pile conditions but reducing time for return of experience by increased temperatures. For PWR application, the uniform corrosion is studied in water at up to 370 deg. C and in high pressure steam at 400 deg. C, and for BWR, the nodular corrosion is studied in high pressure steam at 500-520 deg. C. Particular attention has to be given to the corrosion media, because oxidative traces in the water can significantly affect the corrosion response. An extensive air removal is thus important for all corrosion tests. This links to the different water chemistry conditions that have been investigated as separate effects otherwise difficult to separate under in-pile conditions. Uniform corrosion in 350 deg. C water is usually a cyclic process with repeated rate transitions. In addition, at high exposure times an acceleration of corrosion can occur, e.g. for Zr-Sn alloys with a high Sn content. In 400 deg. C steam, corrosion rate decreases somewhat with increasing time. Uniform corrosion rate of Zr alloys depends on their Sn- and Fe+Cr contents as well as on their annealing parameters with a similar trend as in PWR and on their yield strength, however with an opposite trend compared to BWR conditions. Nodular corrosion of BWR alloys depends on the annealing parameter with a similar trend as in PWR and out-of-reactor also significantly on the Fe+Cr content. The hydrogen pickup fraction (HPUF) depends largely on details of the water chemistry and can particularly depend on autoclave degassing and probably also on autoclave contaminations. Thus any HPUF value from out-of- pile corrosion tests is only

  4. Zirconium behaviour in purex process solutions

    International Nuclear Information System (INIS)

    Shu, J.

    1982-01-01

    The extraction behaviour of zirconium, as fission product, in TBP/diluent- HNO 3 -H 2 O systems, simulating Purex solutions, is studied. The main purpose is to attain an increasing in the zirconium decontamination factor by adjusting the extraction parameters. Equilibrium diagram, TBP concentration, aqueous:organic ratio, salting-out effects and, uranium loading in the organic phase were the main factors studied. All these experiments had been made with zirconium in the 10 - 2 - 10 - 3 concentration range. The extractant degradation products influence uppon the zirconium behaviour was also verified. With the obtained data it was possible to introduce some modification in the standard Purex flow-sheet in order to obtain the uranium product with higher zirconium decontamination. (Author) [pt

  5. 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

  6. 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

  7. Influence of irradiation and radiolysis on the corrosion rates and mechanisms of zirconium alloys

    International Nuclear Information System (INIS)

    Verlet, Romain

    2015-01-01

    The nuclear fuel of pressurized water reactors (PWR) in the form of uranium oxide UO 2 pellets (or MOX) is confined in a zirconium alloy cladding. This cladding is very important because it represents the first containment barrier against the release of fission products generated by the nuclear reaction to the external environment. Corrosion by the primary medium of zirconium alloys, particularly the Zircaloy-4, is one of the factors limiting the reactor residence time of the fuel rods (UO 2 pellets + cladding). To optimize core management and to extend the lifetime of the fuel rods in reactor, new alloys based on zirconium-niobium (M5) have been developed. However, the corrosion mechanisms of these are not completely understood because of the complexity of these materials, corrosion environment and the presence of radiation from the nuclear fuel. Therefore, this thesis specifically addresses the effects of radiolysis and defects induced by irradiation with ions in the matrix metal and the oxide layer on the corrosion rate of Zircaloy-4 and M5. The goal is to separate the influence of radiation damage to the metal, that relating to defects created in the oxide and that linked to radiolysis of the primary medium on the oxidation rate of zirconium alloys in reactor. 1) Regarding effect of irradiation of the metal on the oxidation rate: type dislocation loops appear and increase the oxidation rate of the two alloys. For M5, in addition to the first effect, a precipitation of fines needles of niobium reduced the solid solution of niobium concentration in the metal and ultimately in the oxide, which strongly reduces the oxidation rate of the alloy. 2) Regarding the effect of irradiation of the oxide layer on the oxidation rate: defects generated by the nuclear cascades in the oxide increase the oxidation rate of the two materials. For M5, germination of niobium enriched zones in irradiated oxide also causes a decrease of the niobium concentration in solid solution

  8. Protection of zirconium and its alloys by metallic coatings

    International Nuclear Information System (INIS)

    Loriers, H.; Lafon, A.; Darras, R.; Baque, P.

    1968-01-01

    At 600 deg. C in an atmosphere of carbon dioxide, zirconium and its alloys undergo corrosion which presents two aspects simultaneously: - formation of a surface layer of zirconia, - dissolution of oxygen in the alloy sub-layer leading to brittleness. The two phenomena greatly restrict the possibilities of using zirconium alloys as a canning material for fuel elements in CO 2 cooled nuclear reactors. An attempt has thus been made to limit, and perhaps to suppress, the corrosion effects in zirconium under these conditions by protecting it with metallic coatings. A first attempt to obtain a protection using copper-based coatings did not produce the result hoped for. Aluminium coatings produced by vacuum evaporation, followed by a consolidating thermal treatment make it possible to prevent the formation of the zirconia layer, but they do not eliminate the hardening effect produced by oxygen diffusion. On the other hand, electrolytically produced chromium deposits whose adherence is improved by a thermal vacuum treatment, counteract both these phenomena simultaneously. A similar result has been obtained with coatings of molybdenum produced by the technique of high-frequency inductive plasma sputtering. The particular effectiveness of the last two types of coatings is due to their structures characterized by the existence of an adherent film of chromium or molybdenum in the free state. (authors) [fr

  9. Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

    2002-01-01

    Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

  10. 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

  11. Thermodynamical, kinetic and structural properties of simple and mixed complexes of zirconium (IV) with the hydroxyl and carbonate ions. Study by potentiometry, Raman spectroscopy and NMR spectroscopy

    International Nuclear Information System (INIS)

    Veyland, A.

