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Sample records for zircaloy cladding oxidation

  1. 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

  2. Air Oxidation Behaviors of Zircaloy-4 Cladding During a LOCA In Spent Fuel Pool

    International Nuclear Information System (INIS)

    Bang, Je Geon; Chun, Tae Hyun; Kim, Sun Ki; Koo, Yang Hyun

    2014-01-01

    It is well known that air oxidation induces a serious degradation of the Zircaloy cladding material, compared with steam oxidation. From the oxidant point of view, in comparison with steam, chemical heat release during oxidation in air is higher by 80%, which may lead to a more rapid degradation of the Zircaloy cladding, and further evolution of the accident.. Additionally, the oxidation kinetics in air is much faster than in steam due to the formation of non-protective oxide layer. From the safety point of view, the barrier effect of the cladding against release of fission products is lost much earlier in air compared to steam. The objective of this study is to investigate the oxidation behaviors of fuel cladding in two different conditions such as isothermal and transient condition and to generate its kinetic data under an accident condition in the spent fuel pool. In this study, the oxidation behaviors and its kinetics of the Zircaloy-4 were investigated in air environment for various air ingress scenarios in the temperature range 600 .deg. C-1,400 .deg. C by thermo-gravimetric analysis. In this study, the oxidation behaviors of the Zircaloy-4 for both isothermal condition and transient condition were investigated in air environment. In comparison with isothermal condition, the retardation of oxidation rate in transient condition was observed at both 1,200 .deg. C and 1,400 .deg. C. This seems to be ascribed to the effect of thin oxide formed during a heating

  3. Review of zircaloy oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Iglesias, F.C. [Royal Military College of Canada, Kingston, Ontario (Canada); Lewis, B.J. [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2013-07-01

    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. The effect of internal clad oxidation due to Zircaloy/UO{sub 2} interaction is also discussed. Low-temperature oxidation of Zircaloy due to water-side corrosion is further described. (author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Nadeem Elahi, Waseem; Atif Rana, Muhammad

    2015-01-01

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

  6. 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

  7. A comparison of Zircaloy oxide thicknesses on Millstone-3 and North Anna-1 PWR fuel cladding

    International Nuclear Information System (INIS)

    Polley, M.V.; Evans, H.E.

    1993-08-01

    High concentrations of lithium in the coolant may enhance the corrosion rate of Zircaloy fuel cladding. In the present work, oxide thicknesses on fuel cladding from the Millstone 3 PWR were compared with those from the North Anna 1 PWR. The intention was to identify whether the higher lithium levels (up to 3.5 ppM) in the Millstone 3 primary coolant during cycles 2 and 3 led to significantly greater oxidation rates than in North Anna 1 which operated generally with lithium levels lower than 2.2 ppM. The comparisons were made by comparing the measurements with code predictions of Zircaloy oxidation in order to factor out the effect of operational variables on the oxide thicknesses achieved. Overall, Millstone 3 oxide thicknesses were found to be approximately 14% greater than North Anna 1 values. However, approximately 29% lower oxide thicknesses were found on reload Millstone 3 rods exposed to one cycle of elevated lithium chemistry than on Millstone 3 initial fuel exposed to one cycle of normal lithium chemistry during cycle 1. Furthermore, oxide thicknesses on Millstone 3 rods exposed to two cycles of elevated lithium chemistry were approximately 36% lower than on Millstone 3 rods exposed to one cycle of normal lithium chemistry plus one cycle of elevated lithium chemistry. Therefore, it cannot be concluded that elevated lithium operation in Millstone 3 led to enhanced Zircaloy fuel clad corrosion

  8. 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

  9. 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

  10. Interactions of zircaloy cladding with gallium -- 1997 status

    International Nuclear Information System (INIS)

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

    1997-11-01

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

  11. Study of the response of Zircaloy- 4 cladding to thermal shock during water quenching after double sided steam oxidation at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Sawarn, Tapan K., E-mail: sawarn@barc.gov.in; Banerjee, Suparna; Kumar, Sunil

    2016-05-15

    This study investigates the failure of embrittled Zircaloy-4 cladding in a simulated loss of coolant accident condition and correlates it with the evolved stratified microstructure. Isothermal steam oxidation of Zircaloy-4 cladding at high temperatures (900–1200 °C) with soaking periods in the range 60–900 s followed by water quenching was carried out. The combined oxide + oxygen stabilized α-Zr layer thickness and the fraction of the load bearing phase (recrystallised α-Zr grains + prior β-Zr or only prior β-Zr) of clad tube specimens were correlated with the %ECR calculated using Baker-Just equation. Average oxygen concentration of the load bearing phase corresponding to different oxidation conditions was calculated from the average microhardness using an empirical correlation. The results of these experiments are presented in this paper. Thermal shock sustainability of the clad was correlated with the %ECR, combined oxide+α-Zr(O) layer thickness, fraction of the load bearing phase and its average oxygen concentration. - Highlights: • Response of the embrittled Zircaloy-4 clad towards thermal shock, simulated under LOCA condition was investigated. • Thermal shock sustainability of the clad was correlated with its evolved stratified microstructure. • Cladding fails at %ECR value ≥ 29. • To resist the thermal shock, clad should have load bearing phase fraction > 0.44 and average oxygen concentration < 0.69 wt%.

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

    International Nuclear Information System (INIS)

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

    1998-06-01

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

  13. 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.

  14. Examination of Zircaloy-clad spent fuel after extended pool storage

    International Nuclear Information System (INIS)

    Bradley, E.R.; Bailey, W.J.; Johnson, A.B. Jr.; Lowry, L.M.

    1981-09-01

    This report presents the results from metallurgical examinations of Zircaloy-clad fuel rods from two bundles (0551 and 0074) of Shippingport PWR Core 1 blanket fuel after extended water storage. Both bundles were exposed to water in the reactor from late 1957 until discharge. The estimated average burnups were 346 GJ/kgU (4000 MWd/MTU) for bundle 0551 and 1550 GJ/kgU (18,000 MWd/MTU) for bundle 0074. Fuel rods from bundle 0551 were stored in deionized water for nearly 21 yr prior to examination in 1980, representing the world's oldest pool-stored Zircaloy-clad fuel. Bundle 0074 has been stored in deionized water since reactor discharge in 1964. Data from the current metallurgical examinations enable a direct assessment of extended pool storage effects because the metallurgical condition of similar fuel rods was investigated and documented soon after reactor discharge. Data from current and past examinations were compared, and no significant degradation of the Zircaloy cladding was indicated after almost 21 yr in water storage. The cladding dimensions and mechanical properties, fission gas release, hydrogen contents of the cladding, and external oxide film thicknesses that were measured during the current examinations were all within the range of measurements made on fuel bundles soon after reactor discharge. The appearance of the external surfaces and the microstructures of the fuel and cladding were also similar to those reported previously. In addition, no evidence of accelerated corrosion or hydride redistribution in the cladding was observed

  15. Development of advanced neutron radiography for inspection on irradiated fuels and materials (2). Observation of hydride and oxide film on zircaloy cladding by using neutron radiography

    International Nuclear Information System (INIS)

    Yasuda, Ryou; Nakata, Masahito; Mastubayashi, Masahito; Harada, Katsuya

    2001-02-01

    Neutron radiography has been used as available diagnosis method of integrity on irradiated fuels, and has not been employed for estimating hydride and oxide film, which are influenced on integrity of Zircaloy cladding. Preliminary tests for PIE were carried out to assess possibility of neutron radiography as evaluation tool for hydrided and oxide film on the cladding. In these experiments, Zircaloy claddings with controlled amount of hydrogen absorption (200, 500, and 1000ppm) and thickness of oxide film were radiographed in center axis and in side directions of cladding tube by neutron imaging plate method. It is noted that thickness of oxide film was formed range from 7 μ m to 70 μ m at various temperatures (973, 1173, and 1323K) under steam atmosphere on the Zircaloy claddings. CT (Computed Tomography) restructure calculation was carried out to obtain cross section image of the claddings non-destructively. The Radiographs were qualitatively investigated about structure formation area and dependence of hydrogen absorption amount on PSL (Photo Simulated Luminescence) and CT values using by image analysis processor. At the results of imaging plate test, obvious difference was not found out between hydride formation (except for 1000ppm cladding) and standard claddings in side direction image. However, on the center axis direction image, outer circumference in the cladding cross-section that corresponded with hydride segregation area became blacker. In the case of oxide film formed cladding images, although oxide film could not find out on all speciments in the radiographs taken at the center axis and side directions, cross-section of claddings heat-processed at 973K showed appreciable blackness increasing with oxide film thickness on the radiographs. On the other hand, there is no effective difference between images of oxide film formed claddings processed at 1173K and 1323K and that of standard cladding. In CT image of 1000ppm hydrogen absorbed cladding, it is

  16. Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Chollet, Mélanie, E-mail: melanie.chollet@psi.ch [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland); Valance, Stéphane; Abolhassani, Sousan; Stein, Gene [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland); Grolimund, Daniel [Paul Scherrer Institute, SLS, 5232 Villigen (Switzerland); Martin, Matthias; Bertsch, Johannes [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland)

    2017-05-15

    For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO{sub 2} are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components. - Highlights: •A Zircaloy-2 cladding irradiated 9 cycles was investigated thanks to synchrotron X-ray diffraction. •Microstructure and uniform strain through the oxide layer is revealed. •The m-ZrO{sub 2} uniform strain is oriented presenting compression along the (−111) plane. •Virtual tensor is built based on reflecting planes of families of grains. •Tensor components vary from tensile to compressive along the oxide layer.

  17. Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Chollet, Mélanie; Valance, Stéphane; Abolhassani, Sousan; Stein, Gene; Grolimund, Daniel; Martin, Matthias; Bertsch, Johannes

    2017-01-01

    For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO 2 are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components. - Highlights: •A Zircaloy-2 cladding irradiated 9 cycles was investigated thanks to synchrotron X-ray diffraction. •Microstructure and uniform strain through the oxide layer is revealed. •The m-ZrO 2 uniform strain is oriented presenting compression along the (−111) plane. •Virtual tensor is built based on reflecting planes of families of grains. •Tensor components vary from tensile to compressive along the oxide layer.

  18. The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Swam, L.F.P.; Shann, S.H.

    1991-01-01

    This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

  19. Zircaloy-oxidation and hydrogen-generation rates in degraded-core accident situations

    International Nuclear Information System (INIS)

    Chung, H.M.; Thomas, G.R.

    1983-02-01

    Oxidation of Zircaloy cladding is the primary source of hydrogen generated during a degraded-core accident. In this paper, reported Zircaloy oxidation rates, either measured at 1500 to 1850 0 C or extrapolated from the low-temperature data obtained at 0 C, are critically reviewed with respect to their applicability to a degraded-core accident situation in which the high-temperature fuel cladding is likely to be exposed to and oxidized in mixtures of hydrogen and depleted steam, rather than in an unlimited flux of pure steam. New results of Zircaloy oxidation measurements in various mixtures of hydrogen and steam are reported for >1500 0 C. The results show significantly smaller oxidation and, hence, hydrogen-generation rates in the mixture, compared with those obtained in pure steam. It is also shown that a significant fraction of hydrogen, generated as a result of Zircaloy oxidation, is dissolved in the cladding material itself, which prevents that portion of hydrogen from reaching the containment building space. Implications of these findings are discussed in relation to a more realistic method of quantifying the hydrogen source term for a degraded-core accident analysis

  20. Investigation of Zircaloy-2 oxidation model for SFP accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Yoshiyuki, E-mail: nemoto.yoshiyuki@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki [Global Nuclear Fuel – Japan Co., Ltd., 2-3-1, Uchikawa, Yokosuka-shi, Kanagawa, 239-0836 (Japan)

    2017-05-15

    The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. - Highlights: •An oxidation model of Zircaloy-2 in air environment was developed. •The oxidation model was validated by the comparison with oxidation tests using long cladding tubes in hypothetical spent fuel pool accident condition. •The oxidation model successfully reproduced the typical oxidation behavior in air.

  1. Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Malang, S.; Neitzel, H.J.

    1983-01-01

    Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

  2. Zircaloy cladding ID/OD oxidation studies. Final report

    International Nuclear Information System (INIS)

    Westerman, R.E.; Hesson, G.M.

    1977-11-01

    The ID/OD oxide ratio that forms on Zircaloy tubing at temperatures relevant to postulated LOCA conditions was measured as a function of time, temperature, and distance from the rupture. The average ratio at the rupture position was less than unity, and decreased with decreasing test time and increasing distance from the point of rupture. The maximum observed ID/OD oxide ratio was 1.4. Ratios in excess of unity were typically found to be a consequence of the OD oxide being thinner than would have been anticipated from the nominal test conditions. Confirmatory data were also obtained on the isothermal oxidation kinetics of Zircaloy. These data are in good agreement with those obtained by other investigators and confirm the conservative nature of the Baker-Just equation that is required for use in licensing calculations

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-12

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

  4. Investigation of the high temperature steam oxidation of Zircaloy 4 cladding tubes

    International Nuclear Information System (INIS)

    Leistikow, S.; Berg, H. v.; Kraft, R.; Pott, E.; Schanz, G.

    1979-01-01

    Also for the ORNL Zircaloy 4 cladding material, an intermediate decrease of the proportion of the ZrO 2 /α-phase layer was found, followed by an drastic increase when the breakaway of the ZrO 2 -scale occurred. Other reasons for small divergencies were evaluated, for instance temperature and time measurements, metallographic evaluation of layer thicknesses, consequences of one-sided (ORNL) and double-sided (KfK) oxidation. The so-called anomalous effect of steam oxidation during temperature transients was reproduced qualitatively and-in case that a reduced gain of oxygen was observed-explained by the predominant existence of the monoclinic oxide phase. The creep-rupture tests below 800 0 C showed a moderate prolongation of time-to-rupture when the tests were performed in steam (or after preoxidation in steam) instead of argon. Also slightly reduced maximum circumferential strain could be measured. (orig./RW) [de

  5. Study of the response of Zircaloy cladding to thermal shock during water quenching after double sided steam oxidation at elevated temperatures

    International Nuclear Information System (INIS)

    Banerjee, Suparna; Sawarn, Tapan K.; Kumar, Sunil

    2015-01-01

    This study investigates the failure of embrittled Zircaloy-4 cladding used in the present generation of Indian pressurized heavy water reactors (IPHWRs) in a simulated LOCA condition and its correlation with the evolved stratified microstructure. Isothermal steam oxidation of Zircaloy-4 cladding at high temperatures (900-1200°C) with soaking periods in the range 60-900 seconds followed by water quenching was carried out. None of the pieces broke during quenching except for those heated at 1100, 1150 and 1200°C for longer durations. The combined oxide + oxygen stabilized α-Zr(O) layer thickness and the fraction of the load bearing phase of clad tube specimens were correlated with the %ECR values calculated using Baker-Just equation. Average oxygen concentration of the load bearing prior β-Zr phase corresponding to different oxidation conditions was calculated from the average microhardness values in Vickers scale using an empirical correlation developed by Leistikow. The results of these experiments are presented in this paper. Thermal shock sustainability of the clad was correlated with the %ECR, combined oxide+α-Zr(O) layer thickness, fraction of the prior β-Zr phase and its average oxygen concentration. The thermal shock boundary was observed to be 29% ECR, 0.29 mm combined thickness of ZrO_2+α-Zr(O), 0.16 mm of β-Zr thickness with an average β phase oxygen content of 0.69 wt%. (author)

  6. Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

    International Nuclear Information System (INIS)

    Davis, W. Jr.; West, G.A.; Stacy, R.G.

    1979-01-01

    Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO 2 or 96 to 97% ThO 2 --3 to 4% UO 2 . Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO 2 or ThO 2 --UO 2 sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO 2 from BWRs and of Zircaloy-4-clad UO 2 from PWRs. Median particle sizes of UO 2 from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 μm; particle sizes of ThO 2 --UO 2 , under these same conditions, ranged from 137 to 202 μm. Similarly, median particle sizes of UO 2 from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 μm. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution deduced from experimental data, realistic estimates can be made of fractions of dislodged fuel having dimensions less than specified values

  7. Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

    Science.gov (United States)

    Yan, Yong; Qian, Shuo; Garrison, Ben; Smith, Tyler; Kim, Peter

    2018-04-01

    A nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0 wt. % at 1100 °C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness, and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.

  8. Out-of-pile experiments on the high-temperature behavior of Zircaloy-4 clad fuel rods

    International Nuclear Information System (INIS)

    Hagen, S.

    1984-01-01

    Out-of-pile experiments have been performed to investigate the escalation in temperature of Zircaloy-clad fuel rods during heatup in steam due to the exothermal Zircaloy steam reaction. In these tests single Zircaloy/uranium dioxide (UO 2 ) fuel rod simulators surrounded with a Zircaloy shroud--simulating the Zircaloy of neighboring rods--were heated inside a fiber ceramic insulation. The initial heating rates were varied from 0.3 to 2.5 K/s. In every test an escalation of the temperature rise rate was observed. The maximum measured surface temperature was about 2200 0 C. The temperature decreased after the maximum had been reached without decreasing the input electric power. The temperature decreases were due to inherent processes including the runoff of molten Zircaloy. The escalation process was influenced by the temperature behavior of the shroud, which was itself affected by the insulation and steam cooling. Damage to the fuel rods increased with increasing heatup rate. Fro slow heatup rates nearly no interaction between the oxidized cladding and UO 2 was observed, while for fast heatup rates the entire annular pellet was dissolved by molten Zircaloy

  9. Influence of texture on fracture toughness of zircaloy cladding

    International Nuclear Information System (INIS)

    Grigoriev, V.; Andersson, Stefan

    1997-06-01

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

  10. Influence of texture on fracture toughness of zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  11. Formation of Lamellar Structured Oxide Dispersion Strengthening Layers in Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

    Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures. According to our previous investigations, the tensile strength of Zircaloy-4 was increased by up to 20% with the formation of a thin dispersed oxide layer with a thickness less than 10% of that of the Zircaloy-4 substrate. However, the tensile elongation of the samples decreased drastically. The brittle fracture was a major concern in development of the ODS Zircaloy-4. In this study, a lamellar structure of ODS layer was formed to increase ductility of the ODS Zircaloy-4. The mechanical properties were varied depending on the structure of ODS layer. For example, the partial formation of ODS layer with the thickness of 10% to the substrate thickness induced the increase in tensile strength up to about 20% than fresh Zircaloy-4.

  12. 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.

  13. 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.

  14. Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

    Energy Technology Data Exchange (ETDEWEB)

    Davis, W. Jr.; West, G.A.; Stacy, R.G.

    1979-03-22

    Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO/sub 2/ or 96 to 97% ThO/sub 2/--3 to 4% UO/sub 2/. Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO/sub 2/ or ThO/sub 2/--UO/sub 2/ sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO/sub 2/ from BWRs and of Zircaloy-4-clad UO/sub 2/ from PWRs. Median particle sizes of UO/sub 2/ from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 ..mu..m; particle sizes of ThO/sub 2/--UO/sub 2/, under these same conditions, ranged from 137 to 202 ..mu..m. Similarly, median particle sizes of UO/sub 2/ from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 ..mu..m. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels; however, unirradiated fuel from vendors was not available for performing comparative shearing experiments. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution estimates can be made of fractions of dislodged fuel having

  15. Investigation of in-pile formed corrosion films on zircaloy fuel-rod claddings by impedance spectroscopy and galvanostatic anodization

    International Nuclear Information System (INIS)

    Gebhardt, O.

    1993-01-01

    Hot-cell investigations have been executed to study the corrosion behaviour of irradiated Zircaloy fuel-rod claddings by impedance spectroscopy and galvanostatic anodization. The thickness of the compact oxide at the metal/oxide interface and the thickness of the minimum barrier oxide have been determined at different positions along the claddings. As shown by analysis, both quantities first increase and then decrease with increasing thickness of the total oxide. (author) 6 figs., 33 refs

  16. 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

  17. Nondestructive characterization of hydrogen concentration in zircaloy cladding tubes with laser ultrasound technique

    International Nuclear Information System (INIS)

    Yang, Che Hua; Lai, Yu An

    2006-01-01

    This paper describes a laser ultrasound technique (LUT) for nondestructive characterization of hydrogen concentration (HC) in Zircaloy cladding tubes. With the LUT, guided ultrasonic waves are generated remotely and then propagate in the axial direction of Zircaloy tubes, and finally detected remotely by an optical probe. By measuring the dispersion spectra with the LUT, relations between the dispersion spectra and the HC of the Zircaloy tubes can be established. The LUT is non-contact, capable of remote inspection, and therefore suitable for nondestructive inspection of HC in Zircaloy cladding tubes used in nuclear power plant.

  18. Engineered zircaloy cladding modifications for improved accident tolerance of LWR fuel: US DOE NEUP Integrated Research Project

    International Nuclear Information System (INIS)

    Heuser, Brent

    2013-01-01

    An integrated research project (IRP) to fabricate and evaluate modified zircaloy LWR cladding under normal BWR/PWR operation and off-normal events has been funded by the US DOE. The IRP involves three US academic institutions, a US national laboratory, an intermediate stock industrial cladding supplier, and an international academic institution. A combination of computational and experimental protocols will be employed to design and test modified zircaloy cladding with respect to corrosion and accelerated oxide growth, the former associated with normal operation, the latter associated with steam exposure during loss of coolant accidents (LOCAs) and low-pressure core re-floods. Efforts will be made to go beyond design-base accident (DBA) scenarios (cladding temperature equal to or less than 1204 deg. C) during the experimental phase of modified zircaloy performance characterisation. The project anticipates the use of the facilities at ORNL to achieve steam exposure beyond DBA scenarios. In addition, irradiation of down-selected modified cladding candidates in the ATR may be performed. Cladding performance evaluation will be incorporated into a reactor system modelling effort of fuel performance, neutronics, and thermal hydraulics, thereby providing a holistic approach to accident-tolerant nuclear fuel. The proposed IRP brings together personnel, facilities, and capabilities across a wide range of technical areas relevant to the study of modified nuclear fuel and LWR performance during normal operation and off-normal scenarios. Two pathways towards accident-tolerant LWR fuel are envisioned, both based on the modification of existing zircaloy cladding. The first is the modification of the cladding surface by the application of a coating layer designed to shift the M + O→MO reaction away from oxide growth during steam exposure at elevated temperatures. This pathway is referred to as the 'surface coating' solution. The second is the modification of the bulk

  19. 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

  20. Reaction in air and in nitrogen of pre-oxidised Zircaloy-4 and M5™ claddings

    Energy Technology Data Exchange (ETDEWEB)

    Duriez, C., E-mail: christian.duriez@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SEREX-LE2M, Centre de Cadarache, St Paul-Lez-Durance 13115 (France); Drouan, D. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SEREX-LE2M, Centre de Cadarache, St Paul-Lez-Durance 13115 (France); Pouzadoux, G. [Université Technologique de Troyes, BP 2060, Troyes 10010 (France)

    2013-10-15

    High temperature reactivity in air and in nitrogen of pre-oxidised Zircaloy-4 and M5™ claddings has been studied by thermogravimetry. Claddings were pre-oxidised at low temperature with the aim of simulating spent fuel. Different pre-oxidation modes, inducing significant variation in the pre-oxides microstructure, were compared. The behaviour in air, investigated in the 850–1000 °C temperature range, was found to be strongly dependant on the type of pre-oxide: the compact pre-oxide formed in autoclave (at temperature, pressure, and water chemistry representative of PWR conditions) significantly slows down the degradation in air compared to the bare alloys; on the contrary, a pre-oxide formed at 500 °C at ambient pressure, either in oxygen or in steam, favours the initiation of post-breakaway type oxidation, which in air is associated with nitride formation. The behaviour in nitrogen has been investigated in the 800–1200 °C temperature range, with Zircaloy-4 pre-oxidised at 500 °C in O{sub 2}. Reactivity is low up to 1000 °C but becomes very significant at the highest temperatures investigated, 1100 and 1200 °C. Finally, cladding segments first reacted in N{sub 2} at 1100 °C, were exposed to air and show fast oxidation even at the lowest temperature investigated (600 °C)

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

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1977-06-01

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

  2. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    International Nuclear Information System (INIS)

    Chung, H. M.

    2000-01-01

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

  3. Effect of cyclic loading on the viscoplastic behaviour of Zircaloy 4 cladding tubes

    International Nuclear Information System (INIS)

    Bouffioux, P.; Gabriel, B.; Soniak, A.; Mardon, J.P.

    1995-06-01

    Most of the electricity being generated by nuclear energy load follow and remote control have become normal operating modes in the French PWR. In addition, EDF is developing a strategy of fuel sub-assembly burnup extension. Those operating conditions will lead to cyclic straining of the Zircaloy cladding tube which could induce damages. Therefore, EDF, CEA, and FRAMATOME has started a joint R and D cooperative program in order to investigate the mechanical behaviour of Zircaloy cladding tubes under cyclic loading. This paper is dealing with the effect of a pre-cyclic loading on the plasticity properties of Zircaloy 4 cladding tubes. Load controlled cyclic tests were carried out at 350 deg. C and 0.5 Hz in both axial and hoop directions. The Woehler curves were determined. Sequential tests combining pre-cyclic loading to 50 and 75 % fraction life with tension were then performed. It has ben noticed that the pre-cycling loading does not change the plastic flow curve of the Zircaloy 4 cladding tubes and therefore does not induce observable macroscopic damage. It has been concluded that a linear cumulative damage rule like ΣΔN(σ)/N r(σ) is very conservative. (author)

  4. Temperature estimates from the zircaloy oxidation kinetics in the α plus β phase region

    International Nuclear Information System (INIS)

    Olsen, C.S.