    1999-01-01

    Most of zirconium production is used by the nuclear industry for the cladding of nuclear fuels and for the storage of radioactive wastes. The aim of this work is the qualitative and quantitative study of the complexes made by zirconium with the hydroxyl and carbonate ions in order to evaluate the long-term pollution risks linked with the corrosion of confinement containers. The zirconium(IV)/hydroxyl system is studied by proto-metry and follows a protocol which reduces the local over-concentrations of reagent and the precipitation of zirconium hydroxide. In potassium nitrate environment and in a pH range of 1.5 to 3.5, three soluble species are evidenced: Zr(OH) 3 + , Zr 2 (OH) 7 + and Zr(OH) 4 . Their apparent constant of formation and the solubility product of zirconium hydroxide are determined with 4 ionic forces. Using these results, the corresponding thermodynamic constants are calculated by applying the theory of specific interactions. The formation of zirconium (IV) hydroxo-carbonate complexes is evidenced by proto-metry and 17 O and 13 C NMR. The number of carbonates fixed by zirconium varies from 0 to 4. The dialysis indicates that the degree of poly-condensation of species is an inverse function of the number of complex carbonates. The Raman polarized spectra and the 13 C NMR results demonstrate for all complexes the bidentate character of the complexation mode of the carbonates. The dynamical study made by 13 C NMR of the exchange between complexed carbonates of the tetra-carbonate-zirconate ion and free carbonates in solution allows to determine the kinetic constants and the corresponding velocity law. An associative mechanism is proposed, in agreement with the results obtained by mass spectroscopy with electro-spray ionization. These new thermodynamical and kinetic data allow to model the speciation of zirconium in natural waters. (J.S.)

  12. Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

    International Nuclear Information System (INIS)

    Stout, R.B.

    2001-01-01

    An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

  13. Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

    Energy Technology Data Exchange (ETDEWEB)

    Stout, R.B. [California Univ., Livermore, CA (United States). Lawrence Livermore National Lab

    2001-07-01

    An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

  14. Hydrides blister formation and induced embrittlement on zircaloy-4 cladding tubes in reactivity initiated conditions

    International Nuclear Information System (INIS)

    Hellouin-De-Menibus, A.

    2012-01-01

    Our aim is to study the cladding fracture with mechanical tests more representative of RIA conditions, taking into account the hydrides blisters, representative strain rates and stress states. To obtain hydride blisters, we developed a thermodiffusion setup that reproduces blister growth in reactor conditions. By metallography, nano-hardness, XRD and ERDA, we showed that they are constituted by 80% to 100% of δ hydrides in a Zircaloy-4 matrix, and that the zirconium beneath has some radially oriented hydrides. We modeled the blister growth kinetics taking into account the hysteresis of the hydrogen solubility limit and defined the thermal gradient threshold for blister growth. The modeling of the dilatometric behavior of hydrided zirconium indicates the important role of the material crystallographic texture, which could explain differences in the blister shape. Mechanical tests monitored with an infrared camera showed that significant local heating occurred at strain rates higher than 0.1/s. In parallel, the Expansion Due to Compression test was optimized to increase the bi-axiality level from uniaxial stress to plane strain (HB-EDC and VHB-EDC tests). This increase in loading bi-axiality lowers greatly the fracture strain at 25 C and 350 C only in homogeneous material without blister. Eventually, the ductility decrease of unirradiated Zircaloy-4 cladding tube in function of the blister depth was quantified. (author) [fr

  15. Hydrogen desorption kinetics from zirconium hydride and zirconium metal in vacuum

    International Nuclear Information System (INIS)

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

    2014-01-01

    The kinetics of hydrogen desorption from zirconium hydride is important in many nuclear design and safety applications. In this paper, a coordinated experimental and modeling study has been used to explicitly demonstrate the applicability of existing kinetic theories for hydrogen desorption from zirconium hydride and α-zirconium. A static synthesis method was used to produce δ-zirconium hydride, and the crystallographic phases of the zirconium hydride were confirmed by X-ray diffraction (XRD). Three obvious stages, involving δ-zirconium hydride, a two-phase region, and α-zirconium, were observed in the hydrogen desorption spectra of two zirconium hydride specimens with H/Zr ratios of 1.62 and 1.64, respectively, which were obtained using thermal desorption spectroscopy (TDS). A continuous, one-dimensional, two-phase moving boundary model, coupled with the zero- and second-order kinetics of hydrogen desorption from δ-zirconium hydride and α-zirconium, respectively, has been developed to reproduce the TDS experimental results. A comparison of the modeling predictions with the experimental results indicates that a zero-order kinetic model is valid for description of hydrogen flux away from the δ-hydride phase, and that a second-order kinetic model works well for hydrogen desorption from α-Zr if the activation energy of desorption is optimized to be 70% of the value reported in the literature

  16. Methods for the preparation of ultra-pure anhydrous zirconium tetrafluoride from zirconium tetraborohydride, researches in connection with halide glasses

    International Nuclear Information System (INIS)

    Tortevois, R.

    1990-01-01

    The synthesis of ultrapure zirconium tetrafluoride, the main component of fluorozirconate based optical fibers, was successfully attempted from zirconium tetraborohydride. Of the fluorinating agents used, nitrogen trifluoride doesn't react with zirconium tetraborohydride while xenon difluoride reacts too violently and leads to phases which contain boron. The fluorination in a compatible solvent enabled us to minimize the degradation. The best results were obtained with the fluorination of Zr(BH 4 ) 4 dissolved in CFCl 3 at -40 deg C by anhydrous HF. Using several analytical methods such as graphite furnace atomic absorption and proton activation, we analyzed the purity. The degree of transition element impurities is less than the ppm level for ZrF 4 . The dehydration of ZrF 4 ,H 2 O and ZrF 4 ,3H 2 O at room temperature by CIF 3 in gaseous and liquid state was also investigated. At exceptionally low temperature, this process allows oxide and oxyfluoride components to be reduced

  17. Temperature distributions in cladded configurations of complicated boundary shape

    International Nuclear Information System (INIS)

    Laura, P.A.A.; Gutierrez, R.H.; Sarmiento, G.S.

    1981-01-01

    This pape deals with an approximate analytical solution of certain cladded shapes of basic interest in nuclear engineering technology. In one case a solution of Laplace's equation is obtained. In the other, a solution to a system of partial differential equations (Poisson's and Laplace's) is attained. A approximate conformal mapping approach is also used in both situation, and it is shown that the analytical predictions are in good agreement with the values obtained by means of a finite element code. (orig.)