    1981-01-01

    Oxidation rates of zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near thermocouple locations

  5. Temperature estimates from the Zircaloy oxidation kinetics in the α plus β phase region

    International Nuclear Information System (INIS)

    Olsen, C.S.

    1981-01-01

    Oxidation rates of Zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of Zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near the thermocouple locations

  6. 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)

  7. 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.

  8. Zircaloy behaviour in high temperature irradiated water

    International Nuclear Information System (INIS)

    Urbanic, V.F.

    1982-04-01

    The corrosion and hydriding of Zircaloy during irradiation in high temperature water is strongly dependent on the oxygen concentration of the water. Corrosion tests in the NRX and NRU research reactors using small samples have demonstrated the importance of water chemistry in maintaining Zircaloy corrosion and hydriding within acceptable limits. Zircaloy fuel cladding develops non-uniform, patch-type oxides during irradiation in hich temperature water containing dissolved oxygen. Results from examinations of prototype fuel cladding irradiated in the research reactors are presented to show how local variations in coolant flow, fast neutron flux, metallurgical structure and surface condition can influence the onset of non-uniform corrosion under these conditions. Destructive examinations of CANDU-PHW reactor fuel cladding have emphasized the importance of good chemistry control, especially the dissolved oxygen concentration of the water. When reactor coolants are maintained under normal reducing conditions at high pH (5 to 10 cm 3 D 2 /kg D 2 O; 2 /kg D 2 O; pH > 10 with LiOD), Zircaloy cladding develops non-uniform, patch-type oxides. These patch-type oxides tend to coalesce with time to form a thick, uniform oxide layer after extended exposure. Under reducing coolant conditions, Zircaloy cladding absorbs less than 200 mg D/kg Zr (approximately 2.5 mg/dm 2 equivalent hydrogen) in about 500 days. With oxygen in the coolant, deuterium absorption is considerably less despite the significant increase in corrosion under such conditions

  9. A new method of residual stress distribution analysis for corroded Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Godlewski, J.; Cadalbert, R.

    1992-01-01

    An X-ray diffraction method of residual stress measurement is developed to determine the stress level in the metal near the metal/oxide interface of Zircaloy-4 cladding samples oxidized in steam water at 400degC under a pressure of 10.3 MPa. The stress gradient is obtained and the evolution of the average stress is determined as function of the oxidation time. The presence of tetragonal zirconia phase in quite large quantity near the metal/oxide interface could be correlated to the high stress level in the base metal, adjacent to the interface. (author)

  10. A new method for residual stress distribution - analysis of corroded zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Godlewski, J.; Cadalbert, R.

    1992-01-01

    An X-ray diffraction method for residual stress measurement is developed to determine the stress level in the metal near the metal/oxide interface of Zircaloy-4 cladding samples oxidized in steam water at 400 deg C under a pressure of 10.3 MPa. The stress gradient is obtained and the evolution of the average stress is determined as a function of the oxidation time. The presence of tetragonal zirconia phase in quite large quantity near the metal/oxide interface could be correlated to the high stress level in the base metal, adjacent to the interface. 12 refs., 5 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  12. Interim report on the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  13. Strengthening of Zircaloy-4 using Oxide Particles by Laser Beam Treatment

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    Oxide particles such as Y{sub 2}O{sub 3} and CeO{sub 2} were dispersed homogeneously in a Zircaloy-4 plate surface using an LBS method. From the tensile test at 380 .deg. C, the strength of laser ODS alloying on the Zircaloy-4 sheet was increased more than 50% when compared to the initial state of the sheet, although the ODS alloyed layer was less than 20% of the specimen thickness. This technology showed a good opportunity to increase the strength without major changes in the substrates of zirconium-based alloys. Accident tolerant fuel (ATF) cladding is being developed globally after the Fukushima accident with the demands for the nuclear fuel having higher safety at normal operation conditions as well as even in a severe accident conditions. Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures.

  14. Fabrication and use of zircaloy/tantalum-sheathed cladding thermocouples and molybdenum/rhenium-sheathed fuel centerline thermocouples

    International Nuclear Information System (INIS)

    Wilkins, S.C.; Sepold, L.K.

    1985-01-01

    The thermocouples described in this report are zircaloy/tantalum-sheathed and molybdenum/rhenium alloy-sheathed instruments intended for fuel rod cladding and fuel centerline temperature measurements, respectively. Both types incorporate beryllium oxide insulation and tungsten/rhenium alloy thermoelements. These thermocouples, operated at temperatures of 2000 0 C and above, were developed for use in the internationally sponsored Severe Fuel Damage test series in the Power Burst Facility. The fabrication steps for both thermocouple types are described in detail. A laser-welding attachment technique for the cladding-type thermocouple is presented, and experience with alternate materials for cladding and fuel therocouples is discussed

  15. The anisotropic creep behaviour of zircaloy-4 fuel cladding at 1073 K

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Bowden, J.; Shewfelt, R.S.W.

    1982-04-01

    The anisotropy coefficients (F, G and H) of Hill's equation, suitably modified for creep deformation, have been determined for Zircaloy-4 fuel cladding from steady-state creep tests at an elevated temperature. Creep specimens were subjected to both uniaxial and biaxial loads (via internal pressure) at 1073 K and the strain measured concurrently in the axial and tangential directions. It has been found that Zircaloy-4 fuel cladding is almost, but not completely, isotropic at 1073 K; the values of F, G and H are 0.57, 0.48 and 0.45 respectively

  16. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

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

  17. CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

    International Nuclear Information System (INIS)

    Haste, T.J.

    1982-07-01

    The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    The eutectic melting behavior of FeCrAl/Zircaloy-4 double wall cladding tubes was investigated by annealing at various temperatures ranging from 900 .deg. C to 1300 .deg. C. It was found that significant eutectic melting occurred after annealing at temperatures equal to or higher than 1150 .deg. C. It means that an additional diffusion barrier layer is necessary to limit the eutectic melting between FeCrAl and Zircaloy-4 alloy cladding tubes. Coating of FeCrAl layers on the Zr alloy cladding tube is being investigated for the development of accident tolerant fuel by exploiting of both the oxidation resistance of FeCrAl alloys and the neutronic advantages of Zr alloys. Coating of FeCrAl alloys on Zr alloy cladding tubes can be performed by various techniques including thermal spray, laser cladding, and co-extrusion. Son et al. also reported the fabrication of FeCrAl/Zr ally double wall cladding by the shrink fit method. For the double layered cladding tubes, the thermal expansion mismatch between the dissimilar materials, severe deformation or mechanical failure due to the evolution of thermal stresses can occur when there is a thermal cycling. In addition to the thermal stress problems, chemical compatibilities between the two different alloys should be investigated in order to check the stability and thermal margin of the double wall cladding at a high temperature. Generally, it is considered that Zr alloy cladding will maintain its mechanical integrity up to 1204 .deg. C (2200 .deg. F) to satisfy the acceptance criteria for emergency core cooling systems.

  20. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    International Nuclear Information System (INIS)

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  1. Oxidation of zircaloy-2 in high temperature steam

    International Nuclear Information System (INIS)

    Ikeda, Seiichi; Ito, Goro; Ohashi, Shigeo

    1975-01-01

    Oxidation tests were conducted for zircaloy-2 in steam at temperature ranging from 900 to 1300 0 C to clarify its oxidation kinetics as a nuclear fuel cladding materials in case of a loss-of-coolant accident. The influence of maximum temperature and heating rate of the specimen on its oxidation rate in steam was investigated. The changes in mechanical properties of the specimens after oxidation tests are also studied. The results obtained were summarized as follows: (1) The weight of the specimen after oxidation in steam increased two times as the time required to reach the maximum temperature increased from 1 to 10 mins. (2) The kinetics of oxidation of zircaloy-2 in steam were not affected by the difference in the surface condition before test such as chemical polishing or pre-oxidation in steam. (3) The dominant growth of oxide film on the surface of zircaloy-2 was observed at the initial stage of oxidation in steam. However, the thickness of oxygen-rich solid solution layer under the film increased gradually with the progress of oxidation and the ratio of oxygen in oxide to that in solid solution has a constant value of 8:2. (4) The breakaway took place only in the specimen subjected to 900 0 C repeated heating. This penomenon was caused by the local growth of the oxide below a crack of the oxide film resulting from the reheating of the specimen. (5) The results of bending tests showed that the deflection until fracture of the specimen was smaller for the one heated at a higher temperature even if the weight increase was of the same order of magnitude for both specimens. (6) It was concluded that the ductility of zircaloy-2 decreased remarkably at a heating temperature in excess of 1100 0 C for more than 5 min. (auth.)

  2. Influence of manufacturing process on the in-reactor creep anisotropy of stress-relieved Zircaloy-2 cladding

    International Nuclear Information System (INIS)

    Shann, S.H.; Van Swam, L.F.

    1995-01-01

    A procedure to determine the axial/radial and circumferential/radial contractile strain ratios (the R and P factors respectively in the Backofen-modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 cladding of BWR fuel rods, tie rods and water rods was developed and has been described previously (S.H. Shann and L.F. van Swam, Creep anisotropy of Zircaloy-2 cladding during irradiation, Trans. SMiRT-11, Vol. C, 1991). The present study employs the procedure to determine the anisotropy factors R and P for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding fabricated by various manufacturing processes. The analysis indicates that the cladding manufacturing process can have a pronounced effect on the anisotropy of irradiation-induced creep. Cladding types with identical yield and ultimate tensile strengths but fabricated by different manufacturing processes have different values of R and P during in-reactor creep. ((orig.))

  3. Studies on the Electrochemical Dissolution for the Treatment of 10 g-Scale Zircaloy-4 Cladding Hull Wastes in LiCl-KCl Molten Salts

    International Nuclear Information System (INIS)

    Lee, You Lee; Lee, Jang Hwa; Jeon, Min Ku; Kang, Kweon Ho

    2012-01-01

    The electrochemical behaviors of 10 g-scale fresh and oxidized Zircaloy-4 cladding hulls were examined in 500 degree C LiCl-KCl molten salts to confirm the feasibility of the electrorefining process for the treatment of hull wastes. In the results of measuring the potential-current response using a stainless steel basket filled with oxidized Zircaloy-4 hull specimens, the oxidation peak of Zr appears to be at -0.7 to -0.8 V vs. Ag/AgCl, which is similar to that of fresh Zircaloy-4 hulls, while the oxidation current is found to be much smaller than that of fresh Zircaloy-4 hulls. These results are congruent with the outcome of current-time curves at -0.78 V and of measuring the change in the average weight and thickness after the electrochemical dissolution process. Although the oxide layer on the surface affects the uniformity and rate of dissolution by decreasing the conductivity of Zircaloy-4 hulls, electrochemical dissolution is considered to occur owing to the defect of the surface and phase properties of the Zr oxide layer.

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

    International Nuclear Information System (INIS)

    Jernkvist, L.O.

    2003-01-01

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

  5. Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

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

    1990-12-01

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

  6. 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 −})

  7. Effect of annealing temperature on the mechanical properties of zircaloy-4 cladding

    International Nuclear Information System (INIS)

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

    1977-01-01

    The mechanical properties of zircaloy cladding materials are sensitive to those fabrication variables which have an effect on the preferred crystallographic orientation or texture of the finished tube. The effect of one such variable, the final annealing temperature, on various mechanical properties is examined using tube reduced zircaloy-4 fuel rod cladding annealed at temperatures from 905F to 1060F. This temperature range provides cladding with varying degrees of recrystallization including full recrystallization. Hoop creep characteristics of zircaloy cladding were studied as a function of the annealing temperature using closed-end internal pressurization tests at 750F and hoop stresses of 10, 15, 20 and 25 ksi. The critical annealing temperature at which a minimum creep strain occurs decreases as the applied stress increases. An additional test at 700F and 30 ksi hoop stress was conducted to demonstrate that the critical annealing temperature is essentially independent of the test temperature. Plausible explanations based on differing substructures developed in cold-worked stress-relieved material are forwarded. The effect of annealing temperature on the room temperature mechanical anisotropy parameters, R and P, was studied. R-parameters were determined from in situ transverse strain gage measurements in uniaxial tensile tests. P-parameters were calculated from uniaxial test data (R and yield stress) and hoop yield stress determined in biaxial, closed-end internal pressurization tests

  8. Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

    International Nuclear Information System (INIS)

    Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

    1983-01-01

    To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

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

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1979-02-01

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

  10. Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

    International Nuclear Information System (INIS)

    Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

    1990-01-01

    Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

  11. 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)

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

    International Nuclear Information System (INIS)

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

    1985-02-01

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

  13. 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)

  14. 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

  15. 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.

  16. Iodine stress-corrosion cracking in irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Mattas, R.F.; Yaggee, F.L.; Neimark, L.A.

    1979-01-01

    Irradiated Zircaloy cladding specimens, which had experienced fluences from 0.1 to 6 x 10 21 n/cm 2 (E>0.1 MeV), were gas-pressure tested in an iodine environment to investigate their stress-corrosion cracking (SCC) susceptibility. The test temperatures and hoop stresses ranged from 320 to 360 0 C and 150 to 500 MPa, respectively. The results indicate that irradiation, in general, increases the susceptibility of Zircaloy to iodine SCC. For specimens that experienced fluences >2 x 10 21 n/cm 2 (E>0.1 MeV), the 24-h failure stress was 177+-18 MPa, regardless of the preirradiation metallurgical condition. An analytical model for iodine SCC has been developed which agrees reasonably well with the test results

  17. Thermal creep of Zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Murty, K.L.; Clevinger, G.S.; Papazoglou, T.P.

    1977-01-01

    Data on the hoop creep characteristics of Zircaloy tubing were collected at temperatures between 600 F and 800 F, and at stress levels ranging from 10 ksi to 25 ksi using internal pressurization tests. At low driving forces, exposures as long as 2000 hours were found insufficient to establish steady state creep. The experimental data at temperatures of 650 F to 800 F correlate well with an exponential stress dependence, and the activation energy for creep was found to be in excellent agreement with that for self-diffusion. The range of stresses and temperatures is too small to study the overall effect of these variables on the activation energy for creep. The experimental steady state creep-rates and those predicted from the creep equation used agree within a factor of 1.3. These correlations imply that the mechanism for hoop creep of Zircaloy-4 cladding is characterized by an activation energy of approximately 60 kcal/mole and an activation area of about 20b 3 . In addition, the exponential stress dependence implies that the activation area for creep is stress-independent. These results suggest that the climb of edge dislocations is the rate controlling mechanism for creep of Zircaloy-4. The transient creep regime was also analysed on the premise that primary creep is directly related to the rate of dispersal of dislocation entanglements by climb. (Auth.)

  18. Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

    International Nuclear Information System (INIS)

    Niceno, Bojan; Pouchon, Manuel

    2014-01-01

    The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

  19. Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

  20. Zircaloy cladding corrosion degradation in a Tuff repository: initial experimental plan

    International Nuclear Information System (INIS)

    Smith, H.D.

    1984-07-01

    The projected environmental history of a Tuff repository sited in an unsaturated hydrologic setting is evaluated to identify the potentially most severe corrosion conditions for Zircaloy spent fuel cladding. Three distinct corrosion periods are identified over the projected history. In two of those, liquid water may be present which is believed to produce the most severe corrosive environment for Zircaloy spent fuel cladding. In the time interval 100 to 1000 years after emplacement in the repository, the most severe condition is exposure to 170 0 C water at about 100 psi in an unbreached canister. This condition will be reproduced experimentally in an autoclave. For times after 1000 years, the most severe condition is exposure to 90 0 C water that is equilibrated with the tuff and invades breached canisters. This condition will be reproduced with a water bath system

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  2. Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1980-01-01

    In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

  3. Delayed hydride cracking of Zircaloy-4 fuel cladding

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  4. The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Parsons, P.D.; Hindle, E.D.; Mann, C.A.

    1986-09-01

    The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. 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)

  7. Treatment of stainless steels and zircaloy cladding hulls

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1983-01-01

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

  10. Some observations on pitting corrosion in the zircaloy cladding of fuel pins irradiated in a PWR loop

    International Nuclear Information System (INIS)

    Linde, A. van der; Letsch, A.C.; Hornsveld, E.M.

    1978-11-01

    A three-pins, zircaloy-4 clad, sphere-pac bundle was irradiated in a 280 0 C PWR loop in the HFR at Petten during 131 effective full power days to a bundle average burnup of 0.84 % FIMA. The pins contained a mixture of 61.5 w/o of 1050 μm (U,Pu) 0 2 spheres, 18.5 w/o of 115 μm UO 2 spheres and 20.0 w/o of 2 spheres. The as-fabricated smear density of the vibratory compacted mixture was 81-85 % T.D. The pressure of the pin filling gas was 1 bar helium for pin 306 and 25 bar helium for the pins 308 and 309. The cladding was zircaloy-4 tubing, stress relieved for 4 hours at 540 0 C, with an inner diameter of 9.30 mm and a wall thickness of 0.73 mm. Exposure of the pins in the loop started in the as-pickled, degreased surface condition. The pins operated at an average heat rating of 335 W/cm and at a peak rating of 620 W/cm. The end-of -life peak rating was 425 W/cm. Unfavourable water chemistry conditions of the coolant during the last weeks of the irradition, in particular low NH 3 concentrations resulting in low pH values, caused the deposition of heavy crud layers on the pin surfaces. This crud layer caused a small cladding defect in pin 306 at the axial position of the peak heat rating. The zircaloy-4 wall failed by complete oxidation, which started at and progressed from the outer, coolant side, surface. Immediately after the detection of fission product activity in the loop water, the irradiation of the bundle was terminated. Microscopic investigations on cross sections of the pins 306 and 309 revealed the presence of oxide pits at the outer surface of the zircalloy-4 wall

  11. Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil, E-mail: hgkim@kaeri.re.kr; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun

    2015-10-15

    A 3D laser coating technology using Cr powder was developed for Zr-based alloys considering parameters such as: the laser beam power, inert gas flow, cooling of Zr-based alloys, and Cr powder control. This technology was then applied to Zr cladding tube samples to study the effect of Cr coating on the high-temperature oxidation of Zr-based alloys in a steam environment of 1200 °C for 2000s. It was revealed that the oxide layer thickness formed on the Cr-coated tube surface was about 25-times lower than that formed on a Zircaloy-4 tube surface. In addition, both the ring compression and the tensile tests were performed to evaluate the adhesion properties of the Cr-coated sample. Although some cracks were formed on the Cr-coated layer, the Cr-coated layer had not peeled off after the two tests.

  12. Oxidation behavior of fuel cladding tube in spent fuel pool accident condition

    International Nuclear Information System (INIS)

    Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Ogawa, Chihiro; Nakashima, Kazuo; Tojo, Masayuki

    2017-01-01

    In spent fuel pool (SFP) under loss-of-cooling or loss-of-coolant severe accident condition, the spent fuels will be exposed to air and heated by their own residual decay heat. Integrity of fuel cladding is crucial for SFP safety therefore study on cladding oxidation in air at high temperature is important. Zircaloy-2 (Zry2) and zircaloy-4 (Zry4) were applied for thermogravimetric analyses (TGA) in different temperatures in air at different flow rates to evaluate oxidation behavior. Oxidation rate increased with testing temperature. In a range of flow rate of air which is predictable in spent fuel lack during a hypothetical SFP accident, influence of flow rate was not clearly observed below 950degC for the Zry2, or below 1050degC for Zry4. In higher temperature, oxidation rate was higher in high rate condition, and this trend was seen clearer when temperature increased. Oxide layers were carefully examined after the TGA analyses and compared with mass gain data to investigate detail of oxidation process in air. It was revealed that the mass gain data in pre-breakaway regime reflects growth of dense oxide film on specimen surface, meanwhile in post-breakaway regime, it reflects growth of porous oxide layer beneath fracture of the dense oxide film. (author)

  13. Influence of specimen design on the ductility of zircaloy cladding: Experiment and analysis

    International Nuclear Information System (INIS)

    Bates, D. W.; Majumdar, S.; Koss, D. A.; Motta, A. T.

    1999-01-01

    In a reactivity-initiated accident (RIA), a control rod ejection or drop causes a sudden increase in reactor power, which in turn deposits a large amount of energy into the fuel. The resulting thermal expansion and fission gas release loads the cladding into the plastic regime and may cause it to fail. In order to predict cladding survivability, there has been considerable interest and effort in supplementing integral WA tests with separate-effects ring tests of cladding tubes. Such tests can give one insight into failure mechanisms and measure relevant mechanical properties (such as yield strength, uniform elongation, uniaxial stress-strain curve, etc.), for use in computer codes that attempt to predict cladding response during an RIA. The accuracy of such model predictions obviously depends on appropriate and accurate failure data. This study concerns itself with the proper development of ring tensile tests that (i) are similar to the loading conditions present in an RIA, (ii) measure the relevant mechanical properties and (iii) provide insight regarding the influence of the strain paths on the failure mechanisms present if Zircaloy cladding. Based on both experiments and computational modeling, the authors investigate the failure of Zircaloy tubing as a function of specimen geometry, and discuss the limitations of certain ring-test geometries in yielding failure ductility data that are applicable to RIA situations

  14. 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)

  15. 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)

  16. Hydraulic burst tests at elevated temperatures on Zircaloy cladding from fuel rods irradiated in the Winfrith SGHWR

    International Nuclear Information System (INIS)

    Garlick, A.; Hindmarch, P.

    1980-09-01

    Closed-end hydraulic burst tests have been carried out at 613K on lengths of cladding cut from fuel rods that had been irradiated in the SGHWR to 25 n/m 2 . The effects of reactor exposure on the mechanical properties of the Zircaloy cladding, initially in the stress-relieved and fully recrystallised conditions, have been evaluated from measurements of the 0.2% proof stress, the ultimate burst stress, the total circumferential elongation and the reduction in wall thickness at fracture. It is shown that after irradiation, the measured strength properties of stress-relieved cladding remained higher than for that in the fully recrystallised condition, although the large differences observed before irradiation were considerably reduced. The irradiation-induced increase in proof stress measured during these tests was compared with US results from uniaxial tensile tests and, after correcting for the effect of stress-ratio, it is concluded that close agreement exists between the two sets of data for Zircaloy in the fully recrystallised condition. In contrast, the agreement for stress-relieved Zircaloy is less good, although the maximum increase in proof stress after high neutron doses for this material is similar for data from the two sources. After irradiation, the ductility of fully recrystallised Zircaloy remained higher than that of stress-relieved material and there was no evidence to suggest that a serious loss of ductility had occurred for Zircaloy in either condition of heat-treatment as a result of reactor exposure. (author)

  17. Monitoring the oxidation of nuclear fuel cladding using Raman spectroscopy

    International Nuclear Information System (INIS)

    Mi, Hongyi; Mikael, Solomon; Allen, Todd; Sridharan, Kumar; Butt, Darryl; Blanchard, James P.; Ma, Zhenqiang

    2014-01-01

    In order to observe Zircaloy-4 (Zr-4) cladding oxidation within a spent fuel canister, cladding oxidized in air at 500 °C was investigated by micro-Raman spectroscopy to measure the oxide layer thickness. Systematic Raman scans were performed to study the relationship between typical Raman spectra and various oxide layer thicknesses. The thicknesses of the oxide layers developed for various exposure times were measured by cross-sectional Scanning Electron Microscopy (SEM). The results of this work reveal that each oxide layer thickness has a corresponding typical Raman spectrum. Detailed analysis suggests that the Raman scattering peaks around wave numbers of 180 cm −1 and 630 cm −1 are the best choices for accurately determining the oxide layer thickness. After Gaussian–Lorentzian deconvolution, these two peaks can be quantitatively represented by four peaks. The intensities of the deconvoluted peaks increase consistently as the oxide layer becomes thicker and sufficiently strong signals are produced, allowing one to distinguish the bare and oxidized cladding samples, as well as samples with different oxide layer thicknesses. Hence, a process that converts sample oxide layer thickness to optical signals can be achieved

  18. A new high temperature deformation model for zircaloy clad ballooning under hypothetical LOCA conditions

    International Nuclear Information System (INIS)

    Brzoska, B.; Cheliotis, G.; Kunick, A.; Senski, G.

    1977-01-01

    Assuming Zircaloy clad ballooning occurs predominantly by thermal activated secondary creep, generally a power law is applied to describe the creep rate analytically. According to Norton the creep rate is taken as a power function of the cladding hoop stress multiplied by a numerical constant which is determined by the cladding structural properties and a Boltzmann factor including the creep activation energy, the gas constant and the cladding temperature respectively. As is well known, the stress exponent is not a constant value in the total range of LOCA stresses, but increases steadily with stress. This difficulty is avoided by introducing into the Norton law a plastic flow-factor including a limiting stress, which was derived by G. Senski using plastic crack models from Dugdale and Irwin. For LOCA applications the limiting stress is identified with the burst stress, which is experimentally determined. A total number of about 280 directly heated KWU burst tests including two types of experiments: (i) controlled temperature transient tests, (ii) creep rupture tests, are used to fit the burst stress of KWU zircaloy tubes simulating the whole range of LOCA temperatur

  19. Chemical interactions between as-received and pre-oxidized Zircaloy-4 and Inconel-718 at high temperatures

    International Nuclear Information System (INIS)

    Hofmann, P.; Markiewicz, M.

    1994-06-01

    Isothermal reaction experiments were performed in the temperature range of 1000 - 1300 C in order to determine the chemical interactions between Zircaloy-4 fuel rod cladding and Inconel-718 spacer grids of Pressurized Water Reactors (PWR) under severe accident conditions. It was not possible to apply even higher temperatures since fast and complete liquefaction of the components occurred as a result of eutectic interactions during heatup. The liquid reaction products formed enhance and accelerate the degradation of the material couples and the fuel elements, respectively. Only small amounts of Inconel are necessary to liquefy large amounts of Zircaloy. Thin oxide layers on the Zircaloy surface delay the beginning of the chemical interactions with Inconel but cannot prevent them. In this work the reaction kinetics have been determined for the system: as-received and pre-oxidized Zircaloy-4/Inconel 718. The interactions can be described by parabolic rate laws; the Arrhenius equations for the various interactions are given. (orig.) [de

  20. 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

  1. 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.

  2. Zircaloy-4 and M5 high temperature oxidation and nitriding in air

    Energy Technology Data Exchange (ETDEWEB)

    Duriez, C. [Institut de Radioprotection et Surete Nucleaire, Direction de Prevention des Accidents Majeurs, Centre de Cadarache, 13115 St Paul Lez Durance (France)], E-mail: christian.duriez@irsn.fr; Dupont, T.; Schmet, B.; Enoch, F. [Universite Technologique de Troyes, BP 2060, 10010 Troyes (France)

    2008-10-15

    For the purpose of nuclear power plant severe accident analysis, degradation of Zircaloy-4 and M5 cladding tubes in air at high temperature was investigated by thermo-gravimetric analysis, in isothermal conditions, in a 600-1200 deg. C temperature range. Alloys were investigated either in a 'as received' bare state, or after steam pre-oxidation at 500 {sup o}C to simulate in-reactor corrosion. At the beginning of air exposure, the oxidation rate obeys a parabolic law, characteristic of solid-state diffusion limited regime. Parabolic rate constants compare, for Zircaloy-4 as well as for M5, with recently assessed correlations for high temperature Zircaloy-4 steam-oxidation. A thick layer of dense protective zirconia having a columnar structure forms during this diffusion-limited regime. Then, a kinetic transition (breakaway type) occurs, due to radial cracking along the columnar grain boundaries of this protective dense oxide scale. The breakaway is observed for a scale thickness that strongly increases with temperature. At the lowest temperatures, the M5 alloy appears to be breakaway-resistant, showing a delayed transition compared to Zircaloy-4. However, for both alloys, a pre-existing corrosion scale favours the transition, which occurs much earlier. The post transition kinetic regime is linear only for the lowest temperatures investigated. From 800 deg. C, a continuously accelerated regime is observed and is associated with formation of a strongly porous non-protective oxide. A mechanism of nitrogen-assisted oxide growth, involving formation and re-oxidation of ZrN particles, as well as nitrogen associated zirconia phase transformations, is proposed to be responsible for this accelerated degradation.

  3. Irradiation capsule design capable of continuously monitoring the creepdown of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Thoms, K.R.; Dodd, C.V.; van der Kaa, T.; Hobson, D.O.

    1978-01-01

    An irradiation capsule which permits continuous monitoring of the creepdown of Zircaloy tubing has been designed and given preliminary tests. This design effort is the major element of a cooperative research program between the United States Nuclear Regulatory Commission and the Netherlands Energy Research Foundation (ECN) and is a part of the NRC-sponsored Zircaloy creepdown program. The purpose of the Zircaloy creepdown program is to provide data on the deformation characteristics of Zircaloy tubes, typical of LWR fuel element cladding, under combined axial and tangential compressive stresses. These data will be used to verify and improve the material behavior codes that are used for the description of fuel pin behavior. The first capsule of this series contains a mockup test specimen which was machined with three different diameters, nominally 10.92-mm, 10.54-mm and 11.30-mm (.430-in., .415-in., and .445-in.). This test specimen can be moved axially thereby varying the lift-off and serving as a calibration device for the eddy-current deformation monitoring probes. Fabrication of this capsule has been completed and during out-or-reactor checkout we were able to obtain a resolution of better than 0.01-mm (0.0004-in.). The capsule is scheduled for installation in the HFR on February 8, 1978, for a 26 day irradiation test. The first pressurized capsule, and therefore the first one to monitor in-reactor cladding deformation, will be installed in the HFR on May 3, 1978

  4. Interaction between aluminium oxide pellets and Zircaloy tubes in steam atmospheres at temperatures above 12000C

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.

    1988-09-01

    The burnable poison rods in light water reactors (LWR) consist of Al 2 O 3 /B 4 C pellets surrounded by Zircaloy-4 cladding tubes. In the Al 2 O 3 /B 4 C pellets of a LWR rod alumina is the main constituent (98.6 wt.-%) whereas boron carbide acts as neutron absorber. Failure of the Al 2 O 3 /Zircaloy test rods started at 1350 0 C when first droplets of molten material were observed running down the test bundle forming bundle blockages upon solidification. Post test examinations revealed that the process of liquefaction was initiated by a reduction of alumina by Zircaloy resulting in a (Zr, Al, O) melt which decomposed on cooldown into two metallic phases, a (Zr, Al) alloy and oxygen-stabilized a-Zr(O). The components of an extremely porous ceramic melt were also Zr, Al, and oxygen but with a higher oxygen content compared to the metallic melt. The ceramic melt decomposes on cooldown into an Al 2 O 3 /ZrO 2 eutectic with various amounts of primary constituents. Other types of relocated material were due to melting of essentially unreacted Zircaloy cladding and to debris formation by fracturing of oxidized cladding and Al 2 O 3 pellets stack residues. The interactions between Al 2 O 3 and Zircaloy occurring in a burnable poison rod are furthermore important for the behavior of the entire LWR core because the generated metals are able to attack the UO 2 chemically and dissolve or liquefy the fuel even below the melting point of Zircaloy (1760 0 C). As a result, fuel elements which contain burnable poison rods are expected to fail under severe accident conditions at about 1500 0 C. (orig./HP) [de

  5. Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes

    Directory of Open Access Journals (Sweden)

    Yang-Il Jung

    2018-03-01

    Full Text Available An oxide-dispersion-strengthened (ODS layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide (Y2O3–coated Zircaloy-4 tube to induce the penetration of Y2O3 particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at 380°C, and from 385 to 470 MPa at 500°C. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to 830°C at a heating rate of 5°C/s and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties. Keywords: Laser Surface Treatment, Microstructure, Oxide Dispersion Strengthened Alloy, Tensile Strength, Zirconium Alloy

  6. Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

    International Nuclear Information System (INIS)

    Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho

    2016-01-01

    Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

  7. Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

  8. Out-of pile mechanical test: simulating reactivity initiated accident (RIA) of zircaloy-4 cladding tube

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myung Ho; Kim, Jun Hwan; Choi, Byoung Kwon; Jeong, Young Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    The ejection or drop of a control rod in a reactivity initiated accident (RIA) causes a sudden increase in reactor power and in turn deposits a large amount of energy into the fuel. In a RIA, cladding tubes bear thermal expansion due to sudden reactivity and may fail from the resulting mechanical damage. Thus, RIA can be one of the safety margin reducers because the oxide on the tubes makes their thickness to support the load less as well as hydrides from the corrosion reduce the ductility of the tubes. In a RIA, the peak of reactor power from reactivity change is about 0.1m second and the temperature of the cladding tubes increases up to 1000 .deg. C in several seconds. Although it is hard to fully simulate the situation, several attempts to measure the change of mechanical properties under a RIA situation has done using a reduction coil, ring tension tests with high speed. This research was done to see the effect of oxide on the change of circumferential strength and ductility of Zircaloy-4 tubes in a RIA. The ring stretch tensile tests were performed with the strain rate of 1/sec and 0.01/s to simulate a transient of the cladding tube under a RIA. Since the test results of the ring tensile test are very sensitive to the lubricant, the tests were also carried out to select a suitable lubricant before the test of oxided specimens.

  9. Statistics of the acoustic emission signals parameters from Zircaloy-4 fuel cladding

    International Nuclear Information System (INIS)

    Oliveto, Maria E.; Lopez Pumarega, Maria I.; Ruzzante, Jose E.

    2000-01-01

    Statistic analysis of acoustic emission signals parameters: amplitude, duration and risetime was carried out. CANDU type Zircaloy-4 fuel claddings were pressurized up to rupture, one set of five normal pieces and six with defects included, acoustic emission was used on-line. Amplitude and duration frequency distributions were fitted with lognormal distribution functions, and risetime with an exponential one. Using analysis of variance, acoustic emission was appropriated to distinguish between defective and non-defective subsets. Clusters analysis applied on mean values of acoustic emission signal parameters were not effective to distinguish two sets of fuel claddings studied. (author)

  10. The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

  11. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs

  12. Experimental study of the deformation of Zircaloy PWR fuel rod cladding under mainly convective cooling

    Energy Technology Data Exchange (ETDEWEB)

    Hindle, E.D.; Mann, C.A.

    1982-01-01

    Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 915 degree C in flowing steam at atmospheric pressure. Internal test pressures were in the range 0.69 to 11.0 MPa. The length of cladding strained 33 percent or more was greatest (about 20 times the original diameter) when the initial pressure was 1.38/plus or minus/0.17MPa. This results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilizing the deformation or partial superplastic deformation, or both. For adjacent rods in a fuel assembly not to touch at any temperature, the pressure would have to be less than about 1 MPa. These results are compared with those form multirod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behavior of fuel elements in a loss-of-coolant accident are outlined. 37 refs.

  13. For the world's best cladding tubes, ten years of progress by Zircaloy Special Committee of JAPCO

    International Nuclear Information System (INIS)

    Mishima, Yoshitsugu

    1982-01-01

    The zircaloy special committee was organized in 1971 for the purpose of planning the trial use of two nuclear fuel assemblies for which Japan-made cladding tubes were to be used, for a BWR. Now, seven years later, these two fuel assemblies have completed their service life, and have been submitted to post-irradiation examination after cooling for a year. Zircaloy tubes have been produced by Sumitomo Metal Industries, Ltd., and Kobe Steel, Ltd., and more than ten years have elapsed since wholly Japan-made zircaloy cladding tubes were used for reloading fuel elements for the Japan Power Demonstration Reactor. In this report, the history, progress and significance of the works performed by the committee are summarized. The LWR fuel elements made in Japan have attained the highest performance in the world as the leak has been scarce, and the works of the committee is one of the pioneering activities in the development of LWR fuel technology. The situation for starting the committee, the activity of the committee during ten years, the significance and outcome of the committee activity are reported. (Kako, I.)

  14. Influence of specimen design on the deformation and failure of zircaloy cladding

    International Nuclear Information System (INIS)

    Bates, D.W.; Koss, D.A.; Motta, A.T.; Majumdar, S.

    2000-01-01

    Experimental as well as computational analyses have been used to examine the deformation and failure behavior of ring-stretch specimens of Zircaloy-4 cladding tubes. The results show that, at least for plastically anisotropic unirradiated cladding, specimens with a small gauge length l to width w ratio (l/w ∼ 1) exhibit pronounced non-uniform deformation along their length. As a result, specimen necking occurs upon yielding when the specimen is fully plastic. Finite element analysis indicates a minimum l/w of 4 before a significant fraction of the gauge length deforms homogeneously. A brief examination of the contrasting deformation and failure behavior between uniaxial and plane-strain ring tension tests further supports the use of the latter geometry for determining cladding failure ductility data that are relevant to certain reactivity-initiated accident conditions

  15. 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.

  16. Pressure effects on high temperature steam oxidation of Zircaloy-4

    International Nuclear Information System (INIS)

    Park, Kwangheon; Kim, Kwangpyo; Ryu, Taegeun

    2000-01-01

    The pressure effects on Zircaloy-4 (Zry-4) cladding in high temperature steam have been analyzed. A double layer autoclave was made for the high pressure, high temperature oxidation tests. The experimental test temperature range was 700 - 900 deg C, and pressures were 0.1 - 15 MPa. Steam partial pressure turns out to be an important one rather than total pressure. Steam pressure enhances the oxidation rate of Zry-4 exponentially. The enhancement depends on the temperature, and the maximum exists between 750 - 800 deg C. Pre-existing oxide layer decreases the enhancement about 40 - 60%. The acceleration of oxidation rate by high pressure team seems to be originated from the formation of cracks by abrupt transformation of tetragonal phase in oxide, where the un-stability of tetragonal phase comes from the reduction of surface energy by steam. (author)

  17. Oxidation of Zircaloy-4 under limited steam supply at 1000 and 13000C

    International Nuclear Information System (INIS)

    Uetsuka, H.

    1984-12-01

    With the view of examining the oxidation behavior of Zircaloy-4 under limited steam supply occurring in severe accidents of LWRs, Zircaloy-4 cladding specimens were examined at the isothermal oxidation temperatures of 1000 and 1300 0 C under a steam atmosphere, flowing at a reduced and constant rate in the range of 3proportional170 mg/cm 2 xmin. The effect of steam starvation, which was restricted to very low levels of steam supply rate, was observed at the two examined temperatures. And the critical supply rate of steam starvation was evaluated to be 13 and 20 mg/cm 2 xmin for the oxidation at 1000 and 1300 0 C, respectively. Variation of the oxidation duration between 2 and 60 min at 1000 0 C allowed to compare the reaction kinetics for three different rates of steam supply. The short-term results confirmed the reduced reaction rates for the lower steam supplies. At the longer times, however, a clear trend towards linear kinetics was observed for the lower supplies. This can be interpreted as the result of earlier breakaway transition under limited steam supply. In the test at 1300 0 C, an acceleration of the oxidation rate was measured for the specified steam supply rate between 20 and 60 mg/cm 2 xmin. This related strongly with high hydrogen concentration in the atmosphere. Hydrogen blanketing - the retarding effect of hydrogen on Zircaloy oxidation - was not identified in the examined temperature range. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  19. A unified model to describe the anisotropic viscoplastic behavior of Zircaloy-4 cladding tubes

    International Nuclear Information System (INIS)

    Delobelle, P.; Robinet, P.; Bouffioux, P.; Geyer, P.; Pichon, I. Le

    1996-01-01

    This paper presents the constitutive equations of a unified viscoplastic model and its validation with experimental data. The mechanical tests were carried out in a temperature range of 20 to 400 C on both cold-worked stress-relieved and fully annealed Zircaloy-4 tubes. Although their geometry (14.3 by 1.2 mm) is different, the crystallographic texture was close to that expected in the cladding tubes. To characterize the anisotropy, mechanical tests were performed under both monotonic and cyclic uni- and bi-directional loadings, i.e., tension-compression, tension-torsion, and tension-internal pressure tests. The results obtained at ambient temperatures and the independence of the ratio R p = var-epsilon θθ p /var-epsilon zz p , with respect to temperature would seem to indicate that the set of anisotropy coefficients does not depend on temperature. Zircaloy-4 material also has a slight supplementary hardening during out-of-phase cyclic loading. The authors propose to extend the formulation of a unified viscoplastic model, developed and identified elsewhere for other initially isotropic materials, to the case of Zircaloy-4. Generally speaking, anisotropy is introduced through fourth order tensors affecting the flow directions, the linear kinematical hardening components, as well as the dynamic and static recoveries of the forementioned hardening variables. The ability of the model to describe all the mechanical properties of the material is shown. The application of the model to simulate mechanical tests (tension, creep, and relaxation) performed on true CWSR Zircaloy-4 cladding tubes with low tin content is also presented

  20. High-temperature oxidation kinetics of sponge-based E110 cladding alloy

    Science.gov (United States)

    Yan, Yong; Garrison, Benton E.; Howell, Mike; Bell, Gary L.

    2018-02-01

    Two-sided oxidation experiments were recently conducted at 900°C-1200 °C in flowing steam with samples of sponge-based Zr-1Nb alloy E110. Although the old electrolytic E110 tubing exhibited a high degree of susceptibility to nodular corrosion and experienced breakaway oxidation rates in a relatively short time, the new sponge-based E110 demonstrated steam oxidation behavior comparable to Zircaloy-4. Sample weight gain and oxide layer thickness measurements were performed on oxidized E110 specimens and compared to oxygen pickup and oxide layer thickness calculations using the Cathcart-Pawel correlation. Our study shows that the sponge-based E110 follows the parabolic law at temperatures above 1015 °C. At or below 1015 °C, the oxidation rate was very low when compared to Zircaloy-4 and can be represented by a cubic expression. No breakaway oxidation was observed at 1000 °C for oxidation times up to 10,000 s. Arrhenius expressions are given to describe the parabolic rate constants at temperatures above 1015 °C and cubic rate constants are provided for temperatures below 1015 °C. The weight gains calculated by our equations are in excellent agreement with the measured sample weight gains at all test temperatures. In addition to the as-fabricated E110 cladding sample, prehydrided E110 cladding with hydrogen concentrations in the 100-150 wppm range was also investigated. The effect of hydrogen content on sponge-based E110 oxidation kinetics was minimal. No significant difference was found between as-fabricated and hydrided samples with regard to oxygen pickup and oxide layer thickness for hydrogen contents below 150 wppm.

  1. Determination of the initial oxidation behavior of Zircaloy-4 by in-situ TEM

    International Nuclear Information System (INIS)

    Harlow, Wayne; Ghassemi, Hessam; Taheri, Mitra L.

    2016-01-01

    The corrosion behavior of Zircaloy-4 (Zry-4), specifically by oxidation, is a problem of great importance as this material is critical for current nuclear reactor cladding. The early formation behavior and structure of the oxide layer during oxidation was studied using in-situ TEM techniques that allowed for Zry-4 to be monitored during corrosion. These environmental exposure experiments were coupled with precession electron diffraction to identify and quantify the phases present in the samples before and after the oxidation. Following short-term, high temperature oxidation, the dominant phase was revealed to be monoclinic ZrO 2 in a columnar structure. These samples oxidized in-situ contained structures that correlated well with bulk Zry-4 subjected to autoclave treatment, which were used for comparison and validation of this technique. By using in-situ TEM the effect of microstructure features, such as grain boundaries, on oxidation behavior of an alloy can be studied. The technique presented herein holds the potential to be applied any alloy system to study these effects. - Highlights: • In-situ TEM was used to oxidize samples of Zircaloy-4. • Similar behavior was found in the in-situ oxidized and autoclave-oxidized samples. • Precession diffraction was used to characterize oxide phase and texture.

  2. 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.

  3. Comparison between zircaloy oxidation in steam and air surroundings

    International Nuclear Information System (INIS)

    Shawkat, M.E.; Hasaneln, H.; Ali, M.; Parlatan, Y.; Albasha, H.

    2013-01-01

    The available experimental data for Zircaloy oxidation in air were reviewed. The behavior of the oxidation kinetics at different temperature ranges was described. It was shown that maintaining the oxidation kinetics within the oxide pre-breakaway region can prevent elevated sheath temperatures due to the oxidation process during postulated accidents. The available correlations to model the oxidation kinetics for pre-breakaway region were reviewed and assessed. Zircaloy-air oxidation correlation based on Leistikow-Berg data was determined to be the most suitable correlation to model pre-breakaway kinetics and it was compared to Urbanic-Heidrick correlation which is widely used for Zircaloy oxidation in steam environment. The results showed that the energy release due to the Zircaloy-steam oxidation bounds the energy released due to Zircaloy-air oxidation up to a sheath temperature referred as the “crossover temperature”. Below this temperature, the impact of Zircaloy-air oxidation on fuel sheath temperature transient can be predicted conservatively using the Urbanic-Heidrick steam correlation. The crossover temperature was calculated for isothermal sheath heating as well as transient sheath heat-up assuming three linear heating rates of 0.6, 1.0, and 1.3 K/s. (author)

  4. Cladding the inside surface of a 3 1/4 in. ID Zircaloy-2 pressure tube with 1S aluminum

    International Nuclear Information System (INIS)

    Watson, R.D.

    1966-09-01

    A hot-press sizing technique has been developed for cladding the inside surface of Zircaloy-2 pressure tubes with 1S aluminum. The process is performed in air with the Zircaloy-2 and aluminum at a temperature of approximately 950 o F. A controlled atmosphere is not required, either during preheating or while the cladding is being applied. Tubes 30 inches long and 3 1/4 inches ID have been coated with 1S aluminum in thicknesses ranging from 0.005 inches to more than 0.02 inches; tubes longer than 30 inches have not been attempted. The lining of aluminum is firmly attached to the Zircaloy-2 at all points in the tube but the bond strength varies considerably - from. 6500 to 28000 lbf/in 2 . This work is the subject of Canadian Patent Application No. 955,358 filed March 21, 1966. (author)

  5. 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.

  6. Investigations of the interaction between ballooning Zircaloy cladding and emergency core cooling

    International Nuclear Information System (INIS)

    Wiehr, K.; Barth, S.; Erbacher, F.; Hame, W.; Harten, U.; Just, W.; Megerle, A.; Mueller, S.; Neitzel, H.J.; Reimann; Schaeffner, P.; Schmidt, H.

    1975-01-01

    The development of fabrication methods for the production of fuel rod simulators has been largely terminated. For welding of Zircaloy-4 and inconel 600 explosive welding has proved to be promissory in preliminary tests. A prototype fuel rod simulator was tested at full power. Its performance was faultless and the fuel rod and ring pellets could be easily dismantled and reused after the experiment. Planning of the test rig and electricity supply were terminated. Most of the assembly work has been finished. For electric heating of the fuel rod simulators a special device was built and tested which allows to program the power control. The radiographic system recording ballooning of the Zircaloy clad was erected outside the test space and put into operation. First trial pictures yielded good results. (orig.) [de

  7. High temperature properties of Zircaloy--oxygen alloys

    International Nuclear Information System (INIS)

    Mellinger, G.B.; Bates, J.L.

    1977-03-01

    The effect of oxygen on three properties of Zircaloy-4 cladding relevant to LOCA evaluation codes was determined. Thermal expansion, elastic moduli, and thermal diffusivity were measured over the range room temperature--1200 0 C (2192 0 F) and 0.7 to 28 at.% oxygen. Thermal expansion and elastic moduli showed increases with oxygen concentration, while thermal diffusivity tended to decrease. Zircaloy-2 was examined over the same temperature range, but only to 5 at.% oxygen, differences in the properties between the two alloys were minor. The thermal emittance of Zircaloy-4 was measured in argon over the wavelength range 1.5 to 2.5 μm on previously oxidized tubing and on surfaces in the process of oxidizing in unlimited steam. For the latter, a high emittance (approximately 0.9) was reached at an oxide thickness of about 100 mg/dm 2 , and the tubing surface remained black and substoichiometric as oxidation continued at temperatures to 1200 0 C

  8. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Peck, S.O.; Wallenfels, K.P.

    1983-12-01

    This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO 2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO 2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 2 0 C/s at 1100 0 C increased to approximately 6 0 C/s. The maximum temperature reached was 2250 0 C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO 2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO 2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.) [de

  9. High-temperature oxidation of Zircaloy in hydrogen-steam mixtures

    International Nuclear Information System (INIS)

    Chung, H.M.; Thomas, G.R.

    1982-09-01

    Oxidation rates of Zircaloy-4 cladding tubes have been measured in hydrogen-steam mixtures at 1200 to 1700 0 C. For a given isothermal oxidation temperature, the oxide layer thicknesses have been measured as a function of time, steam supply rate, and hydrogen overpressure. The oxidation rates in the mixtures were compared with similar data obtained in pure steam and helium-steam environments under otherwise identical conditions. The rates in pure steam and helium-steam mixtures were equivalent and comparable to the parabolic rates obtained under steam-saturated conditions and reported in the literature. However, when the helium was replaced with hydrogen of equivalent partial pressure, a significantly smaller oxidation rate was observed. For high steam-supply rates, the oxidation kinetics in a hydrogen-steam mixture were parabolic, but the rate was smaller than for pure steam or helium-steam mixtures. Under otherwise identical conditions, the ratio of the parabolic rate for hydrogen-steam to that for pure steam decreased with increasing temperature and decreasing steam-supply rate

  10. The effect of mechanical restraint on the deformation of Zircaloy cladding

    International Nuclear Information System (INIS)

    Jones, P.M.; Haste, T.J.

    1980-10-01

    Zircaloy cladding, deformed at temperatures postulated for loss-of-coolant accidents, can exhibit considerable ductility. The actual circumferential strain is governed by the temperature uniformity around the rod during the time at which the major part of the deformation occurs. If the bulges in neighbouring rods in a multi-rod array touch before rupture, and the array is large enough for the outer rods to restrain bulges rather than be pushed away by them, then the stress in such bulges drops. However the stress in adjacent axial regions of the cladding which have not contacted remains high and these continue to strain until they also interact, thus propagating the bulging axially. Meanwhile the non-contacted portions of the interacting bulges continue to strain slowly into the remaining sub-channels. Illustrative calculations suggest that the mechanical restraint of bulging cladding will only be effective in increasing sub-channel blockage when the failure strains are greater than 60-70%. This may occur with temperature differences between neighbouring rods of 10-25 0 C if the deformation process is thermally stabilised. (author)

  11. Irradiation effect on fatigue behaviour of zircaloy-4 cladding tubes

    International Nuclear Information System (INIS)

    Soniak, A.; Lansiart, S.; Royer, J.; Waeckel, N.