  18. High-temperature steam-oxidation behavior of Zr-1Nb-1Sn-0.1Fe cladding tube at temperatures of 800-1000

    International Nuclear Information System (INIS)

    Lee, Cheol Min; Cho, Tae Won; Jeong, Gwan Yoon; Kim, Mi Jin; Kim, Ji Hyeon; Lee, Hee Jae; Sohn, Dong Seong; Mok, Yong Kyoon

    2016-01-01

    To prevent cladding failure, NRC issued a regulation Title 10 § 50.46, which specifies cladding temperature of 1204 .deg. C and 17% ECR should not be exceeded. The fundamental reason of the mechanical degradation of cladding is the formation of the oxide which is brittle. Theoretically, the oxide layer is formed following parabolic rate. However, from many experiments, sub-parabolic rates are often observed. There have been many suggestions so far; chemical and stress gradient across the oxide layer could initiate the sub-parabolic rate, the phase transformation of Zirconium dioxide from tetragonal to monoclinic could be the reason, change of the grain size of Zirconium dioxide could cause the cubic oxidation rate, and there is a suggestion that if electron migration is the major mechanism of the oxide growth, then the subparabolic rate can show up. However, the reason why the sub-parabolic rate appears is still not certain. Another important degradation mechanism is breakaway oxidation. A clear explanation that why the breakaway oxidation appears is still not clear. Most of the people believe the phase transformation of Zirconium dioxide cause instability within the oxide, which causes breakaway oxidation to appear. However, how much effect is caused from the phase transformation is not so sure. In this study, detailed analysis about the oxidation kinetics and the breakaway oxidation of Zr-1Nb-1Sn- 0.1Fe were carried out at temperatures between 800 - 1000 .deg. C.

  19. Use of multiscale zirconium alloy deformation models in nuclear fuel behavior analysis

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, Robert, E-mail: robert.montgomery@pnnl.gov [Pacific Northwest National Laboratory (United States); Tomé, Carlos, E-mail: tome@lanl.gov [Los Alamos National Laboratory (United States); Liu, Wenfeng, E-mail: wenfeng.liu@anatech.com [ANATECH Corporation (United States); Alankar, Alankar, E-mail: alankar.alankar@iitb.ac.in [Indian Institute of Technology Bombay (India); Subramanian, Gopinath, E-mail: gopinath.subramanian@usm.edu [University of Southern Mississippi (United States); Stanek, Christopher, E-mail: stanek@lanl.gov [Los Alamos National Laboratory (United States)

    2017-01-01

    Accurate prediction of cladding mechanical behavior is a key aspect of modeling nuclear fuel behavior, especially for conditions of pellet-cladding interaction (PCI), reactivity-initiated accidents (RIA), and loss of coolant accidents (LOCA). Current approaches to fuel performance modeling rely on empirical constitutive models for cladding creep, growth and plastic deformation, which are limited to the materials and conditions for which the models were developed. To improve upon this approach, a microstructurally-based zirconium alloy mechanical deformation analysis capability is being developed within the United States Department of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL). Specifically, the viscoplastic self-consistent (VPSC) polycrystal plasticity modeling approach, developed by Lebensohn and Tomé [1], has been coupled with the BISON engineering scale fuel performance code to represent the mechanistic material processes controlling the deformation behavior of light water reactor (LWR) cladding. A critical component of VPSC is the representation of the crystallographic nature (defect and dislocation movement) and orientation of the grains within the matrix material and the ability to account for the role of texture on deformation. A future goal is for VPSC to obtain information on reaction rate kinetics from atomistic calculations to inform the defect and dislocation behavior models described in VPSC. The multiscale modeling of cladding deformation mechanisms allowed by VPSC far exceed the functionality of typical semi-empirical constitutive models employed in nuclear fuel behavior codes to model irradiation growth and creep, thermal creep, or plasticity. This paper describes the implementation of an interface between VPSC and BISON and provides initial results utilizing the coupled functionality.

  20. Study of the hoop fracture behaviour of nuclear fuel cladding from ring compression tests by means of non-linear optimization techniques

    Energy Technology Data Exchange (ETDEWEB)

    Gómez, F.J., E-mail: javier.gomez@amsimulation.com [Advanced Material Simulation, AMS, Bilbao (Spain); Martin Rengel, M.A., E-mail: mamartin.rengel@upm.es [E.T.S.I. Caminos, Canales y Puertos, Universidad Politécnica de Madrid, C/Professor Aranguren SN, E-28040 Madrid (Spain); Ruiz-Hervias, J.; Puerta, M.A. [E.T.S.I. Caminos, Canales y Puertos, Universidad Politécnica de Madrid, C/Professor Aranguren SN, E-28040 Madrid (Spain)

    2017-06-15

    In this work, the hoop fracture toughness of ZIRLO{sup ®} fuel cladding is calculated as a function of three parameters: hydrogen concentration, temperature and displacement rate. To this end, pre-hydrided samples with nominal hydrogen concentrations of 0 (as-received), 150, 250, 500, 1200 and 2000 ppm were prepared. Hydrogen was precipitated as zirconium hydrides in the shape of platelets oriented along the hoop direction. Ring Compression Tests (RCTs) were conducted at three temperatures (20, 135 and 300 °C) and two displacement rates (0.5 and 100 mm/min). A new method has been proposed in this paper which allows the determination of fracture toughness from ring compression tests. The proposed method combines the experimental results, the cohesive crack model, finite elements simulations, numerical calculations and non-linear optimization techniques. The parameters of the cohesive crack model were calculated by minimizing the difference between the experimental data and the numerical results. An almost perfect fitting of the experimental results is achieved by this method. In addition, an estimation of the error in the calculated fracture toughness is also provided.