    1993-01-01

    Since nuclear electricity has a predominant share in French generating capacity, PWR's are required to fit grid load following and frequency control operating conditions. Consequently cyclic stresses appear in the fuel element cladding. In order to characterize the possible resulting clad damage, fatigue tests were performed at 350 deg C on unirradiated material or irradiated stress relieved Zircaloy-4 tube portions, using a special device for tube fatigue by repeated pressurization. It appears that, for high stress levels, the material fatigue life is not affected by irradiation. But the endurance fatigue limit undergoes a decrease from the 350 MPa value for unirradiated material to the 210 MPa value for the material irradiated for four cycles in a PWR. However, this effect seems to saturate with irradiation dose: no difference could be detected between the two cycles results and the corresponding four cycles results. The corrosion effect and the load following influence were also investigated: they do not appear to modify the fatigue behaviour in our experimental conditions

  12. The steady-state creep of zircaloy-4 fuel cladding from 940 to 1873 K

    International Nuclear Information System (INIS)

    Rosinger, H.E.; Bera, P.C.; Clendening, W.R.

    1978-11-01

    The steady-state creep rates of as-received Zircaloy-4 fuel cladding have been determined in the α-Zr phase (940 -6 and 10 -3 s -1 were determined under constant uniaxial load conditions. Assuming that creep rates can be described by a power law - Arrhenius equation, the creep rate for α-phase Zircaloy-4 is given by: epsilon sub(ss) = 2000σ sup(5.32) exp (-284 600/kT) s -1 and for the β-phase Zircaloy-4 is given by: epsilon sub(ss) = 8.1σ sup(3.79) exp (-142 300/kT) s -1 . For both the α-Zr and β-Zr phases, the activation energies for creep are in agreement with those for self-diffusion of zirconium and the rate-controlling mechanism is attributed to dislocation climb. Because of the scarcity of data, it is not possible to determine the rate equation unambiguously, nor to identify the mechanism for creep in the mixed α + β phase region. (author)

  13. High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments

    International Nuclear Information System (INIS)

    Chung, H.M.; Garde, A.M.; Kassner, T.F.

    1977-01-01

    The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350 0 C

  14. Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

    International Nuclear Information System (INIS)

    Meservey, R.H.

    1976-04-01

    The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800 0 F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests

  15. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    International Nuclear Information System (INIS)

    Clayton, J.C.

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  17. 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

  18. The deformation of Zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-08-01

    Simulated PWR fuel rods clad with Zircaloy-4 were tested under convective steam cooling conditions, by pressurising to 0.69-2.07MPa (100-300lb/in 2 ), then ramping at 10 0 C/s to various temperatures in the region 800-955 0 C and holding until either 600 s elapsed or rupture occurred. The length of cladding strained 33% or more was greatest (about 20 times the original diameter) when the initial internal pressure was 1.38+-0.17 PMa (200+-25lb/in 2 ), and the temperature 885 0 C. It is thought that this results from oxidation strengthening of the surface layers acting as an additional mechanism for stabilising the deformation and/or partial superplastic deformation. To avoid adjacent rods in a fuel assembly touching at any temperature, the pressure would have to be less than about 1MPa (145 1b/in 2 ). If the pressure was 1.38MPa (200lb/in 2 ) then the rods would not swell sufficiently to touch if the temperature did not exceed about 840 0 C. (author)

  19. Crack resistance curve determination of zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Bertsch, J.; Alam, A.; Zubler, R.

    2009-03-01

    Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 o C and 350 o C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could be

  20. Crack resistance curve determination of zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Bertsch, J.; Alam, A.; Zubler, R

    2009-03-15

    Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 {sup o}C and 350 {sup o}C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could

  1. Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

    International Nuclear Information System (INIS)

    Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

    1979-01-01

    Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

  2. Creep behavior of Zircaloy cladding under variable conditions

    International Nuclear Information System (INIS)

    Matsuo, Y.

    1989-01-01

    Various creep tests of Zircaloy cladding tubes under variable conditions were conducted to investigate which hardening rule can be applicable for the creep behavior associated with condition changes. The results show that the strain-hardening rule is applicable in general when either the stress or temperature conditions change, provided that a certain amount of creep strain recovery is observed in case of stress drop. In stress reversal conditions, however, softening of the material was observed. Strain rate after stress reversal is much higher than that predicted by the strain-hardening rule. In this case, the modified strain-hardening model, considering a recoverable creep-hardening range together with the strain recovery, predicts the creep behavior well. The applicability of the model is ascertained through a verification test that includes stress reversal, strain recovery, stress changes, and temperature changes

  3. Mechanical properties of irradiated and non-irradiated Zr1%Nb and Zircaloy claddings

    International Nuclear Information System (INIS)

    Griger, Agnes

    2004-01-01

    The mechanical properties of irradiated and non-irradiated Zr1%Nb were determined and they were compared with the analogous properties of Zircaloy-4 to establish connections between the evolution of mechanical parameters of Zr1%Nb and Zircaloy-4 cladding materials and the measure of irradiation. Samples were irradiated in the vertical channels of the Budapest Research Reactor for different time periods at 50-65 C temperature. The measure of irradiation (fluent) for different samples was estimated by means of flux measurement and using the effective irradiation time. Post irradiation uniaxial tension tests in transverse direction were carried out on ring specimens. The mechanical parameters of the Zr1%Nb alloy significantly improve due to the effect of irradiation. However, the values of mechanical parameters do not further increase when the fluent increases above 10 20 n/cm 2 . These results are in good accordance with the Russian ones [1]. Contrary to the behaviour of Zr1%Nb alloy, the mechanical parameters of the Zircaloy practically do not change on the effect of irradiation. The originally high values of ultimate tensile strength and yield stress change only slightly with the increasing fluent in the investigated fluent-region. (Author)

  4. Mechanistic considerations used in the development of the probability of failure in transient increases in power (PROFIT) pellet-zircaloy cladding (thermo-mechanical-chemical) interactions (pci) fuel failure model

    International Nuclear Information System (INIS)

    Pankaskie, P.J.

    1980-05-01

    A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) interactions (PCI) failure model for estimating the Probability of Failure in Transient Increases in Power (PROFIT) was developed. PROFIT is based on (1) standard statistical methods applied to available PCI fuel failure data and (2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmental and strain-rate dependent Strain Energy Absorption to Failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-dislocation interaction effects in the Zircaloy cladding

  5. Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction ESSI-4 ESSI-11

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauscheck, H.; Wallenfels, K.P.; Buescher, B.J.

    1985-03-01

    The tests had the initial heatup rate as main parameter. The experimental arrangement consisted of a fuel rod simulator (central tungsten heater, UO 2 ring pellets and zircaloy cladding), a zircaloy shroud and the fiber ceramic insulation. A steam flow of ca. 20 g/min was introduced at the lower end of the bundle. A temperature escalation was observed in every test. The maximum cladding surface temperature in the single rod tests never exceeded 2200 0 C. The escalation began in the upper region of the rods and moved down the rods, opposite to the direction of steam flow. For fast initial heatup rates, the runoff of molten zircaloy was a limiting process for the escalation. For slow heatup rates, the formation of a protective oxide layer reduced the reaction rate. The test with less insulation thickness showed a reduction of the escalation. A stronger influence was found for the gap between shroud and insulation. This is caused by convection heat losses to the steam circulating in this gap by natural convection. Removal of the gap between shroud and insulation in essentially the same experimental arrangement produced a faster escalation. The posttest appearance of the fuel rod simulators showed that, at slow heatup rates oxidation of the cladding was complete, and the fuel rod was relatively intact. Conversely, at fast heatup rates, relatively little cladding oxidation with extensive dissolution of the UO 2 pellets and runoff of molten cladding was observed. (orig./HP) [de

  6. 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

  7. Study of hydrogen migration produced during the corrosion of PWR reactors fuel cans in zircaloy 4 and zirconia; Etude du transport de l`hydrogene produit lors de la corrosion des gaines d`elements combustibles des reacteurs a eau sous pression dans la zircone et le zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Aufore, L

    1997-12-12

    The corrosion of Zircaloy-4-claddings by water from the primary circuit of nuclear power plant goes hand in hand with the release of hydrogen which penetrates the oxide and then the metal. This work focuses on the mechanisms of hydrogen transport in oxide and in metal. Hydrogen transport in oxide is studied on the basis of corrosion tests performed in the autoclave at 360 deg C. These tests are performed on Zircaloy-4 claddings under different chemical conditions (pure water, and pure water with lithium hydroxide). The distribution of hydrogen in oxide film is measured by SIMS. Hydrogen profiles in the oxide are dependent on the oxide microstructure and vary with oxidation time. These observations are confirmed by experiments in which some samples, previously oxidized in the autoclave, are immersed in heavy water. In the oxide layer, two zones are observed: one external porous zone and one internal dense zone. Deuterium diffusion coefficients in dense oxide are determined using SIMS profiles and Fischer diffusion model. Hydrogen transport in metal is also studied by means of gas-phase permeation experiments. These are set up at different temperature (400-500 deg. C) and under different hydrogen pressures and make it possible to determine the hydrogen diffusion coefficients in a Zircaloy-4 cladding under experimental conditions. All these results lead us to discuss of hydrogen transport evolution in cladding during oxidation. A model taking into account hydrogen transport in oxide and in metal, and the hydrides precipitations is proposed. (author) 110 refs.

  8. 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

  9. Morphology control of anodic ZrO2 layer for the prevention of H2 production from Zr-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Y. J.; Park, J. W.; Cho, S. O. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    Since the Fukushima disaster happened, studies on accident-resistant nuclear fuel has been carried out actively. There has been an attempt to protect zircaloy fuel cladding by coating SiC. Research on producing oxide layer that can block fuel cladding from water on the surface of zircaloy fuel cladding by means of anodizing to reduce the rate of oxidation of fuel cladding at Loss Of Coolant Accident (LOCA) is an significant ongoing study subject. Applying nanostructured oxide layer to the prevention of thermal deformation of oxide layer was already suggested in our research group, the reasons of which is nanoporous structure is better than nanotube structure in terms of corrosion-resistant structure because nanotube structure can be easily peeled off. In this study, methods which are able to control morphology between nanoporous and nanotube structure were conducted by changing the anodizing conditions. Hence, Using glycerol and ammonium fluoride, Zircaloy-4 was anodized by varying water contents and applied voltage. It reveals that the alloy transition from nanoporous structure to nanotube structure can be changed by varying water contents of anodizing solution and applied voltage. Anodizing conditions determining nanoporous structure were obtained. According to the mechanism already suggested, nanoporous oxide layer that can seal the fuel cladding perfectly, and increase critical heat flux (CHF) due to large surface area is easily produced. This results obtained in this paper expected to be facilitated fabrication of accident-resistant nuclear fuel cladding.

  10. Secondary hydriding of defected zircaloy-clad fuel rods

    International Nuclear Information System (INIS)

    Olander, D.R.; Vaknin, S.

    1993-01-01

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

  11. High-temperature oxidation of Zircaloy-2 and Zircaloy-4 in steam

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Heidrick, T.R.

    1978-01-01

    At temperatures above the (α + β)/β transformation temperature for zirconium alloys, steam reacts with β-Zr to form a superficial layer of zirconium oxide (ZrO 2 ) and an intermediate layer of oxygen-stabilized α-Zr. Reaction kinetics and the rate of growth of the combined (ZrO 2 + α-Zr) layer for Zircaloy-2 and Zircaloy-4 oxidation in steam were measured over the temperature range 1050-1850 o C. The reaction rates for both alloys were similar, obeyed parabolic kinetics and were not limited by gas phase diffusion. The parabolic rate constants were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam. A discontinuity was found in the temperature dependence of both the reaction rate and the rate of growth of the combined (ZrO 2 + α-Zr) layer. The discontinuity is attributed to a change in the oxide microstructure at the discontinuity temperature, an observation which is consistent with the zirconium-oxygen phase diagram. (author)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  13. Zircaloy oxidation studies

    International Nuclear Information System (INIS)

    Prater, J.T.; Beauchamp, R.H.; Saenz, N.T.

    1985-06-01

    The oxidation kinetics of Zircaloy-4 in steam have been determined at 1300-2400 0 C. Growth of the ZrO 2 and α-Zr layers display parabolic behavior over the entire temperature range studied. A discontinuity in the oxidation kinetics at 1510 0 C causes rates to increase above those previously established by the Baker-Just relationship. This increase coincides with the tetragonal-to-cubic phase transformation in ZrO/sub 2-x/. No discontinuity in the oxide growth rate is observed upon melting of Zr(0). The effects of temperature gradients have been taken into account and corrected values representative of near-isothermal conditions have been computed

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

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  16. Effects of stress on the oxide layer thickness and post-oxidation creep strain of zircaloy-4

    International Nuclear Information System (INIS)

    Lim, Sang Ho; Yoon, Young Ku

    1986-01-01

    Effects of compressive stress generated in the oxide layer and its subsequent relief on oxidation rate and post-oxidation creep characteristics of zircaloy-4 were investigated by oxidation studies in steam with and without applied tensile stress and by creep testing at 700 deg C in high purity argon. The thickness of oxide layer increased with the magnitude of tensile stress applied during oxidation at 650 deg C in steam whereas similar phenomenon was not observed during oxidation at 800 deg C. Zircaloy-4 specimens oxidized at 600 deg C in steam without applied stress exhibited higher creep strain than that shown by unoxidized specimens when creep-tested in argon. Zircaloy-4 specimens oxidized at 600 deg C steam under the applied stress of 8.53MPa and oxidized at 800 deg C under the applied stress of 0 and 8.53MPa exhibited lower strain than that shown by unoxidized specimen. The above experimental results were accounted for on the basis of interactions among applied stress during oxidation, compressive stress generated in the oxide layer and elasticity of zircaloy-4 matrix. (Author)

  17. Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions

    Science.gov (United States)

    Narukawa, Takafumi; Yamaguchi, Akira; Jang, Sunghyon; Amaya, Masaki

    2018-02-01

    For estimating fracture probability of fuel cladding tube under loss-of-coolant accident conditions of light-water-reactors, laboratory-scale integral thermal shock tests were conducted on non-irradiated Zircaloy-4 cladding tube specimens. Then, the obtained binary data with respect to fracture or non-fracture of the cladding tube specimen were analyzed statistically. A method to obtain the fracture probability curve as a function of equivalent cladding reacted (ECR) was proposed using Bayesian inference for generalized linear models: probit, logit, and log-probit models. Then, model selection was performed in terms of physical characteristics and information criteria, a widely applicable information criterion and a widely applicable Bayesian information criterion. As a result, it was clarified that the log-probit model was the best among the three models to estimate the fracture probability in terms of the degree of prediction accuracy for both next data to be obtained and the true model. Using the log-probit model, it was shown that 20% ECR corresponded to a 5% probability level with a 95% confidence of fracture of the cladding tube specimens.

  18. The oxidation kinetics of zircaloy - 4 under isothermal conditions

    International Nuclear Information System (INIS)

    Santos, A.M.M. dos; Cardoso, P.E.

    1982-01-01

    The oxidation kinetics of zircaloy-4 tubes was studied by means of isothermal tests in the temperature interval 500 0 C to 900 0 C. Dry oxygen and water steam, were used as oxidant agents. The results show that the oxidation kinetics law exhibits a behaviour from cubic to parabolic in the range of the time and temperatures of the experiment. Dry oxygen shows a stronger oxidation effect than water steam. A special mechanical test to study the embrittlement effect in the small samples of zircaloy tubes was used. (Author) [pt

  19. Stress corrosion cracking of Zircaloys. Final report

    International Nuclear Information System (INIS)

    Cubicciotti, D.; Jones, R.L.; Syrett, B.C.

    1980-03-01

    The overall aim has been to develop an improved understanding of the stress corrosion cracking (SCC) mechanism considered to be responsible for pellet-cladding interaction (PCI) failures of nuclear fuel rods. The objective of the present phase of the project was to investigate the potential for improving the resistance of Zircaloy to iodine-induced SCC by modifying the manufacturing techniques used in the commercial production of fuel cladding. Several aspects of iodine SCC behavior of potential relevance to cladding performance were experimentally investigated. It was found that the SCC susceptibility of Zircaloy tubing is sensitive to crystallographic texture, surface condition, and residual stress distribution and that current specifications for Zircaloy tubing provide no assurance of an optimum resistance to SCC. Additional evidence was found that iodine-induced cracks initiate at local chemical inhomogeneities in the Zircaloy surface, but laser melting to produce a homogenized surface layer did not improve the SCC resistance. Several results were obtained that should be considered in models of PCI failure. The ratio of axial to hoop stress and the temperature were both shown to affect the SCC resistance whereas the difference in composition between Zircaloy-2 and Zircaloy-4 had no detectable effect. Damage accumulation during iodine SCC was found to be nonlinear: generally, a given life fraction at low stress was more damaging than the same life fraction at higher stress. Studies of the thermochemistry of the zirconium-iodine system (performed under US Department of Energy sponsorship) revealed many errors in the literature and provided important new insights into the mechanism of iodine SCC of Zircaloys

  20. Simulation of Zircaloy cladding deformation under accident conditions derived from analysis of data from Three Mile Island-2

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    A limited series of tests has been carried out based on a published analysis of Three Mile Island data. Zircaloy PWR cladding specimens were pressurised to 6.9 MPa at 500 deg C and heated at 0.2-1.0 deg C/sec in slowly flowing steam until they failed. The temperature at which rupture occurred ranged from 700 to 760 deg C. Three specimens were directly heated, and one was indirectly heated using an internal heater. The lengths of cladding strained greater than 33% ranged from 5.7 to 9.7 times the original diameter

  1. 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

  2. Characterization of electron beam welded Zircaloy-4

    International Nuclear Information System (INIS)

    Anishetty, Sharath; Manna, I.; Majumdar, J. Dutta

    2015-01-01

    Zirconium (Zr) alloys are the backbone materials for thermal reactors because of their low neutron absorption cross section and in addition have suitable properties like high temperature mechanical and corrosion properties. For various structural applications, different Zirconium based alloys are used. Zircaloy-4 (Zr-4) is most commonly used as channel boxes in boiling water reactors (BWRs), intermediate grid applications in pressurized water reactors (PWRs) and in fuel cladding. Zircaloy cladding acts as a barrier between the radioactive fuel and exterior coolants. Therefore, the structural integrity of the cladding tube is extremely important in the safe operation of reactors. Efforts are being made to produce Zircaloy-4 products with better mechanical properties. Different routes of processing are involved like forging, pilgering and extrusion are developed over years in fabricating components to improve in-reactor performance. In this study, microstructure and hardness properties of electron beam welded Zr-4 was evaluated

  3. Cold spray deposition of Ti{sub 2}AlC coatings for improved nuclear fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Maier, Benjamin R. [University of Wisconsin, Madison, WI (United States); Garcia-Diaz, Brenda L. [Savannah River National Laboratory, Aiken, SC (United States); Hauch, Benjamin [University of Wisconsin, Madison, WI (United States); Olson, Luke C.; Sindelar, Robert L. [Savannah River National Laboratory, Aiken, SC (United States); Sridharan, Kumar, E-mail: kumar@engr.wisc.edu [University of Wisconsin, Madison, WI (United States)

    2015-11-15

    Coatings of Ti{sub 2}AlC MAX phase compound have been successfully deposited on Zircaloy-4 (Zry-4) test flats, with the goal of enhancing the accident tolerance of LWR fuel cladding. Low temperature powder spray process, also known as cold spray, has been used to deposit coatings ∼90 μm in thickness using powder particles of <20 μm. X-ray diffraction analysis showed the phase-content of the deposited coatings to be identical to the powders indicating that no phase transformation or oxidation had occurred during the coating deposition process. The coating exhibited a high hardness of about 800 H{sub K} and pin-on-disk wear tests using abrasive ruby ball counter-surface showed the wear resistance of the coating to be significantly superior to the Zry-4 substrate. Scratch tests revealed the coatings to be well-adhered to the Zry-4 substrate. Such mechanical integrity is required for claddings from the standpoint of fretting wear resistance and resisting wear handling and insertion. Air oxidation tests at 700 °C and simulated LOCA tests at 1005 °C in steam environment showed the coatings to be significantly more oxidation resistant compared to Zry-4 suggesting that such coatings can potentially provide accident tolerance to nuclear fuel cladding. - Highlights: • Deposited Ti{sub 2}AlC coatings on Zircaloy-4 substrates with a low pressure powder spray process, also known as cold spray. • Coatings have high hardness and wear resistance for both damage resistance during rod insertion and fretting wear resistance. • The oxidation resistance of Ti{sub 2}AlC coated Zircaloy-4 at 700 °C and 1005 °C was significantly superior to uncoated Zircaloy. • Cold spray of Ti{sub 2}AlC demonstrates considerable promise as a near-term solution for accident tolerant Zr-alloy fuel claddings.

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

    Science.gov (United States)

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

    2014-08-15

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

  5. Autoclave corrosion of zircaloy-4 cladding samples in LiOH solutions

    International Nuclear Information System (INIS)

    Hermann, A.

    2010-03-01

    In reactor operation, pH of the cooling water is adjusted by addition of alkaline hydroxides, and LiOH has been found to be the most suitable one. The addition of LiOH above a certain concentration level (depending on temperature) increases the corrosion rate of zirconium and its alloys. Hydrogen pick-up by the metal is also increased, and this effect is used to produce hydrided specimens for different investigations using the corrosion reaction. At the Paul Scherrer Institute several projects were accomplished to investigate the influence of hydrogen in Zircaloy cladding on its mechanical properties. In order to produce hydrided specimens for comparison and for adjusting new equipment, Zircaloy tubing samples were hydrogen charged by autoclave corrosion in lithiated water. Results of the corrosion experiments are outlined in this publication. Because of the great variety of possible experimental parameters these results are still of interest for the scientific community. Autoclave corrosion was accomplished in 0.2 M or 0.5 M LiOH solution at a constant temperature of 340 o C and a pressure of 160 bar. The corrosion rate increases from 84 mg/(dm 2 d) in 0.2 M LiOH to 153 mg/(dm 2 d) in 0.5 M LiOH. The hydrogen pick-up fraction in 0.5 M LiOH amounts to 80%. In 0.5 M LiOH, Zircaloy tubing samples can be charged with ∼ 500 ppm hydrogen in about 40 hours. In the corrosion experiments described in this report a homogeneous distribution of hydrides should be expected (except very high hydride concentrations) because no temperature gradient exists through the tubing wall. Hydrogen stringers are homogeneously distributed with circumferential orientation (stress-relieved tubing samples). (author)

  6. Zirconium metal-water oxidation kinetics. III. Oxygen diffusion in oxide and alpha Zircaloy phases

    International Nuclear Information System (INIS)

    Pawel, R.E.

    1976-10-01

    The reaction of Zircaloy in steam at elevated temperature involves the growth of discrete layers of oxide and oxygen-rich alpha Zircaloy from the parent beta phase. The multiphase, moving boundary diffusion problem involved is encountered in a number of important reaction schemes in addition to that of Zircaloy-oxygen and can be completely (albeitly ideally) characterized through an appropriate model in terms of oxygen diffusion coefficients and equilibrium concentrations for the various phases. Conversely, kinetic data for phase growth and total oxygen consumption rates can be used to compute diffusion coefficients. Equations are developed that express the oxygen diffusion coefficients in the oxide and alpha phases in terms of the reaction rate constants and equilibrium solubility values. These equations were applied to recent experimental kinetic data on the steam oxidation of Zircaloy-4 to determine the effective oxygen diffusion coefficients in these phases over the temperature range 1000--1500 0 C

  7. Study on the improvement of nuclear fuel cladding reliability

    International Nuclear Information System (INIS)

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

    1987-12-01

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

  8. Mechanisms of damage to the oxide layer of cladding of fuel rods under accident conditions like RI

    International Nuclear Information System (INIS)

    Busser, Vincent

    2009-01-01

    During reactivity initiated accident, the importance of cladding tube oxidation on its thermomechanical behavior has been investigated. After RIA tests in experimental reactors oxide damage including radial cracking and spallation of the outer oxide layer has been evidenced. This work aims at better understanding the key mechanisms controlling these phenomena. Laboratory air-oxidation of Zircaloy-4 cladding tubes has been performed at 470 C. SEM micrographs show that radial cracks are initiated from the outer surface of the oxide layer and propagated radially towards the oxide-metal interface. A model predicting the stress evolution within the oxide and the depth of crack has been developed and validated on literature tests and tests of this study. Ring compression tests were used for the experimental study of the oxide degradation under mechanical loading. Experimental data revealed three mechanisms: densification of the radial crack network, propagation of these radial cracks, branching and spallation of oxide fragments. The influence of the circumferential cracks, periodically distributed in the oxide layer, on the stress distribution in oxide fragments has been analysed using finite element modelling. The determining influence of these cracks on the maximum stress oxide fragments has been demonstrated. (author)

  9. Mechanical analysis of surface-coated zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Lee, You Ho; Lee, Jeong Ik; No, Hee Cheon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2017-08-15

    A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

  10. 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)

  11. Plastic deformation and fracture behavior of zircaloy-2 fuel cladding tubes under biaxial stress

    International Nuclear Information System (INIS)

    Maki, Hideo; Ooyama, Masatosi

    1975-01-01

    Various combinations of biaxial stress were applied on five batches of recrystallized zircaloy-2 fuel cladding tubes with different textures; elongation in both axial and circumferential directions of the specimen was measured continuously up to 5% plastic deformation. The anisotropic theory of plasticity proposed by Hill was applied to the resulting data, and anisotropy constants were obtained through the two media of plastic strain loci and plastic strain ratios. Comparison of the results obtained with the two methods proved that the plastic strain loci provide data that are more effective in predicting quantitatively the plastic deformation behavior of the zircaloy-2 tubes. The anisotropy constants change their value with progress of plastic deformation, and judicious application of the effective stress and effective strain obtained on anisotropic materials will permit the relationship between stress and strain under various biaxialities of stresses to be approximated by the work hardening law. The test specimens used in the plastic deformation experiments were then stressed to fracture under the same combination of biaxial stress as in the proceeding experiments, and the deformation in the fractured part was measured. The result proved that the tilt angle of the c-axis which serves as the index of texture is related to fracture ductility under biaxial stress. Based on this relationship, it was concluded that material with a tilt angle ranging from 10 0 to 15 0 is the most suitable for fuel cladding tubes, from the viewpoint of fracture ductility, at least in the case of unirradiated material. (auth.)