  1. Concept Feasibility Report for Electroplating Zirconium onto Uranium Foil - Year 2

    Energy Technology Data Exchange (ETDEWEB)

    Coffey, Greg W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Meinhardt, Kerry D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pederson, Larry R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-01

    The Fuel Fabrication Capability within the U.S. High Performance Research Reactor Conversion Program is funded through the National Nuclear Security Administration (NNSA) NA-26 (Office of Material Management and Minimization). An investigation was commissioned to determine the feasibility of using electroplating techniques to apply a coating of zirconium onto depleted uranium/molybdenum alloy (U-10Mo). Electroplating would provide an alternative method to the existing process of hot roll-bonding zirconium foil onto the U-10Mo fuel foil during the fabrication of fuel elements for high-performance research reactors. The objective of this research was to develop a reproducible and scalable plating process that will produce a uniform, 25 μm thick zirconium metal coating on U-10Mo foil. In previous work, Pacific Northwest National Laboratory (PNNL) established a molten salt electroplating apparatus and protocol to plate zirconium metal onto molybdenum foil (Coffey 2015). During this second year of the research, PNNL furthered this work by moving to the U-10Mo alloy system (90 percent uranium:10 percent molybdenum). The original plating apparatus was disassembled and re-assembled in a laboratory capable of handling low-level radioactive materials. Initially, the work followed the previous year’s approach, and the salt bath composition was targeted at the eutectic composition (LiF:NaF:ZrF4 = 26:37:37 mol%). Early results indicated that the formation of uranium fluoride compounds would be problematic. Other salt bath compositions were investigated in order to eliminate the uranium fluoride production (LiF:NaF = 61:39 mol% and LiF:NaF:KF = 46.5:11.5:42 mol% ). Zirconium metal was used as the crucible for the molten salt. Three plating methods were used—isopotential, galvano static, and pulsed plating. The molten salt method for zirconium metal application provided high-quality plating on molybdenum in PNNL’s previous work. A key advantage of this approach is that

  2. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  3. 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

  4. High-temperature air oxidation of E110 and Zr-1%Nb alloys claddings with coatings

    International Nuclear Information System (INIS)

    Kuprin, A.S.; Belous, V.A.; Voyevodin, V.N.; Bryk, V.V.; Vasilenko, R.L.; Ovcharenko, V.D.; Tolmachova, G.N.; V'yugov, P.N.

    2014-01-01

    Results of experimental study of the influence of protective vacuum-arc claddings on the base of compounds zirconium-chromium and of its nitrides on air oxidation resistance at temperatures 660, 770, 900, 1020, 1100 deg C during 3600 s. of tubes produced of zirconium alloys E110 and Zr-1%Nb (calcium-thermal alloy of Ukrainian production) are presented. Change of hardness, the width of oxide layer and depth of oxygen penetration into alloys from the side of coating and without coating are investigated by the methods of nanoindentation and by scanning electron microscopy. It is shown that the thickness of oxide layer in zirconium alloys at temperatures 1020 and 1100 deg C from the side of the coating doesn't exceed 5 μm, and from the unprotected side reaches the value of ≥ 120 μm with porous and rough structure. Tubes with coatings save their shape completely independently of the type of alloy; tubes without coatings deform with the production of through cracks

  5. 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

  6. 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)

  7. 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

  8. Characterization of zirconium alloy oxidation films by alternating current impedance

    International Nuclear Information System (INIS)

    Rosecrans, P.M.

    1983-11-01

    Kinetics of zirocnium alloy oxidation are highly nonlinear. The results of electrochemical measurements and electron microscopy support the existence of porosity in oxide films formed on zirconium alloys in high temperature aqueous environments. Analytical treatment is presented relating oxidation kinetics to the thickness and distribution of nonporous elements within the oxide. This analysis illustrates that both the level and distribution of porosity within the oxide factor into oxidation kinetics. The barrier layer model can provide a basis for predicting the effect of environmental changes on oxidation rate. In addition, it demonstrates the need for further research into porosity generation mechanisms in oxide films

  9. Bis(indenyl)zirconium(IV) complexes of monofunctional bidentate salicylidimines

    International Nuclear Information System (INIS)

    Sharma, A.K.; Khera, B.; Kaushik, N.K.

    1983-01-01

    A series of (C 9 H 7 ) 2 Zr(SB)Cl complexes where SB- is the anion of bidentate Schiff base derived from salicylaldehyde and 4-substitued anilines, viz. salicylidene-4-ansidine, salicylidene-4-phenetidine, salicylidene-4-chloroaniline, salicylidene-4-bromoaniline, salicylene-4-iodoaniline and salicylidene-4-nitroaniline, have been synthesized by the reaction of bis(indenyl)zirconium(IV) dichloride and Schiff base (SBH) in 1 : 1 molar ratio in refluxing THF in the presence of triethylamine. The new derivatives have been characterized on the basis of their elemental analyses, conductance measurements and spectral (IR, 1 H-NMR, UV-VIS) studies. (Author)

  10. URANIUM DECONTAMINATION WITH RESPECT TO ZIRCONIUM

    Science.gov (United States)

    Vogler, S.; Beederman, M.

    1961-05-01

    A process is given for separating uranium values from a nitric acid aqueous solution containing uranyl values, zirconium values and tetravalent plutonium values. The process comprises contacting said solution with a substantially water-immiscible liquid organic solvent containing alkyl phosphate, separating an organic extract phase containing the uranium, zirconium, and tetravalent plutonium values from an aqueous raffinate, contacting said organic extract phase with an aqueous solution 2M to 7M in nitric acid and also containing an oxalate ion-containing substance, and separating a uranium- containing organic raffinate from aqueous zirconium- and plutonium-containing extract phase.

  11. Chemistry of titanium, zirconium and thorium picramates

    International Nuclear Information System (INIS)

    Srivastava, R.S.; Agrawal, S.P.; Bhargava, H.N.

    1976-01-01

    Picramates of titanium, zirconium and thorium are prepared by treating the aqueous sulphate, chloride and nitrate solutions with sodium picramate. Micro-analysis, colorimetry and spectrophotometry are used to establish the compositions (metal : ligand ratio) of these picramates as 1 : 2 (for titanium and zirconium) and 1 : 4 (for thorium). IR studies indicate H 2 N → Me coordination (where Me denotes the metal). A number of explosive properties of these picramates point to the fact that the zirconium picramate is thermally more stable than the picramates of titanium and thorium. (orig.) [de

  12. Experimental study and numerical simulation of the plastic deformation of zirconium single crystals

    International Nuclear Information System (INIS)

    Lebon, C.

    2011-01-01

    There is only few experimental data in the literature on the zirconium single crystals and no constitutive laws for this single crystal material are provided. The goal of this work is then to create an experimental database like the Critical Resolved Shear Stress (CRSS) for the prismatic slip, the strain-hardening, the activation of the prismatic glide system and the activation volumes. We determine theses parameters from image correlation method. Then, we develop a new multi-scale approach using dislocations dynamics concept and finite element computations. Finally, a first single crystal constitutive law for the zirconium is proposed and a good agreement with the experimental data is obtained. (author) [fr

  13. 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

  14. 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

  15. 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.

  16. METMET fuel with Zirconium matrix alloys

    International Nuclear Information System (INIS)

    Savchenko, A.; Konovalov, I.; Totev, T.

    2008-01-01

    The novel type of WWER-1000 fuel has been designed at A.A. Bochvar Institute. Instead of WWER-1000 UO 2 pelletized fuel rod we apply dispersion type fuel element with uniformly distributed high uranium content granules of U9Mo, U5Nb5Zr, U3Si alloys metallurgically bonded between themselves and to cladding by a specially developed Zr-base matrix alloy. The fuel meat retains a controllable porosity to accommodate fuel swelling. The optimal volume ratios between the components are: 64% fuel, 18% matrix, 18% pores. Properties of novel materials as well as fuel compositions on their base have been investigated. Method of fuel elements fabrication by capillary impregnation has been developed. The primary advantages of novel fuel are high uranium content (more than 15% in comparison with the standard UO 2 pelletized fuel rod), low temperature of fuel ( * d/tU) and serviceability under transient conditions. The use of the novel fuel might lead to natural uranium saving and reduced amounts of spent fuel as well as to optimization of Nuclear Plant operation conditions and improvements of their operation reliability and safety. As a result the economic efficiency shall increase and the cost of electric power shall decrease. (authors)

  17. Stable tracer investigations in humans for assessing the biokinetics of ruthenium and zirconium radionuclides

    International Nuclear Information System (INIS)

    Veronese, I.; Cantone, M.C.; Giussani, A.; Maggioni, T.; Birattari, C.; Bondardi, M.; Groppi, F.; Garlaschelli, I.; Werner, E.; Roth, P.; Hoellriegl, V.; Louvat, P.; Felgenhauer, N.; Zilker, Th.

    2003-01-01

    The interest in the biokinetics of ruthenium and zirconium in humans is justified by the potential radiological risk represented by their radionuclides. Only a few data related to the biokinetics of ruthenium and zirconium in humans are available and, accordingly, the biokinetic models currently recommended by the ICRP for these elements are mainly based on data from animal experiments. The use of stable isotopes as tracers, coupled with a proper analytical technique (nuclear activation analysis with protons) for their determination in biological samples, represents an ethically acceptable methodology for biokinetic investigations, being free from any radiation risk for the volunteer subjects. In this work, the results obtained in eight biokinetic investigations for ruthenium, conducted on a total of three healthy volunteers, and six for zirconium, performed on a total of three subjects, are presented and compared to the predictions of the ICRP models. (author)

  18. 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...

  19. Theory of the frictional interaction between nuclear fuel cladding and a cracked ceramic pellet

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1976-02-01

    A summary is presented of the outcome of theoretical work detailed in five publications, reproduced as appendices, which is concerned with the tendency for the cladding tube of nuclear fuel elements to fracture as the result of power cycling or after a sudden upward power excursion. The relationship is shown between the properties of the clad, those of UO 2 pellets, and the tendency of the clad to fail during upward power excursions. The role of interfacial friction is explored and the benefit to be obtained by reducing it is calculated for cases where a soft metal interlayer is present. It is shown that the experimentally-confirmed magnitude of the strain-concentration in the arc of cladding over a radial pellet crack could not arise if there were interfaceons present. Accordingly, these defects, although they do occur in some sliding situations, are thought to be absent from the pellet clas interface in fuel pins. (author)

  20. 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.

  1. Electrochemical Methods for Reprocessing Defective Fuel Elements and for Decontaminating Equipment

    International Nuclear Information System (INIS)

    Mikheykin, S. V.; Rybakov, K. A.; Simonov, V. P.

    2002-01-01

    Reprocessing of fuel elements receives much consideration in nuclear engineering. Chemical and electrochemical methods are used for the purpose. For difficultly soluble materials based on zirconium alloys chemical methods are not suitable. Chemical reprocessing of defective or irradiated fuel elements requires special methods for their decladding because the dissolution of the clad material in nitric acid is either impossible (stainless steel, Zr alloys) or quite slow (aluminium). Fuel elements are cut in air-tight glove-boxes equipped with a dust collector and a feeder for crushed material. Chemical treatment is not free from limitations. For this reason we started a study of the feasibility of electrochemical methods for reprocessing defective and irradiated fuel elements. A simplified electrochemical technology developed makes it possible to recover expensive materials which were earlier wasted or required multi-step treatment. The method and an electrochemical cell are suitable for essentially complete dissolution of any fuel elements, specifically those made of materials which are difficultly soluble by chemical methods

  2. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-07-01

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

  3. 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 ...

  4. 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.

  5. 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.

  6. 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

  7. 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).

  8. The degradation of zirconium alloys in nuclear reactors - a review

    International Nuclear Information System (INIS)

    Lim, D.; Graham, N.A.