  12. Experimental and calculation results of the integral reflood test QUENCH-14 with M5 (registered) cladding tubes

    International Nuclear Information System (INIS)

    Stuckert, J.; Birchley, J.; Grosse, M.; Jaeckel, B.; Steinbrueck, M.

    2010-01-01

    The QUENCH-14 experiment investigated the effect of M5 (registered) cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institut (Switzerland) using the SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. Follow-on post-test analyses were performed using SCDAP/RELAP5 and MELCOR as part of an ongoing programme of model validation and code assessment. Alternative oxidation correlations were used to examine the possible influence of the M5 (registered) cladding material on hydrogen generation, in comparison with Zircaloy-4. The experiment started with a pre-oxidation phase in steam, lasting ∼3000 s at ∼1500 K peak bundle temperature. After a further temperature increase to maximum bundle temperature of 2073 K the bundle was flooded with 2 g/s/rod water from the bottom. The peak temperature of ∼2300 K was measured on the bundle shroud, shortly after quench initiation. The electrical power was reduced to average value of 2 W/cm during the reflood phase to simulate effective decay heat level. Complete bundle cooling was reached in 300 s after reflood initiation. The development of the oxide layer growth during the test was essentially defined by measurements performed on the three Zircaloy-4 corner rods withdrawn successively from the bundle. The withdrawal of Zircaloy-4 and E110 corner rods after the test allowed a comparison of the different alloys in one test. One heated rod with M5 cladding was withdrawn after the test for a detailed analysis of oxidation degree and measurement of absorbed

  13. 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

  14. Status of Zircaloy deformation and oxidation research at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Chapman, R.H.; Cathcart, J.V.; Hobson, D.O.

    1976-01-01

    The U.S. Nuclear Regulatory Commission sponsors a broad range of research on the response of nuclear fuel assemblies to normal, off-normal, and accident conditions in light-water reactors. The paper reviews the current status of three Zircaloy cladding research programs in progress at the Oak Ridge National Laboratory and presents some preliminary results from each

  15. Study of radiation effects on zircaloy 4 microstructure (Impact on susceptibility to fuel pellet-cladding interaction in PWR)

    International Nuclear Information System (INIS)

    Lefebvre, F.

    1989-01-01

    In PWR the fast neutron flux is an important parameter for fuel can aging by modification of zircaloy-4 microstructure: amorphisation and dissolution of intermetallic precipitates. These phenomena are both analysed and their influence on fuel-cladding interaction is discussed. Irradiations by 1 MeV electrons, Ar ions, Kr ions and fast neutrons are realized for comparison of damages with different defect creation kinetics. Amorphisation is explained as the crystal amorphous state transformation allowing precipitate dissolution by creation of a chemical potential gradient between matrix and amorphous phase. Progressive dissolution of precipitates produced by irradiation decrease the number of potential sites for stress corrosion cracking, improving rupture resistance of the alloy by fuel-cladding interaction [fr

  16. Precipitates in irradiated Zircaloy

    International Nuclear Information System (INIS)

    Chung, H.M.

    1985-10-01

    Precipitates in high-burnup (>20 MWd/kg U) Zircaloy spent-fuel cladding discharged from commercial boiling- and pressurized-water reactors have been characterized by TEM-HVEM. Three classes of primary precipitates were observed in the irradiated Zircaloys: Zr 3 O (2 to 6 nm), cubic-ZrO 2 (greater than or equal to 10 nm), and delta-hydride (35 to 100 nm). The former two precipitations appears to be irradiation induced in nature. Zr(Fe/sub x/Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/Ni/sub 1-x/) intermetallics, which are the primary precipitates in unirradiated Zircaloys, were largely dissolved after the high burnup. It seems, therefore, that the influence of the size and distribution of the intermetallics on the corrosion behavior may be quite different for the irradiated Zircaloys

  17. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    International Nuclear Information System (INIS)

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-01-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  18. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    P. Macheret

    2001-01-01

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

  19. Biaxial creep deformation of Zircaloy-4 PWR fuel cladding in the alpha,(alpha + beta) and beta phase temperature ranges

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Healey, T.; Horwood, R.A.L.

    1985-01-01

    The biaxial creep behaviour of Zircaloy-4 fuel cladding has been determined at temperatures between 973 - 1073 K in the alpha phase range, in the duplex (alpha + beta) region between 1098 - 1223 K and in the beta phase range between 1323 - 1473 K. This paper presents the creep data together with empirical equations which describe the creep deformation response within each phase region. (author)

  20. Interaction between zircaloy tube and inconel spacer grid at high temperature

    International Nuclear Information System (INIS)

    Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi; Furuta, Teruo

    1990-09-01

    In order to investigate the interaction between fuel cladding and spacer grid of the pressurized water reactor during a severe accident, isothermal reaction tests were performed at the temperature range from 1248 to 1673K. A specimen consisted of a short Zircaloy-4 cladding tube and a piece of spacer grid of Inconel-718. In the tests in an argon atmosphere, eutectic reaction between Zircaloy and Inconel was observed at the contact points at 1248K. Rapid reaction was observed at higher test temperatures. For example, in the test at 1373K for 300s, Zircaloy reacted with Inconel over the entire thickness (0.62mm) of the tube in the vicinity of the contact point. In the present tests, Zircaloy which has higher melting point than Inconel was dissolved preferentially due to eutectic formation. In the tests in an oxygen atmosphere, no eutectic reaction was observed at temperatures below 1437K. A trace of interaction was found at the contact point of specimen heated at 1573 and 1623K. However, decrease in Zircaloy thickness was not measured. The possibility of eutectic reaction between Zircaloy cladding and Inconel spacer grid seems to be quite limited when sufficient oxygen is supplied. (author)

  1. Out-of-pile UO2/Zircaloy-4 experiments under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Hofmann, P.

    1983-01-01

    Chemical interactions between UO 2 fuel and Zircaloy-4 cladding up to the melting point of zircaloy (Zry) are described. Out-of-pile UO 2 /zircaloy reaction experiments have been performed to investigate the chemical interaction behavior under possible severe fuel damage conditions (very high temperatures and external overpressure). The tests have been conducted in inert gas (1 to 80 bar) with 10-cm-long zircaloy cladding specimens filled with UO 2 pellets. The annealing temperature varied between 1000 and 1700 deg. C and the annealing period between 1 and 150 min. The extent of the chemical reaction depends decisively on whether or not good contact between UO 2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO 2 to form oxygen-stabilized α-Zr(O) and uranium metal. The uranium reacts with zircaloy to form a (U,Zr) alloy rich in uranium. The (U,Zr) alloy, which is liquid above approx. 1150 deg. C, lies between two α-Zr(O) layers. The UO 2 /zircaloy reaction obeys a parabolic rate law. The degree of chemical interaction is determined by the extent of oxygen diffusion into the cladding, and hence by the time and temperature. The affinity of zirconium for oxygen, which results in an oxygen gradient across the cladding, is the driving force for the reaction. The growth of the reaction layers can be represented in an Arrhenius diagram. The UO 2 /Zry-4 reaction occurs as rapidly as the steam/Zry-4 reaction above about 1100 deg. C. The extent of the interaction is independent of external pressure above about 10 bar at 1400 deg. C and 5 bar at 1700 deg. C. The maximum measured oxygen content of the cladding is approx. 6wt.%. Up to approx. 9 volume % of the UO 2 can be chemically dissolved by the zircaloy. In an actual fuel rod, complete release of the fission products in this region of the fuel must therefore be assumed. (author)

  2. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Post test investigations of bundle test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Buescher, B.

    1986-11-01

    This KfK report describes the post test investigation of bundle experiment ESBU-2a. ESBU-2a was the second of two bundle tests on the temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS-Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (central tungsten heater, UO 2 -ring pellet and zircaloy cladding). The length was 0.4 meter. The bundle was heated to a maximum temperature of 2175 0 C. Molten cladding which dissolved part of the UO 2 pellets and slumped away from the already oxidized cladding formed a lump in the lower part of the bundle. After the test the bundle was embedded in epoxy and sectioned with a diamand saw, in the region of the refrozen melt. The cross sections were investigated by metallographic examination. The refrozen (U,Zr,O) melt consists variously of three phases with increasing oxygen content (metallic α-Zry, metallic (U,Zr) alloy and a (U,Zr)O 2 mixed oxide), two phases (α-Zry, (U,Zr)O 2 mixed oxide), or one phase ((U,Zr)O 2 mixed oxide). The cross sections show the increasing oxidation of the cladding with increasing elevation (temperature). A strong azimuthal dependency of the oxidation is found. In regions where the initial oxidized cladding is contacted by the melt one can recognize the interaction between the metallic melt and ZrO 2 of the cladding. Oxygen is taken away from the ZrO 2 . If the melt is in direct contact with steam a relatively well defined oxide layer is formed. (orig.) [de

  3. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  4. Dynamic strain aging of zircaloy-4 PWR fuel cladding in biaxial stress state

    International Nuclear Information System (INIS)

    Park, Ki Seong; Lee, Byong Whi

    1989-01-01

    The expanding copper mandrel test performed at three strain rates (3.2x10E-5/s,2.0x10E-6/s and 1.2x10E-7/s) over 553-873 K temperature range by varying the heating rates (8-10deg C/s,1-2deg C/s and 0.5deg C/s) in air and in vacuum (5x10E-5 torr). The yield stress peak, the strain rate sensitivity minimum and the activation volume peaks could be explained in terms of the dynamic strain aging. The activation energy for dynamic strain aging obtained from the yield stress peak temperature and strain rate was 196 KJ/mol and this value was in good agreement with the activation energy for oxygen diffusion in α-zirconium and Zircaloy-2 (207-220KJ/mol). Therefore, oxygen atoms are responsible for the dynamic strain aging which appeared between 573K and 673K. The yield stress increase due to the oxidation was obtained by comparing the yield stress in air with that in vacuum and represented by the percentage increase of yield stress (σ y a -σ y v /σ y v ). The slower the strain rate, the greater the percentage increase occurs. In order to estimate the yield stress of PWR fuel cladding material under the service environment, the yield stress in water was obtained by comparing the oxidation rate in air that in water assuming the relationship between the oxygen pick-up amount and the yield stress increase. (Author)

  5. Reaction behavior between B{sub 4}C, 304 grade of stainless steel and Zircaloy at 1473 K

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, Ryosuke [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan); Ueda, Shigeru, E-mail: tie@tagen.tohokku.ac.jp [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan); Kim, Sun-Joong [Dept. of Materials Science and Engineering, Chosun University, 309, Pilmun-daero, Dong-gu, Gwangju (Korea, Republic of); Gao, Xu; Kitamura, Shin-ya [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan)

    2016-08-15

    For a better understanding of the decommissioning of the Fukushima-daiichi nuclear power plant, the melting behavior of the control blade and the channel box should be clarified. In Fukushima nuclear reactor, the channel box was made of Zircaloy-4, and the control rode is made of B{sub 4}C together with stainless steel cladding and sheath. In the study, the interaction among B{sub 4}C, stainless steel (SUS), and Zircaloy-4 was investigated at 1473 K in either argon or air atmosphere. In argon, Zircaloy is melted by the diffusion of elements from SUS, and SUS was melted at 1473 K by the diffusion of C and B. In air, SUS reacted with B{sub 2}O{sub 3} and formed an oxides melt firstly. Then, the oxidized Zircaloy contacted with this melt and fused. Moreover, the progress of core melting process during severe accident under different atmospheres was firstly discussed. - Highlights: • The interaction among the system of B{sub 4}C, grade 304 stainless steel and Zircaloy-4 were studied at 1473 K in Ar and air. • In argon, Zircaloy is melted by the diffusion of elements from SUS, and SUS was melted by the diffusion of C and B. • In air, SUS reacted with B{sub 2}O{sub 3} and formed an oxides melt. Then, the oxidized Zircaloy contacted with this melt and fused.

  6. Zircaloy-4 corrosion in PWR's

    International Nuclear Information System (INIS)

    Fyfitch, S.; Smalley, W.R.; Roberts, E.

    1985-01-01

    Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

  7. Stress corrosion crack growth in unirradiated zircaloy

    International Nuclear Information System (INIS)

    Pettersson, K.

    1978-10-01

    Experimental techniques suitable for the determination of stress corrosion crack growth rates in irradiated Zircaloy tube have been developed. The techniques have been tested on unirradiated. Zircaloy and it was found that the results were in good agreement with the results of other investigations. Some of the results were obtained at very low stress intensities and the crack growth rates observed, gave no indication of the existance of a K sub(ISCC) for iodine induced stress corrosion cracking in Zircaloy. This is of importance both for fuel rod behavior after a power ramp and for long term storage of spent Zircaloy-clad fuel. (author)

  8. Iodine-induced stress corrosion cracking of fixed deflection stressed slotted rings of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Sejnoha, R.; Wood, J.C.

    1978-01-01

    Stress corrosion cracking of Zircaloy fuel cladding by fission products is thought to be an important mechanism influencing power ramping defects of water-reactor fuels. We have used the fixed-deflection stressed slotted-ring technique to demonstrate cracking. The results show both the sensitivity and limitations of the stressed slotted-ring method in determining the responses of tubing to stress corrosion cracking. They are interpreted in terms of stress relaxation behavior, both on a microscopic scale for hydrogen-induced stress-relief and on a macroscopic scale for stress-time characteristics. Analysis also takes account of nonuniform plastic deformation during loading and residual stress buildup on unloading. 27 refs

  9. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    International Nuclear Information System (INIS)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F.

    2000-01-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  10. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F

    2000-07-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  11. Gallium-cladding compatibility testing plan. Phases 1 and 2: Test plan for gallium corrosion tests. Revision 2

    International Nuclear Information System (INIS)

    Wilson, D.F.; Morris, R.N.

    1998-05-01

    This test plan is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water-Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. The plan summarizes and updates the projected Phases 1 and 2 Gallium-Cladding compatibility corrosion testing and the following post-test examination. This work will characterize the reactions and changes, if any, in mechanical properties that occur between Zircaloy clad and gallium or gallium oxide in the temperature range 30--700 C

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

    Science.gov (United States)

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

    2018-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-25

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

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

    International Nuclear Information System (INIS)

    Heuser, Brent; Stubbins, James; Kozlowski, Tomasz; Uddin, Rizwan; Trinkle, Dallas; Downar, Thoms; Was, Gary; Ang, Yong; Phillpot, Simon; Sabharwall, Piyush

    2017-01-01

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

  15. Modelling of pellet-clad interaction during power ramps

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  17. Conversion of zircaloy to a massive chemically inert form

    International Nuclear Information System (INIS)

    Atkinson, A.; Kearsey, H.A.; Knibbs, R.H.; Mercer, A.C.; Nickerson, A.K.; Pearson, D.; Sambell, R.A.J.; Taylor, R.I.

    1985-01-01

    The report covers work carried out in the period July 1980 - December 1982 on the development and assessment of an aqueous route for the conversion of Zircaloy fuel element cladding to a stable oxide form and on alternative methods for incorporating the oxide into monolithic waste forms suitable for long-term storage and disposal. The work included two aspects, preliminary process development studies aimed at demonstrating the key steps in the process, and studies on the alternative immobilization techniques and the properties of the resulting waste forms. Experimental studies have shown that the ''hydrous zirconium oxide'' (with a residual fluoride content), following calcination at about 500 0 C, can be hot-pressed at 800-1000 0 C and 22.5 MPa to a high density ceramic waste form with good capacity for the incorporation of active species, such as U 4+ and Sr 2+ , and high leach resistance. Parallel studies have been carried out on the incorporation of the washed ''hydrous zirconium oxide'' in a range of cement matrices. A preliminary chemical engineering assessment of the overall process has been made and flowsheets for a plant to convert 250 kg Zircaloy/day have been prepared

  18. The effect of stimulated fission products on the structure and the mechanical properties of zircaloy

    International Nuclear Information System (INIS)

    Holub, F.

    1982-01-01

    The objective of investigation was to study the long-term effects of individual simulated fission products on the mechanical properties and the structure of Zircaloy. Tensile Test specimens of Zircaloy were annealed with important simulated fission products at 350 0 C up to 10,000 hours and at higher temperatures (500, 700 0 C) up to 2,000 hours. The principal methods of investigation on annealed Zircaloy specimens were tension tests at room temperature and at 400 0 C, scanning electron microscopy and microprobe technique, X-ray diffraction, X-ray fluorescence, optical metallography. The action of fission products at normal temperatures of reactor operation will give rise to a small enhancement of strength and a small drop of ductility of the fuel cladding material only. At high fuel pin temperatures which may be realized under abnormal operation conditions, some of the fission products potentially will produce detrimental consequences on the integrity of fuel pins. The most effective fission products will be: lanthanum oxide, followed by the earth alkaline oxides and the other rare earth oxides, molybdenum, iodine and cadmium

  19. Chemical and microstructural characterization of recycled zircaloy

    International Nuclear Information System (INIS)

    Martinez, Luis G.; Pereira, Luiz A.T.; Rossi, Jesualdo L.; Takiishi, Hidetoshi; Sato, Ivone M.; Scapin, Marcos A.; Orlando, Marcos T.D.

    2011-01-01

    PWR reactors employ as nuclear fuel UO 2 pellets with Zircaloy clad. Brazil is autonomous in the nuclear fuel cycle, from uranium mining to enrichment and nuclear fuel manufacture. However, the industrial production of nuclear zirconium alloys does not meet the demand, leading to importation of Zircaloy for fuel manufacturing. In the fabrication of fuel elements parts, machining chips of alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is strategic in economical and environmental aspects. In this work are described two methods that are being developed to recycle Zircaloy chips. The first method the Zircaloy machining chips are melted using an electric arc furnace to obtain small laboratory ingots. The second method uses powder metallurgy technique. By this later method, the Zircaloy chips are submitted to a hydriding process and the resulting material is milled in a high-energy ball mill. The powder is cold isostatically pressed and vacuum sintered. The elemental composition of the materials obtained using both methods is being determined using X-ray fluorescence techniques and compared to the specifications of nuclear grade Zircaloy and to the composition of the starting chips. The phase composition of the laboratory ingots was determined using X-ray diffraction. The ingots were vacuum annealed and the microstructures resulting from both processing methods before and after heat treatments were characterized using optical and scanning electron microscopy. The hardness of the materials was evaluated. A methodology of chemical analysis using X-ray fluorescence spectrometry, for composition certification, was established and tested. The results showed that recycled Zircaloy presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding cap-ends, using near net shape sintering. (author)

  20. Modeling of the cold work stress relieved Zircaloy-4 cladding tubes mechanical behavior under PWR operating conditions

    International Nuclear Information System (INIS)

    Richard, F.; Delobelle, P.; Leclercq, S.; Bouffioux, P.; Rousselier, G.

    2003-01-01

    This paper proposes a damaged viscoplastic model to simulate, for different isotherms (320, 350, 380, 400 and 420 degC), the out-of-flux anisotropic mechanical behavior of cold work stress relieved Zircaloy-4 cladding tubes over the fluence range 0-85.1024 nm -2 (E > 1 MeV). The model, identified from uni and biaxial tests conducted at 350 and 400 degC, is validated from tests performed at 320, 380 and 420 degC. This model is able to simulate strain hardening under internal pressure followed by a stress relaxation period (thermal creep), which is representative of a pellet cladding mechanical interaction occurring during a power transient (class 2 incidental condition). Both the integration of a scalar state variable, characterizing the damage caused by a bombardment with neutrons, and the modification of the static recovery law allowed us to simulate the fast neutron flux effect (irradiation creep). (author)

  1. Release of indigenous gases from LWR fuel and the reaction kinetics with Zircaloy cladding

    International Nuclear Information System (INIS)

    Beyer, C.E.; Hann, C.R.

    1977-04-01

    The objective of this study was to evaluate the open literature data to estimate: the rate of gaseous impurity release from oxide fuel, the amount and composition of the gaseous impurities, and their subsequent rate of reaction with the fuel or Zircaloy

  2. Chemical interactions between as-received and pre-oxidized Zircaloy-4 and stainless steel at high temperatures

    International Nuclear Information System (INIS)

    Hofmann, P.

    1994-05-01

    The chemical reaction behavior between Zircaloy-4 and 1.4919 (AISI 316) stainless steel, which are used in absorber assemblies of Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR), has been studied in the temperature range 1000 - 1400 C. Zircaloy was used in the as-received, pre-oxidized and oxygen-containing condition. The maximum temperature was limited by the fast and complete liquefaction of the reaction couple as a result of eutectic chemical interactions. Liquefaction of the components occurs below their melting point. The effect of oxygen dissolved in Zircaloy plays an important role in the interaction; oxide layers on the Zircaloy surface delay the chemical interactions with stainless steel but cannot prevent them. Oxygen dissolved in Zircaloy reduces the reaction rates and shift the liquefaction temperature to slightly higher levels. The interaction experiments at the examined temperatures with or without pre-oxidized Zircaloy can be described by parabolic rate laws. The Arrhenius equations for the various conditions of interactions are given. (orig.) [de

  3. Deformation behavior of Zircaloy-4 cladding tubes under inert gas conditions in the temperature range from 600 to 12000C

    International Nuclear Information System (INIS)

    Hofmann, P.; Raff, S.; Gausmann, G.

    1981-07-01

    Within the temperature range from 600 0 to 1200 0 isothermal, isobaric creep rupture experiments were performed under inert gas with short Zircaloy-4 tube specimens in order to obtain experimental data supporting the development of the NORA cladding tube deformation model. The values of the tube inner pressure were so selected that the time-to-failure values varied between 2 and 2000 s. The corresponding creep rupture curves are indicated. Besides the temperature and the burst pressure the development of deformation over time of the tube specimens was measured. This allowed to draw diagrams of stress, strain rate and strain. On account of the type of specimen heating applied (radiation heating) the temperature difference at the cladding tube circumference is very small ( [de

  4. 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

  5. Effect of thermal transients on the hardness of Zircaloy fuel cladding

    International Nuclear Information System (INIS)

    Hobson, D.O.

    1976-06-01

    This study is directed toward the determination of the effects of annealing cycles with rapid heating rates, short hold times at specific temperatures, and rapid cool-down rates on the hardness of Zircaloy fuel cladding. These rapid annealing cycles are designed to provide preliminary annealing behavior data on Loss-of-Fluid-Test Reactor cladding samples. Information has been obtained on (1) the time dependence of the hardness as a function of annealing temperature, and (2) a correlation of single- and multitransient annealing relationships. Both single- and triple-cycle transients were used; four hold times at each of five maximum temperatures comprised the data set (each portion of the triple-cycle experiments had isothermal hold times equal to one-third of their analogous single-cycle times). It was found that there was little difference in the hardness response between single- and triple-cycle transients for a given total hold time at a particular temperature. Test temperatures range from 1000 to 1400 0 F (538 to 760 0 C) and hold times from 5 to 135 sec. The 1100 0 F (593 0 C) level was found to be the transition level for hardness changes, with shorter times (5 and 15 sec) effecting little or no hardness decrease and the longer times (45 and 135 sec) producing partially and fully annealed material, respectively. Temperatures equal to or greater than 1300 0 F (704 0 C) resulted in fully annealed material for all hold times. The 1000 0 F (538 0 C) tests produced no measurable softening

  6. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    D. Stahl

    2004-01-01

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

  7. Laboratory simulation of rod-to-rod mechanical interactions during postulated loss-of-coolant accidents in a PWR involving cladding oxidation

    International Nuclear Information System (INIS)

    Hindle, E.D.; Haste, T.J.; Harrison, W.R.