    1986-01-01

    This report presents the findings of a survey of available non-Canadian literature on the oxidation and hydriding of zirconium alloys. Much of the literature was found to address the Zircaloys, particularly when used as fuel cladding subjected to a radioactive and oxidizing environment. Hydriding of Zircaloys is mainly attributed to oxidation. The survey revealed that Zr-Nb alloys have been included in some investigations; however, data on the long-term degradation of Zr-2.5 wt% Nb, in particular, were scarce. The reviewed literature did not lead to conclusions regarding the potential for accelerated hydriding due to corrosion at crevices and/or second-phase particles, nor did it lead to conclusions as to the potential for a 'breakaway' in oxidation and hydrogen acquisition in long service life of Zr-Nb alloys. Specific information on service experience in U.S.S.R. power reactors could not be obtained; however, most of the information surveyed leads to the conclusion that fuel channels having Zr-2.5 wt% Nb pressure tubes should perform satisfactorily with respect to degradation from corrosion and hydriding provided they are installed correctly and are not operated under conditions that are far removed from those anticipated in design. 91 refs

  9. Multiscale modelling of hydrogen embrittlement in zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Majevadia, Jassel; Wenman, Mark; Balint, Daniel; Sutton, Adrian [Imperial College London (United Kingdom); Nazarov, Roman [MPIE, Dusseldorf (Germany)

    2013-07-01

    Delayed Hydride Cracking (DHC) is a commonly occurring embrittlement phenomenon in zirconium alloy fuel cladding within Pressurized Water Reactors (PWRs). DHC is caused by the accumulation of hydrogen atoms taken up by the metal, and the formation of brittle hydrides in the vicinity of crack tips. The rate of crack growth is limited by the rate of hydrogen diffusion to the crack, which can be modelled by solving a stress driven diffusion equation that incorporates the elastic interaction between defects. This of interest in the present work. The elastic interaction is calculated by combining defect forces determined through Density Functional Theory (DFT) simulations, and an exact solution for the anisotropic elastic field of an edge dislocation in Zr. making it possible to determine the interaction energy without the need to simulate directly a hydrogen atom in the presence of a crack or dislocation, which is computationally prohibitive with DFT. The result of the elastic interaction energy calculations can be utilised to determine the segregation of hydrogen to a crack tip for varying crack tip geometries, and in the presence of other crystal defects. This is done by implementing a diffusion equation for hydrogen within a discrete dislocation dynamics simulation. In the present work a model has been developed to demonstrate the effect of a single dislocation on hydrogen diffusion to create a Cottrell atmosphere.

  10. Evaluation of cladding residual stresses in clad blocks by measurements and numerical simulations

    International Nuclear Information System (INIS)

    Dupas, P.; Moinereau, D.

    1996-01-01

    Reactor pressure vessels are internally clad with austenitic stainless steel. This welding operation generates residual stresses which can have an important role in integrity assessments. In order to evaluate these stresses, an experimental and numerical programme has been conducted. The experiments includes cladding operations, macrographic analyses, temperature and residual stresses measurements with different methods. According to these measurements, transversal stresses (perpendicular to the welding direction) and longitudinal stresses (parallel to the welding direction) are highly tensile in stainless steel and they are compressive in the HAZ. Finite element calculations were used to simulate both welding operations and post weld heat treatment. These calculations coupled the thermal, metallurgical and mechanical aspects in a 2D representation. Different models were studied including effect of generalised plane strain, transformation plasticity, creep and tempering. The transversal stresses calculated are similar to the measured ones, but the longitudinal stresses showed to be very sensitive to the model used. As expected because of the two-dimension model, the longitudinal stresses can't be well estimated. More work is needed to improve measurements of stresses in depth (important differences appeared between the different methods). A predictive model would be also very useful to determine the thermal loading which is at present dependant on measurements. A 3D calculation appears to be necessary to evaluate longitudinal stresses. (orig.)

  11. Oxidized zirconium on ceramic; Catastrophic coupling.

    Science.gov (United States)

    Ozden, V E; Saglam, N; Dikmen, G; Tozun, I R

    2017-02-01

    Oxidized zirconium (Oxinium™; Smith & Nephew, Memphis, TN, USA) articulated with polyethylene in total hip arthroplasty (THA) appeared to have the potential to reduce wear dramatically. The thermally oxidized metal zirconium surface is transformed into ceramic-like hard surface that is resistant to abrasion. The exposure of soft zirconium metal under hard coverage surface after the damage of oxidized zirconium femoral head has been described. It occurred following joint dislocation or in situ succeeding disengagement of polyethylene liner. We reported three cases of misuse of Oxinium™ (Smith & Nephew, Memphis, TN, USA) heads. These three cases resulted in catastrophic in situ wear and inevitable failure although there was no advice, indication or recommendation for this use from the manufacturer. Copyright © 2016 Elsevier Masson SAS. All rights reserved.

  12. Study for the chlorination of zirconium oxide

    International Nuclear Information System (INIS)

    Seo, E.S.M.; Takiishi, H.; Paschoal, J.O.A.; Andreoli, M.

    1990-12-01

    In the development of new ceramic and metallic materials the chlorination process constitutes step in the formation of several intermediate compounds, such as metallic chlorides, used for the production of high, purity raw materials. Chlorination studies with the aim of fabrication special zirconium-base alloys have been carried out at IPEN. Within this program the chlorination technique has been used for zirconium tetrachloride production from zirconium oxide. In this paper some relevant parameters such as: time and temperature of reaction, flow rate of chloride gas and percentage of the reducing agent which influence the efficiency of chlorination of zirconium oxide are evaluated. Thermodynamical aspects about the reactions involved in the process are also presented. (author)

  13. Towards an understanding of zirconium alloy corrosion

    International Nuclear Information System (INIS)

    Cox, B.

    1976-08-01

    A brief historical summary is given of the development of a programme for understanding the corrosion mechanisms operating for zirconium alloys. A general summary is given of the progress made, so far, in carrying through this programme. (author)

  14. Zirconium determination in refractories (gravimetric method)

    International Nuclear Information System (INIS)

    Capiotto, N.; Narahashi, Y.; Perish, C.G.; Souza, J.R.

    1991-01-01

    The zirconium determination in refractories is described, consisting in two separation methods for eliminating the interferences. The formatted product is calcined at 1100 0 C and determined gravimetrically as Zr P z 07. (author)

  15. 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

  16. METHOD OF IMPROVING CORROSION RESISTANCE OF ZIRCONIUM

    Science.gov (United States)

    Shannon, D.W.