    1987-01-01

    Creep deformation of Zircaloy cladding in postulated PWR loss-of-coolant accidents may lead to rod-to-rod mechanical interactions. Tests have been performed in the electrically heated FOURSQUARE rig at 750 0 C and 850 0 C in steam to investigate this effect. Conservatisms inherent in a simple 'square with rounded corners' coolant channel blockage model have been quantified; about 5-10% flow area may remain even at strains which in ideal circumstances would give total blockage. Reduction of average burst strains produced by an oxide layer (up to 13 μm) has been demonstrated, resulting from strain concentration at oxide cracks. (author)

  8. In-situ neutron diffraction study of Zircaloy 4 subjected to biaxial tension

    Energy Technology Data Exchange (ETDEWEB)

    Gharghouri, M.A. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, ON (Canada); McDonald, D.; Xiao, L. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Zircaloy-4 is widely used as fuel element cladding in nuclear reactors. Pellet-clad interaction (PCI) failure is a concern for many water reactor fuel designs. Extensive work on the mechanism of PCI failure has led to the conclusion that stress corrosion cracking (SCC) induced by iodine vapour in the temperature range relevant to fuel operation is the most probable cause of PCI failure in zirconium alloy fuel element cladding. In-situ neutron diffraction measurements performed on tubular Zircaloy-4 specimens simultaneously pulled in tension and pressurized internally will provide information on the effects of stress biaxiality on the distribution of stresses at the crystal level during loading. (author)

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

    International Nuclear Information System (INIS)

    El Genk, M.S.

    1979-01-01

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

  10. 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

  11. Effect of reactor chemistry and operating variables on fuel cladding corrosion in PWRs

    International Nuclear Information System (INIS)

    Park, Moon Ghu; Lee, Sang Hee

    1997-01-01

    As the nuclear industry extends the fuel cycle length, waterside corrosion of zircaloy cladding has become a limiting factor in PWR fuel design. Many plant chemistry factors such as, higher lithium/boron concentration in the primary coolant can influence the corrosion behavior of zircaloy cladding. The chemistry effect can be amplified in higher duty fuel, particularlywhen surface boiling occurs. Local boiling can result in increased crud deposition on fuel cladding which may induce axial power offset anomalies (AOA), recently reported in several PWR units. In this study, the effect of reactor chemistry and operating variables on Zircaloy cladding corrosion is investigated and simulation studies are performed to evaluate the optimal primary chemistry condition for extended cycle operation. (author). 8 refs., 3 tabs., 16 figs

  12. 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)

  13. Effect of the aluminum flow pattern on the bonding of aluminum to oxidized Zircaloy-2

    International Nuclear Information System (INIS)

    Watson, R.D.; Lambert, J.P.

    1965-04-01

    The bonds produced when hot aluminum is allowed to flow smoothly from an extrusion die to the oxidized surface of a heated tube of Zircaloy-2 are consistently inferior to those produced with back-extruded flow. The difference is believed to be due to the reduction in, or elimination of, the oxide layer on the aluminum that comes in contact with the surface of the Zircaloy-2. This method of bonding aluminum to Zircaloy-2 is covered by Canadian patent 702,438 January 1965. (author)

  14. Apparatus for study of transient oxidation of Zircaloy-4 tubing

    International Nuclear Information System (INIS)

    Sagat, S.; Iglesias, F.C.; Newell, G.W.

    1985-11-01

    Complex transient oxidation tests on Zircaloy-4 tubing were performed to provide data for validation of the computer code FROM2. This code was developed to calculate oxygen distribution through oxidized Zircaloy tubing. The test temperature histories consisted of ramp, hold and cool cycles. The heating and cooling rates were in the range of 1 to 100 K/s and the maximum temperature was 1875 K. The apparatus developed to perform these experiments is described. In principle, Joule heating is used to heat the specimen and the temperature is controlled by a computer in conjunction with temperature and SCR power controllers. Using this combination, fast heating and cooling rates were achieved without sacrificing the accuracy of temperature control

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  16. Experiments on ballooning in pressurized and transiently heated Zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Markiewicz, M.E.; Erbacher, F.J.

    1988-02-01

    Single-rod burst tests were performed with Atucha I Zircaloy-4 cladding tubes in the REBEKA burst equipment of KfK. The objective was to investigate the ballooning and burst behavior of argentine cladding tubes obtained from NRG, Germany and CONVAR, Argentina. The burst data were compared with those of cladding tubes used in german PWR's. It was found that the burst data e.g. burst temperature, circumferential burst strain and its response to azimuthal temperature differences are identical for the Argentine and German tubing quality. The burst data are in good agreement with those of German PWR-Zircaloy tubes. Thus, the fuel rod behavior codes developed for German PWR's can also be used for the Argentine reactor Atucha I. (orig.) [de

  17. High temperature steam oxidation of Al3Ti-based alloys for the oxidation-resistant surface layer on Zr fuel claddings

    International Nuclear Information System (INIS)

    Park, Jeong-Yong; Kim, Il-Hyun; Jung, Yang-Il; Kim, Hyun-Gil; Park, Dong-Jun; Choi, Byung-Kwon

    2013-01-01

    We investigated the feasibility to apply Al 3 Ti-based alloys as the surface layer for improving the oxidation resistance of Zr fuel claddings under accident conditions. Two types of Al 3 Ti-based alloys with the compositions of Al–25Ti–10Cr and Al–21Ti–23Cr in atomic percent were prepared by arc-melting followed by homogenization annealing at 1423 K for 48 h. Al–25Ti–10Cr alloy showed an L1 2 quasi-single phase microstructure with a lot of needle-shaped minor phase and pores. Al–21Ti–23Cr alloy consisted of an L1 2 matrix and Cr 2 Al as the second phase. Al 3 Ti-based alloys showed an extremely low oxidation rate in a 1473 K steam for up to 7200 s when compared to Zircaloy-4. Both alloys exhibited almost the same oxidation rate in the early stage of oxidation, but Al–25Ti–10Cr showed a little lower oxidation rate after 4000 s than Al–21Ti–23Cr. The difference in the oxidation rate between two types of Al 3 Ti-based alloys was too marginal to distinguish the oxidation behavior of each alloy. The resultant oxide exhibited almost the same characteristics in both alloys even though the microstructure was explicitly distinguished from each other. The crystal structure of the oxide formed up to 2000 s was identified as Al 2 O 3 in both alloys. The oxide morphology consisted of columnar grains whose length was almost identical to the average oxide thickness. On the basis of the results obtained, it is considered that Al 3 Ti-based alloy is one of the promising candidates for the oxidation-resistant surface layer on Zr fuel claddings

  18. Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Suman, Siddharth [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Khan, Mohd Kaleem, E-mail: mkkhan@iitp.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Pathak, Manabendra [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Singh, R.N.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2016-10-15

    Highlights: • Modelling of nuclear fuel cladding during loss-of-coolant accident transient. • Phase transformation, corrosion, and creep combined to evaluate burst criterion. • Effect of oxygen concentration on burst stress and burst strain. • Effect of heating rate, internal pressure fluctuation, shear modulus incorporated. - Abstract: A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  20. Coating of Zircaloy sheaths with silica glass using the Sol-Gel technique for protection against oxidation

    International Nuclear Information System (INIS)

    De Sanctis, O.; Pellegri, N.; Gomez, L.

    1990-01-01

    With the aim of improving corrosion resistance of Zircaloy, a few Zircaloy sheaths were covered with vitreous silica. Deposition was made by dip coating in tetraetilortosilicate (TEOS) solutions and later densification treatment at 500 degrees C. Oxidation tests were performed and compared with sheaths not covered with silica. As a result, an effective increase in the resistance to dry oxidation was found in sheaths which had been protected. The coating-Zircaloy interface was studied using XPS (scanner). (Author). 6 refs., 3 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-01

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

  2. Oxidation properties of laser clad Nb-Al alloys

    International Nuclear Information System (INIS)

    Tewari, S.K.; Mazumder, J.

    1992-01-01

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

  3. The development of a burst criterion for Zircaloy fuel cladding under LOCA conditions

    International Nuclear Information System (INIS)

    Neitzel, H.J.; Rosinger, H.E.

    1980-10-01

    A burst criterion model, which assumes that deformation is controlled by steady-state creep, has been developed for a thin-walled cladding, in this case Zircaloy-4, subjected to a differential pressure and high temperature. The creep equation is integrated to obtain a burst time at the singularity of the strain. Once that urst time is known, the burst temperature and burst pressure can be calculated from the known temperature and pressure histories. A further relationship between burst stress and burst temperature is used to calculate the burst strain. Comparison with measured burst data shows good agreement between theory and experiment. It was found that, if the heating rate is constant, the burst temperature increases with decreasing stress, and that, if the stress level is constant, the burst temperature increases with increasing heating rate. It was also found that anisotropy alters the burst temperature and burst strain, and that thest conditions in the α-Zr temperature range have no influence on the burst data. (orig.) [de

  4. The development of a burst criterion for zircaloy fuel cladding under LOCA conditions

    International Nuclear Information System (INIS)

    Neitzel, H.J.; Rossinger, H.E.

    1980-02-01

    A burst criterion model, which assumes that deformation is controlled by steady-state creep, has been developed for a thin-walled cladding, in this case Zircaloy-4, subjected to a differential pressure and high temperature. The creep equation is integrated to obtain a burst time at the singularity of the strain. Once the burst time is known, the burst temperature and burst pressure can be calculated from the known temperature and pressure histories. A further relationship between burst stress and burst temperature is used to calculate the burst strain. Comparison with measured burst data shows good agreement between theory and experiment was found that, if the heating rate is constant, the burst temperature increases with decreasing stress, and that, if the stress level is constant, the burst temperature increases with increasing heating rate. It was also found that anisotropy alters the burst temperature and burst strain, and that test conditions in the α-Zr temperature range have no influence on the burst data. (auth)

  5. Oxidation during reflood of reactor core with melting cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  6. 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

  7. 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

  8. Study of the mechanisms controlling the oxide growth under irradiation: characterization of irradiated zircaloy-4 and Zr-1 Nb-O oxide scales

    International Nuclear Information System (INIS)

    Bossis, Ph.; Thomazet, J.; Lefebvre, F.

    2002-01-01

    In PWRs, the Zr-1Nb-O alloy shows a marked enhancement in corrosion resistance in comparison with Zircaloy-4. The aim of this work is to analyze the reasons for these different behaviors and to determine the respective nature of the oxide growth controlling mechanisms under irradiation. Samples taken from Zircaloy-4 irradiated 1, 2, and 4 cycles and Zr-1Nb-O irradiated 1 and 3 cycles have been systematically characterized by optical microscopy, SEM coupled with image analysis, hydride distribution, and XRD. Specific TEM characterizations have been performed on the Zr-1Nb-O samples. A XPS analysis of a nonirradiated sample is also reported. It has been shown that under irradiation the slow oxidation kinetics of the Zr-1Nb-O alloy is associated with very regular metal-oxide interface and oxide layer. On the contrary, the accelerated oxidation kinetics of Zircaloy-4 is associated with highly perturbed metal-oxide interface and oxide layer. On both irradiated alloys, cracks are observed to initiate preferentially above the delayed parts of the oxidation front. Hydrogen intake during water oxidation in PWR environment is found to be much lower on the Zr-1Nb-O alloy than on Zircaloy-4. More β-ZrO 2 is found on the oxide layer formed on Zircaloy-4 than on Zr-1NbO after oxidation in PWR. Classical irradiation-induced microstructural evolution is observed in the Zr-1Nb-O metallic alloy after 3 cycles, i.e., a fine β-Nb precipitation. β-Nb precipitates are observed to undergo a delayed oxidation associated with a crystalline to amorphous transformation. After water oxidation in autoclave, a pronounced Nb segregation is detected on the oxide surface of a Zr-1Nb-O sample. These results suggest that the oxidation kinetics of Zircaloy-4 is controlled essentially by oxygen diffusion through the inner barrier layer, which is significantly accelerated under irradiation. The oxidation kinetics of Zr-1Nb-O is controlled by both oxygen diffusion through the inner barrier and by

  9. Transmission electron microscopy characterization of Zircaloy-4 and ZIRLO™ oxide layers

    International Nuclear Information System (INIS)

    Gabory, Benoit de; Motta, Arthur T.; Wang, Ke

    2015-01-01

    Waterside corrosion of zirconium alloy nuclear fuel cladding varies markedly from one alloy to another. In addition, for a given alloy, the corrosion rate evolves during the corrosion process, most notably when the oxide loses its stability at the oxide transition. In an effort to understand the mechanism resulting in the variations of corrosion rate observed at the oxide transition, oxide layers formed on Zircaloy-4 and ZIRLO™ in high temperature water autoclave environments, and archived before and after the transition, are characterized using transmission electron microscopy. The study characterizes and compares the oxide morphology in both alloys at different times during the corrosion process, in an effort to understand the oxide growth mechanism for these alloys. Results show that the oxide is mainly composed of monoclinic ZrO 2 , with a preponderance of columnar oxide grains which extend to the oxide/metal interface. The oxide formed right after the transition has occurred, exhibits a 150 nm-wide layer of small equiaxed grains with high tetragonal oxide fraction. This layer has a similar morphology and structure as the first oxide layer formed (observed near the oxide/water interface). A study of the oxygen-rich region near the oxide/metal interface reveals a complex structure of different phases at different stages of corrosion. The interface exhibits an intermediate layer, identified as ZrO, a discontinuous layer of “blocky” Zr 3 O grains embedded in the ZrO layer, and a suboxide layer corresponding to an oxygen saturated solid solution in the metal matrix side. The thickness of this interfacial layer decreased markedly at the transition. Hydrides are also observed in that region, with a definite orientation relationship with the matrix. The observations of the oxide/metal interface are qualitatively similar for the two alloys but quantitatively different. The incorporation of intermetallic precipitates into the oxide layer is also studied, and

  10. Oxiding and hydriding properties of Zr-1Nb cladding material in comparison with zircaloys

    Energy Technology Data Exchange (ETDEWEB)

    Vrtilkova, V; Molin, L [Nuclear Fuel Inst., Zbraslav (Czech Republic); Valach, M [Nuclear Research Inst., Rez plc (Czech Republic)

    1997-02-01

    This paper presents an overview of experimental research related to the Zr-1Nb corrosion behaviour in water and steam environment performed in the Czech Republic. Presented work is focused on the differences between Zr1Nb and Zircaloy corrosion performance. The effects of steam pressure, temperature transients and preoxidation are discussed. (author). 14 refs, 15 figs.

  11. Influence de l'orientation des hydrures sur les modes de déformation, d'endommagement et de rupture du Zircaloy-4 hydruré.

    OpenAIRE

    Racine , Aude

    2005-01-01

    In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300°C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embitterment depends on many parameters, among which hydrogen content and orientation of hydrid...

  12. The effect of oxide microstructure on kinetic transition in out-of-pile steam corrosion test for Zircaloy-2 and Nb-added Zircaloy-2

    Energy Technology Data Exchange (ETDEWEB)

    Nanikawa, Shuichi [Japan Nuclear Fuel Co. Ltd., Yokosuka, Kanagawa (Japan); Etoh, Yoshinori [Japan Nuclear Fuel Co. Ltd., Yokohama, Kanagawa (Japan)

    2001-06-01

    In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS. Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly. The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of {alpha}-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition. The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain. (author)

  13. Treatment of zircaloy cladding hulls by isostatic pressing

    International Nuclear Information System (INIS)

    Tegman, R.; Burstroem, M.

    1984-12-01

    A method for the treatment of Zircaloy fuel hulls is proposed. It involves hot isostatic pressing (HIP) for making large, completely densified metallic bodies of the waste. The hulls are packed into a bellows-shaped container of steel. On packing the fuel hulls give a filling factor of only 14%, which is too low for non-deformable compaction in a normal container, but by using a belloped container, a non-deformable compaction can be obtained without any pretreatment of the hulls. Fully dense and mechanically strong blocks of Zircaloy can be fabricated by holding them at temperatures of around 1000 degrees C for three hours. It is also feasible to incorporate the other metallic parts of the fuel bundle, such as top and bottom tie plates and spacers, in the pressing. The HIP-densified hulls provide an effective means of self-containment of radioactive waste due to the excellent corrosion resistance of Zircaloy. A waste loading factor of close to 100% can be realized. Futher, a volume reduction factor of 7 and a surface reduction factor of aout 250 for a 1-ton canister can be achieved. Equilibrium calculations have shown that tritium present in the hulls can quantitatively be contained in the HIPed block. A study has been made of a possible process for industrilscale use. (Author)

  14. Stress corrosion crack initiation of Zircaloy-4 cladding tubes in an iodine vapor environment during creep, relaxation, and constant strain rate tests

    Science.gov (United States)

    Jezequel, T.; Auzoux, Q.; Le Boulch, D.; Bono, M.; Andrieu, E.; Blanc, C.; Chabretou, V.; Mozzani, N.; Rautenberg, M.

    2018-02-01

    During accidental power transient conditions with Pellet Cladding Interaction (PCI), the synergistic effect of the stress and strain imposed on the cladding by thermal expansion of the fuel, and corrosion by iodine released as a fission product, may lead to cladding failure by Stress Corrosion Cracking (SCC). In this study, internal pressure tests were conducted on unirradiated cold-worked stress-relieved Zircaloy-4 cladding tubes in an iodine vapor environment. The goal was to investigate the influence of loading type (constant pressure tests, constant circumferential strain rate tests, or constant circumferential strain tests) and test temperature (320, 350, or 380 °C) on iodine-induced stress corrosion cracking (I-SCC). The experimental results obtained with different loading types were consistent with each other. The apparent threshold hoop stress for I-SCC was found to be independent of the test temperature. SEM micrographs of the tested samples showed many pits distributed over the inner surface, which tended to coalesce into large pits in which a microcrack could initiate. A model for the time-to-failure of a cladding tube was developed using finite element simulations of the viscoplastic mechanical behavior of the material and a modified Kachanov's damage growth model. The times-to-failure predicted by this model are consistent with the experimental data.

  15. Comparison of the air oxidation behaviors of Zircaloy-4 implanted with yttrium and cerium ions at 500 deg. C

    International Nuclear Information System (INIS)

    Chen, X.W.; Bai, X.D.; Xu, J.; Zhou, Q.G.; Chen, B.S.

    2002-01-01

    As a valuable process for surface modification of materials, ion implantation is eminent to improve mechanical properties, electrochemical corrosion resistance and oxidation behaviors of varieties of materials. To investigate and compare the oxidation behaviors of Zircaloy-4, implantation of yttrium ion and cerium ion were respectively employed by using an MEVVA source at the energy of 40 keV with a dose ranging from 1x10 16 to 1x10 17 ions/cm 2 . Subsequently, weight gain curves of the different specimens including as-received Zircaloy-4 and Zircaloy-4 specimens implanted with the different ions were measured after oxidation in air at 500 deg. C for 100 min. It was obviously found that a significant improvement was achieved in the oxidation behaviors of implanted Zircaloy-4 compared with that of the as-received Zircaloy-4, and the oxidation behavior of cerium-implanted Zircaloy-4 was somewhat better than that of yttrium-implanted specimen. To obtain the valence and the composition of the oxides in the scale, X-ray photoemission spectroscopy was used in the present study. Glancing angle X-ray diffraction, employed to analyze the phase transformation in the oxide films, showed that the addition of yttrium transformed the phase from monoclinic zirconia to tetragonal zirconia, yet the addition of cerium transformed the phase from monoclinic zirconia to hexagonal zirconia. In the end, the mechanism of the improvement of the oxidation behavior was discussed

  16. Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction: Tests ESSI-1,2,3

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

    1983-08-01

    This report discusses the test conduct, results, and posttest appearance of three scoping tests (ESSI-1,2,3) investigating temperature escalation in zircaloy clad fuel rods. The experiments are part of an out-of-pile program using electrically heated fuel rod simulators to investigate PWR fuel element behavior up to temperatures of 2000 0 C. These experiments are part of the PNS Severe Fuel Damage Program. The temperature escalation is caused by the exothermal zircaloy/steam reaction, whose reaction rate increases exponentially with the temperature. The tests were performed using different initial oxide layers as a major parameter, obtained by varying the heatup rates and steam exposure times. (orig./RW) [de

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. Eddy-current testing and analysis of a sample of Zircaloy fuel cladding for the OECD Halden 'Round-Robin' exercise (Phase II)

    International Nuclear Information System (INIS)

    Watson, P.C.; Cross, M.T.

    1987-02-01

    Two samples of Zircaloy fuel cladding were supplied, one containing pre-measured defects of known type and size, and the other containing unknown defects. Eddy-current testing techniques were used to ascertain the nature of the unknown defects. By using a high resolution encircling coil and a probe coil and then processing digitally the data with specially prepared software, nine internal defects, of volume 0.18 to 0.86 mm 3 were located positively and identified, despite interference from heavily fluctuating background signals. (author)

  19. Phase transformations in neutron-irradiated Zircaloys

    International Nuclear Information System (INIS)

    Chung, H.M.

    1986-04-01

    Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after ∼3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr 3 O and cubic-ZrO 2 particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/,Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/,Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of ∼4 x 10 21 ncm -2 in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs

  20. Development of zircaloy deformation model to describe the zircaloy-4 cladding tube during accidents

    International Nuclear Information System (INIS)

    Raff, S.

    1978-01-01

    The development of a high-temperature deformation model for Zircaloy-4 cans is primarily based on numerous well-parametrized tensile tests to get the material behaviour including statistical variance. It is shown that plastic deformation may be described by a power creep law, the coefficients of which show strong dependence on temperature in the relevant temperature region. These coefficients have been determined. A model based on these coefficients has been established which, apart from best estimate deformation, gives upper and lower bounds of possible deformation. The model derived from isothermal uniaxial tests is being verified against isothermal and transient tube burst tests. The influence of preoxidation and increased oxygen concentration during deformation is modeled on the basis of the pseudobinary Zircaloy-oxygen phase diagram. (author)

  1. Effects of deposited pyrolytic carbon on some mechanical properties of zircaloy-4 tubes. Vol. 3

    Energy Technology Data Exchange (ETDEWEB)

    Shrkawy, S W; Abdel-razek, I D; El-Sayed, H A [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    Zircaloy cladding tubes are not compatible with the uranium fuel pellets as they suffer from failure due to pelletclad interaction (PCI). A carbon coating, as used in the canadian CANLUB fuel elements, is thought to improve the cladding performance with respect to the PCI problem. In this paper pyrolytic carbon coating was deposited on zircaloy-4 cladding tubes by the thermal cracking of commercial butant gas at the temperature range 250-450 degree C. In order to evaluate the effect of gaseous species on the mechanical properties of the tubes tensile and microhardness testing measurements were performed on samples prepared from the coated tubes. The fractured surface of the tensile zircaloy tubes and the deposited carbon coating, both, were examined by the SEM. The results of the tensile tests of zircaloy-4 tubes indicated that the coating process has insignificant effect on the ultimate strength of the tubes tested. The values of Vickers hardness numbers were not significantly changed across the tubes thickness. The microstructure of deposited carbon, due to the cracking process, was granular in all the temperature range (250-450 degree C) studied. 9 figs., 1 tab.

  2. 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

  3. Some aspects of the utilization of zicaloy and austenitic steel as cladding material for PWR reactor fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Perrotta, J.A.

    1985-01-01

    The behaviour under irradiation of fuel rods for light water reactors was simulated by using fuel performance codes. Two types of cladding were analyzed: zircaloy and austenitic stainless steel. The fuel performance codes, originally made for zircaloy cladding, were adapted for austenitic stainless steel. The simulation results for the two types of cladding are presented, compared and discussed. (F.E.) [pt

  4. Effects of oxidation in the mechanical behavior of zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Santos, A.M.M. dos.

    1981-07-01

    The kinetics of oxidation of zircaloy-4 is isothermally studied utilizing discontinous gravimetric method under two different oxidizing conditions, using gaseous oxigen and steam. The total weight gain during oxidation occurs in two different way: formation of oxide and solid solution. A mechanical test for studying the effect of embrittlement due to the absorption of oxygen in small zircalloy tubes have been developed. (Author) [pt

  5. Development of processes for zircaloy chips recycling by electric arc furnace remelting and powder metallurgy

    International Nuclear Information System (INIS)

    Pereira, Luiz Alberto Tavares

    2014-01-01

    PWR reactors employ, as nuclear fuel, UO 2 pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopy. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering. (author)

  6. The oxidation kinetics and the structure of the oxide film on Zircaloy before and after the kinetic transition

    International Nuclear Information System (INIS)

    Arima, T.; Masuzumi, T.; Furuya, H.; Idemitsu, K.; Inagaki, Y.

    2001-01-01

    Oxidation kinetics of Zircaloy-4 have been measured using a micro-balance technique in CO-CO 2 gas mixtures between 450 deg. C and 600 deg. C. Oxidation kinetics of Zircaloy-4 obeyed a cubic rate law with time at 450-600 deg. C up to 24 h. At 600 deg. C, the kinetic transition occurred after about 36 h. After the transition, oxidation kinetics obeyed a linear rate law. X-ray diffraction patterns for the samples oxidized at 600 deg. C showed that the volume fraction of tetragonal phase of zirconia decreased with time until the kinetic transition occurred and was almost constant after that. In addition, stresses in the oxide films were found to be larger for the pre-transition samples than for the post-transition ones. (authors)

  7. Development of processes for zircaloy chips recycling by electric arc furnace remelting and powder metallurgy; Desenvolvimento de processos de reciclagem de cavacos de zircaloy via refusao em forno eletrico a arco e metalurgia do po

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Luiz Alberto Tavares

    2014-09-01

    PWR reactors employ, as nuclear fuel, UO{sub 2} pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopy. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering. (author)

  8. 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

  9. Application of Coating Technology for Accident Tolerant Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  10. 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)

  11. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    International Nuclear Information System (INIS)

    Chung, H.M.