    1961-03-28

    An improved intermediate rinse for zirconium counteracts an anomalous deposit that often results in crevices and outof-the-way places when ordinary water is used to rinse away a strong fluoride etching solution designed to promote passivation of the metal. The intermediate rinse, which is used after the etching solution and before the water, is characterized by a complexing agent for fluoride ions such as aluminum or zirconium nitrates or chlorides.

  17. Air oxidation of Zircaloy-4, M5 (registered) and ZIRLOTM cladding alloys at high temperatures

    International Nuclear Information System (INIS)

    Steinbrueck, M.; Boettcher, M.

    2011-01-01

    The paper presents the results of isothermal and transient oxidation experiments of the advanced cladding alloys M5 (registered) and ZIRLO TM in comparison to Zircaloy-4 in air at temperatures from 973 to 1853 K. Generally, oxidation in air leads to a strong degradation of the cladding material. The main mechanism of this process is the formation of zirconium nitride and its re-oxidation. From the point of view of safety, the barrier effect of the fuel cladding is lost much earlier than during accident transients with a steam atmosphere only. Comparison of the three alloys investigated reveals a qualitatively similar, but quantitatively varying oxidation behavior in air. The mainly parabolic oxidation kinetics, where applicable, is comparable for the three alloys. Strong differences of up to 500% in oxidation rates were observed after transition to linear kinetics at temperatures below 1300 K. The paper presents kinetic rate constants as well as critical times and oxide scale thicknesses at the point of transition from parabolic to linear kinetics.

  18. Influence of water chemistry on fuel cladding behaviour. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-02-01

    For the purpose of the meeting water chemistry included the actual practice, the water chemistry monitoring and the on-going research. Corrosion included also hydriding, recent observations made in reactors, modelling and the recent research carried out. Fifty seven participants representing twenty countries attended the thirty formal presentations and the subsequent discussions. The thirty papers presented were split into five sessions covering, Reactor experience, Mechanism and Modelling, Oxidation and hydriding, On-line monitoring of water chemistry and the review of existing and advanced water chemistries. Four panel discussions including ''Corrosion mechanism and Modelling'', ''Corrosion and Hydriding'', ''Plant Experience and Loop Experiments'', Water Chemistry, Current Practice and Emerging Solutions'' and ''On-line Monitoring of Water Chemistry and Corrosion'' were organized. The main points of discussion focussed on the optimization of water chemistry, the compatibility of potassium water chemistry with the utilization of Zircaloy 4 or the utilization of zirconium niobium cladding with lithium water chemistry. The effect of the fabrication route and of the cladding composition (Sn content) on the corrosion kinetics, the state of the art and the correlative gaps in cladding corrosion modelling and the recent developments of on-line monitoring of water chemistry together with examination of suitable developments, were also discussed. Refs, figs, tabs

  19. Influence of water chemistry on fuel cladding behaviour. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-02-01

    For the purpose of the meeting water chemistry included the actual practice, the water chemistry monitoring and the on-going research. Corrosion included also hydriding, recent observations made in reactors, modelling and the recent research carried out. Fifty seven participants representing twenty countries attended the thirty formal presentations and the subsequent discussions. The thirty papers presented were split into five sessions covering, Reactor experience, Mechanism and Modelling, Oxidation and hydriding, On-line monitoring of water chemistry and the review of existing and advanced water chemistries. Four panel discussions including ``Corrosion mechanism and Modelling``, ``Corrosion and Hydriding``, ``Plant Experience and Loop Experiments``, Water Chemistry, Current Practice and Emerging Solutions`` and ``On-line Monitoring of Water Chemistry and Corrosion`` were organized. The main points of discussion focussed on the optimization of water chemistry, the compatibility of potassium water chemistry with the utilization of Zircaloy 4 or the utilization of zirconium niobium cladding with lithium water chemistry. The effect of the fabrication route and of the cladding composition (Sn content) on the corrosion kinetics, the state of the art and the correlative gaps in cladding corrosion modelling and the recent developments of on-line monitoring of water chemistry together with examination of suitable developments, were also discussed. Refs, figs, tabs.

  20. Chemical interaction at the FBR cladding fuel interfaces

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  1. Estimation of niobium, tantalum and zirconium in a carbonate rich ore

    International Nuclear Information System (INIS)

    Navale, A.S.; Venkatakrishnan, R.R.; Sreenivas, T.

    2013-01-01

    Tantalum, niobium and zirconium form refractory compounds and therefore pose problems in estimation by methods like spectrophotometry, AAS etc. Their estimation is best carried out with ICP-OES as the plasma is capable of breaking the refractory compounds and releases these elements in the form of excited atoms. But another major problem is bringing these elements in solution as they require very strong reagents and tend to hydrolyze easily in solution. Complexing agents need to be added to maintain these elements in dissolved condition. The commonly employed dissolution techniques for these elements are fusion with potassium pyrosulfate (for Nb and Ta), sodium hydroxide or peroxide (for zirconium). Digestion with mineral acids in combination with hydrofluoric acid is also employed

  2. Development of metallic fuel fabrication - A study on the interdiffusion behavior between ternary metallic fuel and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Soo; Seol, Kyung Won; Shon, In Jin [Chonbuk National University, Chonju (Korea)

    1999-04-01

    To study a new ternary metallic fuel for liquid metal reactor, various U-Zr-X alloys have been made by induction melting. The specimens were prepared for thermal stability tests at 630 deg. C upto 5000 hours in order to estimate the decomposition of the lamellar structure. Interdiffusion studies were carried out at 700 deg. C for 200 hours for the diffusion couples assembled with U-Zr-X ternary fuel versus austenitic stainless steel D9 and martensitic stainless steel HT9, respectively, to investigate the fuel-cladding compatibility. The ternary alloy, especially U-Zr-Mo and U-Zr-Nb alloys showed relatively good thermal stability as long as 5000hrs at 630 deg. C. From the composition profiles of the interdiffusion study, Fe penetrated deeper to the fuel side than other cladding elements such as Ni and Cr, whereas U did to the cladding side of fuel elements in the fuel/D9 couples. On the contrary, the reaction layers of Fuel/HT9 couple were thinner than that of Fuel/D9 couples and were less affected by cladding element, which was believed to be due to Zr rich layer between the fuel-cladding interface. HT9 is considered to be superior to D9 and a favorable choice as a cladding material in terms of fuel-cladding compatibility. 21 refs., 24 figs., 7 tabs. (Author)

  3. Surface coating Zr or Zr alloy nuclear fuel elements

    International Nuclear Information System (INIS)

    Donaghy, R.E.; Sherman, A.H.