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 μm in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307 degree C rather than the normal 288 degree C, a relatively thick (50 to 70 μm) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs

  12. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

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

  13. SiC-CMC-Zircaloy-4 Nuclear Fuel Cladding Performance during 4-Point Tubular Bend Testing

    Energy Technology Data Exchange (ETDEWEB)

    IJ van Rooyen; WR Lloyd; TL Trowbridge; SR Novascone; KM Wendt; SM Bragg-Sitton

    2013-09-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE NE) established the Light Water Reactor Sustainability (LWRS) program to develop technologies and other solutions to improve the reliability, sustain the safety, and extend the life of current reactors. The Advanced LWR Nuclear Fuel Development Pathway in the LWRS program encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. Recent investigations of potential options for “accident tolerant” nuclear fuel systems point to the potential benefits of silicon carbide (SiC) cladding. One of the proposed SiC-based fuel cladding designs being investigated incorporates a SiC ceramic matrix composite (CMC) as a structural material supplementing an internal Zircaloy-4 (Zr-4) liner tube, referred to as the hybrid clad design. Characterization of the advanced cladding designs will include a number of out-of-pile (nonnuclear) tests, followed by in-pile irradiation testing of the most promising designs. One of the out-of-pile characterization tests provides measurement of the mechanical properties of the cladding tube using four point bend testing. Although the material properties of the different subsystems (materials) will be determined separately, in this paper we present results of 4-point bending tests performed on fully assembled hybrid cladding tube mock-ups, an assembled Zr-4 cladding tube mock-up as a standard and initial testing results on bare SiC-CMC sleeves to assist in defining design parameters. The hybrid mock-up samples incorporated SiC-CMC sleeves fabricated with 7 polymer impregnation and pyrolysis (PIP) cycles. To provide comparative information; both 1- and 2-ply braided SiC-CMC sleeves were used in this development study. Preliminary stress simulations were performed using the BISON nuclear fuel performance code to show the stress distribution differences for varying lengths between loading points

  14. 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

  15. 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.

  16. Investigation of effect of air flow rate on Zircaloy-4 oxidation kinetics and breakaway phenomenon in air at 850 .deg. C

    International Nuclear Information System (INIS)

    Maeng, Yunhwan; Lee, Jaeyoung; Park, Sanggil

    2016-01-01

    This paper analyzed an effect of flow rate on oxidation kinetics of Zircaloy-4 in air at 850 .deg. C. In case of the oxidation of Zircaloy-4 in air at 850 .deg. C, acceleration of oxidation kinetics from parabolic to linear (breakaway phenomenon) occurs. Oxidation and breakaway kinetics of the Zircaloy-4 in air was experimentally studied by changing a flow rate of argon/air mixture. Tests were conducted at 850 .deg. C under constant ratio of argon and air. The effects of flow rate on the oxidation and breakaway kinetics was observed. This paper is based on a revised and considerably extended presentation given at the 21 st International Quench Workshop. The effects of flow conditions on the oxidation kinetics of Zircaloy-4 samples were explained with residence time and percent flow efficiency. In addition, several issues were observed from this study, interdiffusion at breakaway and deformation of oxide structure by breakaway phenomenon

  17. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  18. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

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

    1975-10-01

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

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

    International Nuclear Information System (INIS)

    Zmitko, M.

    2002-01-01

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

  20. Zirconium metal-water oxidation kinetics. V. Oxidation of Zircaloy in high pressure steam

    International Nuclear Information System (INIS)

    Pawel, R.E.; Cathcart, J.V.; Campbell, J.J.; Jury, S.H.

    1977-12-01

    A series of scoping tests to determine the influence of steam pressure on the isothermal oxidation kinetics of Zircaloy-4 PWR tubing was undertaken. The oxidation experiments were conducted in flowing steam at 3.45, 6.90, and 10.34 MPa (500, 1000, and 1500 psi) at 905 0 C (1661 0 F), and at 3.45 and 6.90 MPa at 1101 0 C (2014 0 F). A comparison of the results of these experiments with those obtained for oxidation in steam at atmospheric pressure under similar conditions indicated that measurable enhancement of the oxidation rate occurred with increasing pressure at 905 0 C, but not at 1100 0 C

  1. 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

  2. Influence of foreign matter on the flammability of Zircaloy

    International Nuclear Information System (INIS)

    Praetorius, R.; Muenzel, H.

    1990-01-01

    When cutting Zircaloy cladding in the head end of a reprocessing plant, fine particles with a high chemical reactivity are produced. Spontaneous ignition may cause fire or dust explosion. Therefore their ignition and fire behaviour was studied. As a result it can be stated that sugar or a concentrated sugar solution (syrup) poured over a Zircaloy fire is particularly suited as a fire-extinguishing agent. The developing caramel melt prevents air access and sparking. In addition, the sugar can be washed out easily before cementing, and so additional waste arising can be avoided. (DG) [de

  3. Thermomechanical behavior and modeling of zircaloy cladding tubes from an unirradiated state to high burn-up

    International Nuclear Information System (INIS)

    Schaeffler-Le Pichon, I.; Geyer, P.; Bouffioux, P.

    1997-01-01

    Creep laws are nowadays commonly used to simulate the fuel rod response to the solicitations it faces during its life. These laws are sufficient for describing the base operating conditions (where only creep appears), but they have to be improved for power ramp conditions (where hardening and relaxation appear). The modification due to a neutronic irradiation of the thermomechanical behavior of stress-relieved Zircaloy 4 fuel tubes that have been analysed for five different fluences ranging from a non-irradiated material to a material for which the combustion rate was very high is presented. In the second part, a viscoplastic model able to simulate, for different isotherms, out-of-flux anisotropic mechanical behavior of the cladding tubes irradiated until high burn-up is proposed. Finally, results of numerical simulations show the ability of the model to reproduce the totality of the thermomechanical experiments. (author)

  4. Charpy impact test of oxidized and hydrogenated zircaloy using a thin strip specimen

    International Nuclear Information System (INIS)

    Otsuka, Teppei; Hashizume, Kenichi; Sugisaki, Masayasu

    2004-01-01

    The impact properties of an oxidized and a hydrogenated Zircaloy have been studied with an instrumented Charpy machine by using a strip Charpy V-notch specimen (1 mm thick by 4mm wide). Fracture processes such as crack initiation and propagation were examined using load-displacement curves obtained in this study. In the case of the hydrogenated specimen containing preferentially oriented hydrides, an appreciable decrease in the absorbed energy was observed in the crack propagation rather than in the crack initiation. From results of fractographs of the specimen, it was suggested that the reduction of the crack propagation energy of hydrogenated specimen could be attributed to the change of the stress state in the Zircaloy matrix, which was caused by the fracture of hydride in the inner part of specimen. In the case of the specimen oxidized at 973k for 60 min, on which an oxide layer (4 μm in thickness) and oxygen incursion layer (4μm) were formed, the surface layers affected the crack initiation process. The growing oxygen incursion layer, in particular, resulted in the constraint of plastic deformation of the Zircaloy matrix not only in the crack initiation but also in the crack propagation as its thickness increased. (author)

  5. In-cell facility for performing mechanical-property tests on irradiated cladding

    International Nuclear Information System (INIS)

    Yaggee, F.L.; Haglund, R.C.; Mattas, R.F.

    1978-11-01

    A new facility was developed for testing cladding sections of LWR fuel rods. This facility and the accompanying test procedures have improved the level of in-cell mechanical-testing capabilities, making them comparable to existing capabilities for unirradiated cladding. The new facility is currently being used to study the susceptibility of irradiated Zircaloy cladding from LWR fuel rods to iodine stress-corrosion cracking. Preliminary testing results indicate a systematic effect of temperature, stress and irradiation on the susceptibility of annealed and stress-relieved Zircaloy-2. Experimental data obtained to date are being used to develop a stress-corrosion cracking model for LWR fuel rod failure. SEM examination of the undisturbed fracture surface of specimens that failed by pinhole leakage provides useful information on crack propagation and morphology

  6. Critical heat flux on micro-structured zircaloy surfaces for flow boiling of water at low pressures

    International Nuclear Information System (INIS)

    Haas, C.; Miassoedov, A.; Schulenberg, T.; Wetzel, T.

    2012-01-01

    The influence of surface structure on critical heat flux for flow boiling of water was investigated for Zircaloy tubes in a vertical annular test section. The objectives were to find suitable surface modification processes for Zircaloy tubes and to test their critical heat flux performance in comparison to the smooth tube. Surface structures with micro-channels, porous layer, oxidized layer, and elevations in micro- and nano-scale were produced on a section of a Zircaloy cladding tube. These modified tubes were tested in an internally heated vertical annulus with a heated length of 326 mm and an inner and outer diameter of 9.5 and 18 mm. The experiments were performed with mass fluxes of 250 and 400 kg/(m 2 s), outlet pressures between 120 and 300 kPa, and constant inlet subcooling enthalpy of 167 kJ/kg. Only a small influence of modified surface structures on critical heat flux was observed for the pressure of 120 kPa in the present test section geometry. However, with increasing pressure the critical heat flux could increase up to 29% using the surface structured tubes with micro-channels, porous and oxidized layers. Capillary effects and increased nucleation site density are assumed to improve the critical heat flux performance. (authors)

  7. Biaxial mechanical tests in zircaloy-4

    International Nuclear Information System (INIS)

    Mintzer, S.R.; Bordoni, R.A.A.; Falcone, J.M.

    1980-01-01

    The texture of the zircaloy-4 tubes used as cladding in nuclear fuel elements determines anisotropy of the mechanical properties. As a consequence, the uniaxial tests to determine the mechanical behaviour of the tubes are incomplete. Furthermore, the cladding in use is subject to creep with a state of biaxial tensions. For this reason it is also important to determine the biaxial mechanical properties. The creep tests were performed by internal pressure for a state of axial to circumferential tensions of 0.5. Among the experimental procedures are described: preparation of the test specimens, pressurizing equipment, and the implementation of a device that permits a permanent register of the deformation. For the non-irradiated Atucha type zircaloy-4 sheaths, experimental curves of circumferential deformation versus time were obtained, in tests at constant pressure and for different values of temperature and pressure. An empirical function was determined to adjust the experimental values for the speed of the circumferential deformation in terms of the initial tension applied, temperature and deformation, and the change of the corresponding parameters in accordance to the range of the tensions. Also the activation energy for creep was determined. (M.E.L.) [es

  8. Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Katanishi, S.

    1987-03-01

    This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO 2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL) [de

  9. Zircaloy-steam reaction under a simulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kawasaki, Satoru; Furuta, Teruo; Hashimoto, Masao

    1975-07-01

    Under a simulated loss-of-coolant condition, the reaction between zircaloy and steam and the embrittlement of the zircaloy oxidized by this reaction have been studied. The parabolic rate constant, ksub(p), in the zircaloy-steam reaction is represented as ksub(p)=3.24x10 6 exp(-40500/RT) (mg 2 /cm 4 . sec) Ring compression test was made on the steam-reacted zircaloy tubes, and following results were obtained: Embrittlement of the steam-reacted zircaloy tube increases with oxidation at each oxidation temperature. For a given quantity of the oxidation, the incursion of α-phase into β-phase is more remarkable in the specimens reacted at low temperatures than those at high temperatures. The embrittlement, however, is larger in the specimens oxidized at high temperatures than those at low temperatures. (auth.)

  10. Modification of hydrogen determinator for total hydrogen analysis in irradiated zircaloy cladding tube

    International Nuclear Information System (INIS)

    Park, Soon Dal; Choi, Kwnag Soon; Kim, Jong Goo; Joe, Kih Soo; Kim, Won Ho

    1999-01-01

    A hydrogen determinator was modified and installed in the glove box to analyse total hydrogen content in irradiated zircaloy tube. The analysis method of hydrogen is Inert Gas Fusion(IGF)-Thermal Conductivity Detection(TCD). The hydrogen recoveries of no tin method using Ti and Zr matrix standards, respectively, were available within 3 μg of hydrogen. Also the smaller size of sample showed the better hydrogen recovery. It was found that the hydrogen standard of Ti matrix is available to hydrogen analysis in zircaloy sample. The mean radioactivity of irradiated zircaloy sample was 10 mR/hr and hydrogen concentration was 130 ppm

  11. Thermal diffusion of hydrogen in zircaloy-2 containing hydrogen beyond terminal solid solubility

    International Nuclear Information System (INIS)

    Maki, Hideo; Sato, Masao.

    1975-01-01

    The thermal diffusion of hydrogen is one of causes of uneven hydride precipitation in zircaloy fuel cladding tubes that are used in water reactors. In the diffusion model of hydrogen in zircaloy, the effects of the hydride on the diffusibility of hydrogen has been regarded as negligibly small in comparison with that of hydrogen dissolved in the matrix. Contrary to the indications given by this model, phenomena are often encountered that cannot be explained unless hydride platelets have considerable ostensible diffusibility in zircaloy. In order to determine quantitatively the diffusion characteristics of hydrogen in zircaloy, a thermal diffusion experiment was performed with zircaloy-2 fuel cladding tubes containing hydrogen beyond the terminal solid solubility. In this experiment, a temperature difference of 20 0 --30 0 C was applied between the inside and outside surfaces of the specimen in a thermal simulator. To explain the experimental results, a modified diffusion model is presented, in which the effects of stress are introduced into Markowitz's model with the diffusion of hydrogen in the hydride taken into account. The diffusion equation derived from this model can be written in a form that ostensibly represents direct diffusion of hydride in zircaloy. The apparent diffusion characteristics of the hydride at around 300 0 C are Dsub(p)=2.3x10 5 exp(-32,000/RT), (where R:gas constant, T:temperature) and the apparent heat of transport Qsub(p) =-60,000 cal/mol. The modified diffusion model well explains the experimental results in such respects as reaches a steady state after several hours. (auth.)

  12. 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.

  13. Temperature escalation in PWR fuel rod simulator bundles due to the Zircaloy/steam reaction: Test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

    1984-07-01

    This report describes the test conduct and results of the bundle test ESBU-2A, which was run to investigate the temperature escalation of zircaloy clad fuel rods. This investigation of temperature escalation is part of a series of out-of-pile experiments, performed within the framework of the PNS Severe Fuel Damage Program. The test bundle was of a 3 x 3 array of fuel rod simulators with a 0.4 m heated length. The fuel rod simulators were electrically heated and consisted of tungsten heaters, UO 2 annular pellets, and zircaloy cladding. A nominal steam flow of 0.7 g/s was inlet to the bundle. The bundle was surrounded by a zircaloy shroud which was insulated with ZrO 2 fiber ceramic wrap. The initial heatup rate of the bundle was 0.4 0 C/s. The temperature escalation began at the 255 mm elevation after 1200 0 C had been reached. At this elevation, the measured peak temperature was limited to 1500 0 C. It was concluded from different thermocouple results, that induced by this first escalation melt was formed in the lower part of the bundle. Consequently, the escalation in the lower part must be much higher, at least up to the melting temperature of zircaloy. Due to the failure in the steam production system, steam starvation in the upper region may explain the beginning of the escalation at the 255 mm elevation. The maximum temperature reached was 2175 0 C on the center rod at the end of the test. The unregularities in the steam supply may be the reason for less oxidation than expected. (orig./GL) [de

  14. Comparison of two analytical methods for the local quantitative determination of lithium and boron contents in cladding materials

    International Nuclear Information System (INIS)

    Gavillet, D.; Guenther-Leopold, I.; Martin, M.; Guillong, M.; Hellwig, Ch.; Sell, H.J.

    2008-01-01

    Pressurized water reactors contain boric acid for reactivity control. As the acidic coolant conditions result in an increased attack of the circuit materials, LiOH is added to render the coolant slightly alkaline. However, LiOH can affect corrosion of the Zr alloy cladding. Thus the Li content in the oxide layers of irradiated fuel rods is of high interest, especially for new alloys (pathfinder rods). At the 'Paul Scherrer Institut' the lithium as well as the boron content in the oxide layers of claddings are determined by Secondary Ion Mass Spectrometry (SIMS). Quantification is performed by direct comparison with a Zircaloy-oxide layer implanted with B and Li. A new and independent method using Laser Ablation Inductively Coupled Plasma Mass Spectrometry was applied to cross-check the SIMS data. (authors)

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

    International Nuclear Information System (INIS)

    Chapman, R.H.

    1976-01-01

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

  16. 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

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  19. In-situ electrochemical impedance spectroscopy measurements of zirconium alloy oxide conductivity: Relationship to hydrogen pickup

    International Nuclear Information System (INIS)

    Couet, Adrien; Motta, Arthur T.; Ambard, Antoine; Livigni, Didier

    2017-01-01

    Highlights: • In-situ electrochemistry on zirconium alloys in 360 °C pure water show oxide layer resistivity changes during corrosion. • A linear relationship is observed between oxide resistivity and instantaneous hydrogen pickup fraction. • The resistivity of the oxide layer formed on Zircaloy-4 (and thus its hydrogen pickup fraction) is higher than on Zr-2.5Nb. - Abstract: Hydrogen pickup during nuclear fuel cladding corrosion is a critical life-limiting degradation mechanism for nuclear fuel. Following a program dedicated to zirconium alloys, corrosion, it has been hypothesized that oxide electronic resistivity determines hydrogen pickup. In-situ electrochemical impedance spectroscopy experiments were performed on Zircaloy-4 and Zr-2.5Nb alloys in 360 °C water. The oxide resistivity was measured as function of time. The results show that as the oxide resistivity increases so does the hydrogen pickup fraction. The resistivity of the oxide layer formed on Zircaloy-4 is higher than on Zr-2.5Nb, resulting in a higher hydrogen pickup fraction of Zircaloy-4, compared to Zr-2.5Nb.

  20. 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

  1. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  2. Corrosion behavior of Zircaloy 4 cladding material. Evaluation of the hydriding effect

    International Nuclear Information System (INIS)

    Blat, M.

    1997-04-01

    In this work, particular attention has been paid to the hydriding effect in PIE and laboratory test to validate a detrimental hydrogen contribution on Zircaloy 4 corrosion behavior at high burnup. Laboratory corrosion tests results confirm that hydrides have a detrimental role on corrosion kinetics. This effect is particularly significant for cathodic charged samples with a massive hydride outer layer before corrosion test. PIE show that at high burnup a hydride layer is formed underneath the metal/oxide interface. The results of the metallurgical examinations are discussed with respect to the possible mechanisms involved in this detrimental effect of hydrogen. Therefore, according to the laboratory tests results and PIE, hydrogen could be a strong contributor to explain the increase in corrosion rate at high burnup. (author)

  3. Influence of hydrogen on the oxygen solubility in Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Guilbert-Banti, Séverine, E-mail: severine.guilbert@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France); Lacote, Pauline; Taraud, Gaëlle [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France); Berger, Pascal [NIMBE, CEA, CNRS, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Desquines, Jean; Duriez, Christian [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France)

    2016-02-15

    Despite the influence of hydrogen on the behavior of zirconium fuel cladding in many nuclear safety issues, the pseudo-binary Zircaloy-4 – oxygen phase diagram still lacks of data, especially above 1000 °C. The aim of this study was to provide experimental data to better assess the influence of hydrogen on the oxygen solubility in Zircaloy-4. Homogenized two-phase Zircaloy-4 samples were elaborated from 300 to 1000 wppm pre-hydrided samples. Local distributions were characterized thoroughly using Electron Probe Micro-Analysis (EPMA) for oxygen and Elastic Recoil Detection Analysis (ERDA) for hydrogen. The data obtained in this work were included in the pseudo-binary Zircaloy-4 – oxygen phase diagram and have shown that hydrogen has limited influence on the α + β → β transus. Regarding the α → α + β transus, no influence of hydrogen concentration in the α phase below 400 wppm was evidenced.

  4. 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)

  5. Quantification and characterization of zirconium hydrides in Zircaloy-4 by the image analysis method

    International Nuclear Information System (INIS)

    Zhang, J.H.; Groos, M.; Bredel, T.; Trotabas, M.; Combette, P.

    1992-01-01

    The image analysis method is used to determine the hydrogen content in specimens of Zircaloy-4. Two parameters, surface density of hydride, S v , and degree of orientation, Ω, are defined to represent separately the hydrogen content and the orientation of hydrides. By analysing the stress-relieved Zircaloy-4 specimens with known hydrogen content from 100 to 1000 ppm, a relationship is established between the parameter S v and the hydrogen content when the magnifications of the optical microscope are 1000 and 250. The degree of orientation for the hydride in the stress-relieved Zircaloy-4 cladding is about 0.3. (orig.)

  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

    cladding environment. Our interpretation of the sub-cooled nucleate boiling impact on the PWR fuel cladding's chemical environment explains the most important change of the fuel cladding's environmental characteristics - the creation of oxidizing conditions. This conclusion requires new criteria for corrosion resistance properties of fuel cladding materials. These materials must have high corrosion resistance by oxidizing conditions. The presence of oxidizing conditions in PWR fuel cladding's environment explains also in practice observed fact that the nickel oxide formation dominates over the metallic nickel in corrosion product depositions on fuel elements' cladding. Our conclusions offer a logically explanation of prime root cause for Axial Offset Anomaly in US Westinghouse NPPs - the lower corrosion resistance of Zircaloy-4 cladding materials by oxidizing conditions occurring by sub-cooled nucleate boiling. (authors)

  7. CANDU fuel sheath integrity and oxide layer thickness determination by Eddy current technique

    International Nuclear Information System (INIS)

    Gheorghe, Gabriela; Man, Ion; Parvan, Marcel; Valeca, Serban

    2010-01-01

    This paper presents results concerning the integrity assessment of the fuel elements cladding and measurements of the oxide layer on sheaths, using the eddy current technique. Flaw detection using eddy current provides information about the integrity of fuel element sheath or presence of defects in the sheath produced by irradiation. The control equipment consists of a flaw detector with eddy currents, operable in the frequency range 10 Hz to 10 MHz, and a differential probe. The calibration of the flaw detector is done using artificial defects (longitudinal, transversal, external and internal notches, bored and unbored holes) obtained on Zircaloy-4 tubes identical to those out of which the sheath of the CANDU fuel element is manufactured (having a diameter of 13.08 mm and a wall thickness of 0.4 mm). When analyzing the behavior of the fuel elements' cladding facing the corrosion is important to know the thickness of the zirconium oxide layer. The calibration of the device measuring the thickness of the oxide layer is done using a Zircaloy-4 tube identical to that which the cladding of the CANDU fuel element is manufactured of, and calibration foils, as well. (authors)

  8. Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1979-03-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel, and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCI far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel > 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  10. Sensitivity of measured steam oxidation kinetics to atmospheric control and impurities

    Energy Technology Data Exchange (ETDEWEB)

    Sooby Wood, E., E-mail: sooby@lanl.gov [Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM (United States); Terrani, K.A. [Nuclear Fuels Materials Group, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Nelson, A.T. [Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-08-15

    The most direct means of improving the ability of water cooled reactors to withstand excessive cladding oxidation during a loss of coolant accident is to either modify or replace zirconium cladding. It is important to understand what level of agreement is to be expected as a function of systematic differences in steam oxidation testing techniques and instrumentation among testing facilities. The present study was designed to assess the sensitivities of some of the current and proposed reactor cladding materials. Steam oxidation sensitivity of Zircaloy-2, FeCrAl and Mo to O{sub 2} impurities in steam were examined. It was shown that the effect of O{sub 2} impurities is negligible for the two former materials while significant in the case of Mo.

  11. Iodine induced stress corrosion cracking of zircaloy cladding tubes

    International Nuclear Information System (INIS)

    Brunisholz, L.; Lemaignan, C.

    1984-01-01

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

  12. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    Science.gov (United States)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  13. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    International Nuclear Information System (INIS)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  14. Variation in the strain anisotropy of Zircaloy with temperature and strain

    International Nuclear Information System (INIS)

    Hindle, E.D.; Worswick, D.

    1984-01-01

    The strong crystallographic texture which is developed during the fabrication of zirconium-based alloys causes pronounced anisotropy in their mechanical properties, particularly deformation. The tendency for circular-section tension specimens with a high concentration of basal poles in one direction to become elliptical when deformed in tension has been used in this study to provide quantitative data on the effects of both strain and temperature on strain anisotropy. Tension tests were carried out over a temperature range of 293 to 1193 K on specimens machined from Zircaloy-2 plate. The strain anisotropy was found to increase markedly at temperatures over 923 K, reaching a maximum in the region of 1070 K. The strain anisotropy increased with increasing strain in this temperature region. The study was extended to Zircaloy-4 pressurized-water reactor fuel cladding by carrying out tube swelling tests and evaluating the axial deformation produced. Although scatter in the test results was higher than that exhibited in the tension tests, the general trend in the data was similar. The effects of the strain anisotropy observed are discussed in relation to the effects of temperature on the ductility of Zircaloy fuel cladding tubes during postulated largebreak loss-of-coolant accidents

  15. 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

  16. Hydriding and neutron irradiation in zircaloy-4

    International Nuclear Information System (INIS)

    Ramos, Ruben Fortunato; Martin, Juan Ezequiel; Orellano, Pablo; Dorao, Carlos; Analia Soldati; Ghilarducci, Ada Albertina; Corso, Hugo Luis; Peretti, Hernan Americo; Bolcich, Juan Carlos

    2003-01-01

    The composition of Zircaloy-4 for nuclear applications is specified by the ASTM B350 Standard, that fixes the amount of alloying elements (Sn, Fe, Cr) and impurities (Ni, Hf, O, N, C, among others) to optimize good corrosion and mechanical behavior.The recycling of zircaloy-4 scrap and chips resulting from cladding tube fabrication is an interesting issue.However, changes in the final composition of the recycled material may occur due to contamination with tool pieces, stainless steel chips, turnings, etc. while scrap is stored and handled. Since the main components of the possible contaminants are Fe, Cr and Ni, it arises the interest in studying up to what limit the Fe, Ni and Cr contents could be exceeded beyond the standard specification without affecting significantly the alloy properties.Zircaloy-4 alloys elaborated with Fe, Cr and Ni additions and others of standard composition in use in nuclear plants are studied by tensile tests, SEM observations and EDS microanalysis.Some samples are tested in the initial condition and others after hydriding treatments and neutron irradiation in the RA6

  17. Treatment of cladding hulls by the HIPOW process

    International Nuclear Information System (INIS)

    Larker, H.T.; Tegman, R.