    1980-01-01

    A method is disclosed for preventing stress corrosion cracking or metal embrittlement of a zirconium or zirconium alloy container that is to be coated on the inside surface with a layer of a metal such as copper, a copper alloy, nickel, or iron and used for holding nuclear fuel material as a nuclear fuel element. The zirconium material is etched in an etchant solution, desmutted mechanically or ultrasonically, oxidized to form an oxide coating on the zirconium, cleaned in an aqueous alkaline cleaning solution, activated for electroless deposition of a metal layer and contacted with an electroless metal plating solution. This method provides a boundary layer of zirconium oxide between the zirconium container and the metal layer. (author)

  4. 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

  5. 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

  6. 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.

  7. Progress in Understanding of Fuel-Cladding Chemical interaction in Metal Fuel

    International Nuclear Information System (INIS)

    Inagaki, Okenta; Nakamura, Kinya; Ogata, Takanari

    2013-01-01

    Conclusion: Representative phases formed in FCCI were identified: • The reaction between lanthanide elements and cladding; • The reaction between U-PU-Zr and cladding (Fe). Characteristics of the wastage layer were clarified: • Time and temperature dependency of the growth ratio of the wastage layer formed by lanthanide elements; • Threshold temperature of the liquid phase formation in the reaction between U-Pu-Zr and Fe. These results are used: - as a basis for the FCCI modeling; - as a reference data in post-irradiation examination of irradiated metallic fuels

  8. 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...

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-09

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

  11. Separation of zirconium through extraction in hydrochloric medium with tri-n-octilamine and its spectrophotometric determination with chloroanilic acid

    International Nuclear Information System (INIS)

    Floh, B.; Abrao, A.; Federgruen, L.

    1976-01-01

    A procedure is outlined for the spectrophotometric determination of zirconium using its complex with chloroanilic acid in HC10 4 2M. Interfering elements like Fe, Zn, U, Cy, Cd, Sb, Co, Pb, Hg, Tl, Pt, Au, Pd, Ga, In, Mo and W are previously extracted with tri-n-octylamine 7,5%-benzene from 4 M HCL. Then, the acid content of the solution is ascertained to 10 M HCL and zirconium is extracted with the amine. Nb is a strong interference, being extracted by the amine as well as zirconium and absorbing at the same region as zirconium chloroanilate. Zirconium is stripped from the organic phase with Na 2 CO 3 . The colour development is done with chloroanilic acid in 2 M HC10 4 and the measurements at 340 nm. The method allows the determination of 5 micrograms of Zr. The work curve covers the 0.2 - 2.0 μg Zr/mL range. The procedure is being applied to the determination of zirconium in several alloys and in samples containing zinc, magnesium, iron, aluminium, uranium and thorium [pt

  12. 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

  13. 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)

  14. 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)

  15. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

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

    Science.gov (United States)

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

    2008-12-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  18. Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

    International Nuclear Information System (INIS)

    Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R.

    2017-01-01

    The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples

  19. Analysis of zirconium and nickel based alloys and zirconium oxides by relative and internal monostandard neutron activation analysis methods

    Energy Technology Data Exchange (ETDEWEB)

    Shinde, Amol D.; Acharya, Raghunath; Reddy, Annareddy V. R. [Bhabha Atomic Research Centre, Mumbai (India)

    2017-04-15

    The chemical characterization of metallic alloys and oxides is conventionally carried out by wet chemical analytical methods and/or instrumental methods. Instrumental neutron activation analysis (INAA) is capable of analyzing samples nondestructively. As a part of a chemical quality control exercise, Zircaloys 2 and 4, nimonic alloy, and zirconium oxide samples were analyzed by two INAA methods. The samples of alloys and oxides were also analyzed by inductively coupled plasma optical emission spectroscopy (ICP-OES) and direct current Arc OES methods, respectively, for quality assurance purposes. The samples are important in various fields including nuclear technology. Samples were neutron irradiated using nuclear reactors, and the radioactive assay was carried out using high-resolution gamma-ray spectrometry. Major to trace mass fractions were determined using both relative and internal monostandard (IM) NAA methods as well as OES methods. In the case of alloys, compositional analyses as well as concentrations of some trace elements were determined, whereas in the case of zirconium oxides, six trace elements were determined. For method validation, British Chemical Standard (BCS)-certified reference material 310/1 (a nimonic alloy) was analyzed using both relative INAA and IM-NAA methods. The results showed that IM-NAA and relative INAA methods can be used for nondestructive chemical quality control of alloys and oxide samples.

  20. Collaborative analysis for certification of zirconium and zirconium base alloy reference materials JAERI-Z11 to Z16

    International Nuclear Information System (INIS)

    1985-03-01

    The second Sub-Committee on Zircaloy Analysis was organized in April 1978, under the Committee on Analytical Chemistry on Nuclear Fuels and Reactor Materials, JAERI, for the renewal of zirconium and zirconium base alloy certified reference materials (CRMs). The Sub-Committee carried out collaborative analysis among 13 participating laboratories for the certification of the CRMs, JAERI-Z11 to Z18, after development, improvement and evaluation of analytical methods during the period of May 1978 to June 1982. As the result of the collaborative analysis, the certified value was given for 18 elements (Sn, Fe, Ni, Cr, B, Cd, U, Cu, Co, Mn, Pb, Al, Ti, Si, Mo, W, Hf, C) in the CRMs. The first part of this report includes general discussion, the second part principles of certification, the third part development and verification of analytical methods, and the fourth part evaluation of analytical results on 17 elements. Preparation of Z11 to Z18, and certification for carbon in JAERI-Z17 and Z18 were reported separately in JAERI-M 83-241 and M 83-035, respectively. (author)