    1981-01-01

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

  18. The accelerated oxidation of zircaloy-4 at 700∼900 .deg. C in high pressure steam

    International Nuclear Information System (INIS)

    Kim, K. P.; Park, K. H.

    1999-01-01

    To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The specimens used in experiments are commercially available Zircaloy-4 used in Kori nuclear power plants. All the measurements were done at 700∼900 .deg. C in steam. Pressure effects were noticed. The oxide thickness was much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. The enhancement of oxide growth rate at 700∼900 .deg. C in high pressure steam is approximately propotion to the power of 1.0∼1.6 of the ratio of experimental steam pressure to critical steam pressure. There is a critical steam pressure above that the oxidation rate enhances. The critical steam pressure was measured as 30∼40 bar. The enhanced oxidation seems from the oxide cracking due to the tetragonal to monoclinic phase transformation at high pressure steam

  19. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1980-11-01

    The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

  20. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

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

    2016-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  2. The characteristics of surface oxidation and corrosion resistance of nitrogen implanted zircaloy-4

    International Nuclear Information System (INIS)

    Tang, G.; Choi, B.H.; Kim, W.; Jung, K.S.; Kwon, H.S.; Lee, S.J.; Lee, J.H.; Song, T.Y.; Shon, D.H.; Han, J.G.

    1997-01-01

    This work is concerned with the development and application of ion implantation techniques for improving the corrosion resistance of zircaloy-4. The corrosion resistance in nitrogen implanted zircaloy-4 under a 120 keV nitrogen ion beam at an ion dose of 3 x 10 17 cm -2 depends on the implantation temperature. The characteristics of surface oxidation and corrosion resistance were analyzed with the change of implantation temperature. It is shown that as implantation temperature rises from 100 to 724 C, the colour of specimen surface changes from its original colour to light yellow at 100 C, golden at 175 C, pink at 300 C, blue at 440 C and dark blue at 550 C. As the implantation temperature goes above 640 C, the colour of surface changes to light black, and the surface becomes a little rough. The corrosion resistance of zircaloy-4 implanted with nitrogen is sensitive to the implantation temperature. The pitting potential of specimens increases from 176 to 900 mV (SCE) as the implantation temperature increases from 100 to 300 C, and decreases from 900 to 90 mV(SCE) as the implantation temperature increases from 300 to 640 C. The microstructure, the distribution of oxygen, nitrogen and carbon elements, the oxide grain size and the feature of the precipitation in the implanted surface were investigated by optical microscope, TEM, EDS, XRD and AES. The experimental results reveal that the ZrO 2 is distributed mainly on the outer surface. The ZrN is distributed under the ZrO 2 layer. The characteristics of the distribution of ZrO 2 and ZrN in the nitrogen-implanted zircaloy-4 is influenced by the implantation temperature of the sample, and in turn the corrosion resistance is influenced. (orig.)

  3. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  4. 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)

  5. A pneumatic bellows-driven setup for controlled-distance electrochemical impedance measurements of Zircaloy-2 in simulated BWR conditions

    International Nuclear Information System (INIS)

    Arilahti, E.; Bojinov, M.; Hansson-Lyyra, L.

    2004-01-01

    This paper describes a novel pneumatic bellows-driven arrangement designed for controlled distance electrochemistry (CDE) measurements. The feasibility of the new arrangement has been verified by performing contact electric impedance measurements to study corrosion of Zircaloy-2 in a re-circulation loop simulating the BWR conditions. Until now, the measurements have been carried out using a step-motor driven controlled-distance electrochemistry (CDE) arrangement. The electrical and electrochemical properties of the pre transition oxide on Zircaloy-2 determined from these measurements were in good agreement with those estimated from measurements with a step-motor driven CDE. Furthermore, the results indicate that the bellows-driven CDE device is less sensitive to the contact pressure variation than the step-motor driven arrangement. This property combined with the bellows driven displacement mechanism provides a clear advantage for future in-core corrosion studies of fuel cladding materials. (Author)

  6. 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.

  7. SIMS and TEM study on oxide characteristics of Zircaloy

    International Nuclear Information System (INIS)

    Jung, Y. H.; Baek, J. H.; Kim, S. J.; Kim, K. H.; Choi, B. K.; Jung, Y. H.

    1998-01-01

    Long-term corrosion test, SIMS analysis, and TEM study were carried out to investigate the corrosion characteristics and corrosion mechanism of Zircaloy-4 in LiOH solution. The corrosion tests were performed in alkali solutions at 350 deg C for 500days. SIMS analysis was performed for the specimens prepared to have an equal oxide thickness to measure the cation content. TEM studies on the samples formed in various alkali solutions were also conducted. Based on the corrosion test, SIMS analysis, and TEM study, the cation is considered to control the corrosion in LiOH solution and its effect is dependent on the concentration of alkali and the oxide thickness. The slight acceleration of corrosion rate at a low concentration is thought to be caused by the cation incorporation into oxide while the significant acceleration at a high concentration is due to the transformation of oxide microstructure that would be induced by the cation incorporation

  8. 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)

  9. Simulated LOCA Test and Characterization Study Related to High Burn-Up Issue

    International Nuclear Information System (INIS)

    Park, D. J.; Jung, Y. I.; Choi, B. K.; Park, S. Y.; Kim, H. G.; Park, J. Y.

    2012-01-01

    For the safety evaluation of fuel cladding during the injection of emergency core coolant, simulated Loss-of-coolant accident (LOCA) test was performed by using Zircaloy-4 fuel cladding samples. Zircaloy-4 tube samples with and without prehydring were oxidized in a steam environment with the test temperature of 1200 .deg. C. Prehydrided cladding was prepared from as-fabricated Zircaloy-4 to study the effects of hydrogen on mechanical properties of cladding during high temperature oxidation and quench conditions. In order to measure the ductility of the tube samples embrittled by quenching water, ring compression test was carried out by using 8 mm ring sample sectioned from oxidized tube sample and microstructural analysis was also performed after simulated LOCA test. The results showed that hydrogen increases oxygen solubility and pickup rate in the beta layer. This reduces ductility of prehydrided fuel cladding compared with as-fabricated cladding. Trend in ductility decrease for prehydrided sample under simulated LOCA condition was very similar with data obtained from tests conducted using irradiated high burn-up fuel claddings

  10. Mechanical behavior of zircaloy-4 tubes under complexe state of stress

    International Nuclear Information System (INIS)

    Costa Viana, C.S. da

    1980-01-01

    The use of zircaloy-4 tubing as cladding material for fuel elements is reviewed with respect to its microstructural, textural and loading conditions. Its anisotropic plastic behaviour is studied through the experimental determination of its yield locus by mechanical testing and Knoop hardness and compared to Hill's anisotropic yield criterion. (Author) [pt

  11. Strengthening of Zircaloy-4 with Oxide Particles by Surface Treatment using Laser Beam

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang Il; Park, Jung Hwan; Park, Dong Jun; Kim, Hyun Gil; Yang, Jae Ho; Koo, Yang Hyun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Accident tolerant fuel (ATF) cladding is being developed globally after the Fukushima accident with the demands for the nuclear fuel having higher safety at normal operation conditions as well as even in a severe accident conditions. Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. High-power laser beam was exposed on the zirconium surface previously coated by oxides. Various oxides such as Y{sub 2}O{sub 3}, CeO{sub 2}, Gd{sub 2}O{sub 3}, Er{sub 2}O{sub 3} were used for the ODS treatment. In this study, the effect of strengthening by the ODS treatment was investigated. The oxide particles of Y{sub 2}O{sub 3} were dispersed well in the Zr matrix at the surface region.

  12. Influence of neutron irradiation on the stability of recipitates in zircaloy: a critical review

    International Nuclear Information System (INIS)

    Lobo, Raquel M.; Andrade, Arnaldo H. P.

    2013-01-01

    The realization of RMB enterprise (Brazilian Multipurpose Reactor) will give the country a powerful tool to investigate the behavior materials subjected to irradiation. Among them, zirconium alloys, used as cladding of nuclear fuel in reactors type LWR. It is know that neutron irradiation can affect the stability of precipitates in zircaloys, generating as a result changes in theirs mechanical properties, important application of this alloys. This paper present a critical review of neutron irradiation effects on microstructural stability of zircaloys (2 and 4). (author)

  13. Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

    International Nuclear Information System (INIS)

    Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.

    1977-01-01

    A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

  14. 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)

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

    International Nuclear Information System (INIS)

    Siefken, L.J.

    1999-01-01

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

  16. Influence of temperature and hydrogen content on stress-induced radial hydride precipitation in Zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Desquines, J., E-mail: jean.desquines@irsn.fr; Drouan, D.; Billone, M.; Puls, M.P.; March, P.; Fourgeaud, S.; Getrey, C.; Elbaz, V.; Philippe, M.

    2014-10-15

    Radial hydride precipitation in stress relieved Zircaloy-4 fuel claddings is studied using a new thermal–mechanical test. Two maximum temperatures for radial hydride precipitation heat treatment are studied, 350 and 450 °C with hydrogen contents ranging between 50 and 600 wppm. The new test provides two main results of interest: the minimum hoop stress required to precipitate radial hydrides and a maximum stress above which, all hydrides precipitate in the radial direction. Based on these two extreme stress conditions, a model is derived to determine the stress level required to obtain a given fraction of radial hydrides after high temperature thermal–mechanical heat treatment. The proposed model is validated using metallographic observation data on pressurized tubes cooled down under constant pressure. Most of the samples with reoriented hydrides are further subjected to a ductility test. Using finite element modeling, the test results are analyzed in terms of crack nucleation within radial hydrides at the outer diameter and crack growth through the thickness of the tubular samples. The combination of test results shows that samples with hydrogen contents of about 100 wppm had the lowest ductility.

  17. 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.

  18. Influence of surface treatment on the oxidation behavior of zirconium and zircaloy-4

    International Nuclear Information System (INIS)

    Costa, I.; Ramanathan, L.V.

    1986-01-01

    The influence of fluoride concentration in surface treatment solutions on the oxidation behavior of Zr and Zircaloy-4 in the temperature range 350-760 0 C have been studied by means of thermogravimetric analysis. Two solutions containing different concentrations of hydrofluoric acid have been used for surface treatments, following which surface roughness measurements were also carried out. The influence of fluoride ion concentration on oxidation behavior has been found to be significant at higher temperatures. (Author) [pt

  19. 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)

  20. 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.

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

    Science.gov (United States)

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

    2017-07-24

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

  2. NORA-2, a model for creep deformation and rupture of zircaloy at high temperatures

    International Nuclear Information System (INIS)

    Raff, S.; Meyder, R.

    1983-01-01

    A model has been developed to describe Zircaloy cladding behaviour under LOCA and small leak conditions within specified temperature range and strain rates. The deformation model consists of a strain rate equation with two components representing strain rate controlled contributions from different deformation mechanisms. Transition from one mechanism to the other produces the strain rate dependence of the stress exponent of steady state creep. During transient creep the change of creep mechanisms produces a flow softening behaviour which induces unstable creep. Together with a strain hardening model, the strain history can be described for low and high strain values. The influence of oxidation is taken into account by modelling hardening due to solid solution of oxygen, cracking of the brittle oxide and oxygen stabilised α-phase layers, and by an oxidation-induced creep component in steam atmosphere. The rupture criterion is based on a strain fraction rule whose variables are temperature, strain rate or applied stress, and oxygen content. (author)

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

    KAUST Repository

    Myzaferi, A.

    2016-08-11

    The bottom cladding design of semipolar III-nitride laser diodes is limited by stress relaxation via misfit dislocations that form via the glide of pre-existing threading dislocations (TDs), whereas the top cladding is limited by the growth time and temperature of the p-type layers. These design limitations have individually been addressed by using limited area epitaxy (LAE) to block TD glide in n-type AlGaN bottom cladding layers and by using transparent conducting oxide (TCO) top cladding layers to reduce the growth time and temperature of the p-type layers. In addition, a TCO-based top cladding should have significantly lower resistivity than a conventional p-type (Al)GaN top cladding. In this work, LAE and indium-tin-oxide cladding layers are used simultaneously in a (202⎯⎯1) III-nitride laser structure. Lasing was achieved at 446 nm with a threshold current density of 8.5 kA/cm2 and a threshold voltage of 8.4 V.

  4. Stress corrosion cracking of zircaloy. The use of laboratory data to predict in-reactor behaviour

    International Nuclear Information System (INIS)

    Miller, A.K.; Ocken, H.

    1981-01-01

    Pellet-cladding interaction (PCI) can lead to failure of the Zircaloy tubing used as cladding in water-cooled reactors. Many investigations have shown that the mechanism directly responsible for such fuel rod failures is stress corrosion cracking (SCC) of Zircaloy tubing. Laboratory studies have yielded extensive data on the time-to-failure (tsub(f)) behaviour of Zircaloy tubing specimens as a function of such important variables as the applied hoop stress (σ sub(h)), the iodine concentration (I 2 ), the temperature (T) and the fluence (F). These data have been used to predict the response of Zircaloy tubing exposed in-reactor. A typical approach is to fit laboratory data to obtain an empirical equation for tsub(f) in terms of the variables identified above. The question can then be posed as to whether it is appropriate to use such an empirical expression for predicting in-reactor behaviour. This paper describes the approach which has been taken in modelling the SCC process. It first reviews the experimental observations upon which the model is based. A summary of the key features of the model is then presented. The model's capabilities, emphasizing those predictions that are independent of data used to evaluate empirical constants, are briefly discussed. Finally, it is shown how the model can be used to predict important differences between the response of tubing specimens exposed in the laboratory and the response of large quantities of tubing exposed in-reactor

  5. 76 FR 4391 - Calvert Cliffs Nuclear Power Plant, LLC, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2...

    Science.gov (United States)

    2011-01-25

    ... rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall... uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency... fuel design consists of low enriched uranium oxide fuel within M5 zirconium alloy cladding. Since the...

  6. Probabilistic assessment of spent-fuel cladding breach

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  7. Probabilistic assessment of spent-fuel cladding breach

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. Mechanical test of E110 cladding material oxidized in hydrogen rich steam atmosphere

    International Nuclear Information System (INIS)

    Windberg, P.; Perez-Fero, E.

    2005-01-01

    The behavior of the fuel cladding under accidental conditions has been studied at the AEKI for more than a decade. Earlier, the effect of oxygen and hydrogen content on the mechanical properties was studied separately. The present experiments can help to understand what kind of processes took place in the cleaning tank at Paks NPP (2003). The purpose of our experiments was to investigate high temperature oxidation of E110 cladding in steam + hydrogen mixture. A high temperature tube furnace was used for oxidation of the samples. The oxidation was carried out at three different temperatures (900 0 C, 1000 0 C, 1100 0 C). The hydrogen content in the steam was varied between 19-36 vol%. The oxygen content of the sample was defined as oxidation ratio. Two sizes (length: 2 and 8 mm) of cladding rings and 100 mm long E110 cladding tubes were oxidized. After the oxidation we made compression and tensile tests for rings, and ballooning experiments for 100 mm long tube. The most important conclusions were the following. Oxidation in H-rich steam atmosphere need longer time to get the same oxidation ratio compared to the steam oxidation without hydrogen. The shorter oxidation time results in a more compact oxide layer. The longer oxidation time leads to a cracked oxide layer. (author)

  9. Deformation of PWR cladding following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.

    1979-07-01

    A review is presented of recent experiments to investigate the deformation behaviour of Zircaloy cladding in simulated loss-of-coolant accidents. The behaviour of Zircaloy cladding is shown to be controlled by a complex interaction of metallurgical and heat transfer variables, with the latter having a major influence. There is a significant increase in both diametral strain and the axial extent of deformation in multi-rod compared with single-rod tests. The extent to which this will occur in nuclear-heated tests is not yet known; however, it is expected that the 'smearing' of the gamma-radiation portion of decay heat in such tests will tend to reduce circumferential temperature variations. Opposing this is the influence of the colder control rods in an assembly. The resolution of this dichotomy will require a series of in-reactor multi-rod tests and attendant code development. (author)

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

    International Nuclear Information System (INIS)

    Shewfelt, R.S.W.

    1988-01-01

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

  11. Behavior and properties of Zircaloys in power reactors: A short review of pertinent aspects in LWR fuel

    International Nuclear Information System (INIS)

    Garzarolli, F.; Stehle, H.; Steinberg, E.

    1996-01-01

    Zircaloy-2 and -4, developed mainly in the US, have been used in Germany for fuel rod claddings and in-core structural components from the beginning of reactor technology. Extensive studies of the material properties of the Zircaloys have been performed in Siemens laboratories since 1957. Zircaloy-2 and -4 turned out to be very reliable materials that fulfilled all requirements for normal operation and likewise the requirements for postulated accidental conditions and for intermediate storage for many years. Optimization of Zircaloy-2 and -4 during recent years includes both optimization of microstructure and of chemical composition. BWRs and PWRs need differently optimized materials. Today's more demanding operation conditions and discharge burnups required a further optimization of the Zircaloys and for hot PWRs even the development of more corrosion-resistant Zr alloys. A significant improvement of PWR corrosion behavior can be achieved with Zr alloys using the alloying elements of Zircaloy with somewhat modified concentrations. Sn should be below or at least in the lower range of the ASTM specification range for Zircaloy-4, Fe and Cr should be somewhat higher, and Si should be specified as an alloying element rather than as an impurity

  12. Chemical interaction of fuel and cladding tubes

    International Nuclear Information System (INIS)

    Kirihara, Tomoo; Yamawaki, Michio; Obata, Naomi; Handa, Muneo.

    1983-01-01

    It was attempted to take up the behavior of nuclear fuel in cores and summarize it by the expert committee on the irradiation behavior of nuclear fuel from fiscal 1978 to fiscal 1980 from the following viewpoints. The behavior of nuclear fuel in cores has been treated separately according to each reactor type, accordingly this point is reconsidered. The clearly understood points and the uncertain points are discriminated. It is made more easily understandable for people in other fields of atomic energy. This report is that of the group on the chemical interaction, and the first report of this committee. The chemical interaction as the behavior of fuel in cores is in the unseparable relation to the mechanical interaction, but this relation is not included in this report. The chemical interaction of fuel and cladding tubes under irradiation shows different phenomena in LWRs and FBRs, and is called SCC and FCC, respectively. But this point of causing the difference must be understood to grasp the behavior of fuel. The mutual comparison of oxide fuels for FBRs and LWRs, the stress corrosion cracking of zircaloy tubes, and fuel-cladding chemical interaction in FBRs are reported. (Kako, I.)

  13. Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4

    International Nuclear Information System (INIS)

    Racine, A.

    2005-09-01

    In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

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

    International Nuclear Information System (INIS)

    Chung, H.M.; Daum, R.S.; Hiller, J.M.; Billone, M.C.

    2002-01-01

    Transmission electron microscopy (TEM) was used to examine Zircaloy fuel cladding, either discharged from several PWRs and a BWR after irradiation to fluence levels of 3.3 to 8.6 X 10 21 n cm -2 (E > 1 MeV) or hydrogen-charged and heat-treated under stress to produce radial hydrides; the goal was to determine the microstructural and crystallographic characteristics of hydride precipitation. Morphologies, distributions, and habit planes of various types of hydrides were determined by stereo-TEM. In addition to the normal macroscopic hydrides commonly observed by optical microscopy, small 'microscopic' hydrides are present in spent-fuel cladding in number densities at least a few orders of magnitude greater than that of macroscopic hydrides. The microscopic hydrides, observed to be stable at least up to 333 deg C, precipitate in association with -type dislocations. While the habit plane of macroscopic tangential hydrides in the spent-fuel cladding is essentially the same as that of unirradiated unstressed Zircaloys, i.e., the [107] Zr plane, the habit plane of tangential hydrides that precipitate under high tangential stress is the [104] Zr plane. The habit plane of radial hydrides that precipitate under tangential stress is the [011] Zr pyramidal plane, a naturally preferred plane for a cladding that has 30 basal-pole texture. Effects of texture on the habit plane and the threshold stress for hydride reorientation are also discussed. (authors)

  15. Analytical approaches and experimental verification to describe the influence of cold work and heat treatment on the mechanical properties of zircaloy cladding tubes

    International Nuclear Information System (INIS)

    Steinberg, E.; Schaa, A.; Weidinger, H.G.

    1984-01-01

    Well-controlled laboratory heat treatments were performed in the range from 460 to 610 0 C(733 to 883 K) and from 1 to 8 h at temperature on Zircaloy-4 cladding tubes with three different degrees of initial cold work (40%, 64%, and 76%). Within this range the influence of annealing temperature T and time t and of cold work on the yield strength R /SUB pO.2/ at 400 0 C(673 K) and on the degree R of recrystallization was experimentally determined. This data base was used to verify a semi-empirical approach to describe analytically the dependence of yield strength and recrystallization on the aforementioned technological parameters T and t for the annealing and /phi/ = ln l/l /SUB o/ as a measure for the applied cold work

  16. Plating on Zircaloy-2

    International Nuclear Information System (INIS)

    Dini, J.W.; Johnson, H.R.; Jones, A.

    1979-03-01

    Zircaloy-2 is a difficult alloy to coat with an adherent electroplate because it easily forms a tenacious oxide film in air and aqueous solutions. Procedures reported in the literature and those developed at SLL for surmounting this problem were investigated. The best results were obtained when specimens were first etched in either an ammonium bifluoride/sulfuric acid or an ammonium bifluoride solution, plated, and then heated at 700 0 C for 1 hour in a constrained condition. Machining threads in the Zircaloy-2 for the purpose of providing sites for mechanical interlocking of the plating also proved satisfactory

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-04-01

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

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

    International Nuclear Information System (INIS)

    Boltax, A.; Biancheria, A.

    1977-01-01

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

  19. 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

  20. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Lunde, L.; Olshausen, K.D.

    1980-01-01

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

  1. Fabrication of oxide dispersion strengthened ferritic clad fuel pins

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  2. Microstructure and Oxidation Behavior of CrAl Laser-Coated Zircaloy-4 Alloy

    Directory of Open Access Journals (Sweden)

    Jeong-Min Kim

    2017-02-01

    Full Text Available Laser coating of a CrAl layer on Zircaloy-4 alloy was carried out for the surface protection of the Zr substrate at high temperatures, and its microstructural and thermal stability were investigated. Significant mixing of CrAl coating metal with the Zr substrate occurred during the laser surface treatment, and a rapidly solidified microstructure was obtained. A considerable degree of diffusion of solute atoms and some intermetallic compounds were observed to occur when the coated specimen was heated at a high temperature. Oxidation appears to proceed more preferentially at Zr-rich region than Cr-rich region, and the incorporation of Zr into the CrAl coating layer deteriorates the oxidation resistance because of the formation of thermally unstable Zr oxides.

  3. The hydrogen generated as a gas and storage in Zircaloy during water quenching

    International Nuclear Information System (INIS)

    Garcia, Eduardo A.

    1999-01-01

    A simple one-dimensional diffusion model has been developed for the complex process of Zircaloy oxidation during water quenching, calculating the hydrogen liberated as a gas and the hydrogen stored in the metal. The model was developed on the basis of small-scale separate-effects quench experiments performed at Forschungszentrum Karlsruhe. The new oxide surface and the new metallic surface produced by cracking of the oxide during quenching are calculated for each experiment performed at 1200 , 1400 and 1600 C degrees using as-received Zircaloy-4 (no pre oxidation) and with Zircaloy specimens pre oxidised to give oxide thicknesses of 100μm and 300μm. The results are relevant to accident management in light water reactors. (author)

  4. The hydrogen generated as a gas and storage in Zircaloy during steam quenching

    International Nuclear Information System (INIS)

    Garcia, Eduardo A.

    2000-01-01

    A simple one-dimensional diffusion model has been developed for the complex process of Zircaloy oxidation during steam quenching, calculating the hydrogen liberated as a gas and the hydrogen stored in the metal. The model was developed on the basis of small-scale separate-effects quench experiments performed at Forschungszentrum Karlsruhe. The new oxide surface and the new metallic surface produced by cracking of the oxide during quenching are calculated for each experiment performed at 1200 centigrade, 1400 centigrade and 1600 centigrade using as-received Zircaloy-4 (no pre-oxidation) and with Zircaloy specimens pre-oxidized to give oxide thickness of 100μm and 300μm. The results are relevant to accident management in nuclear power plants. (author)

  5. Development of remote welding technology for nuclear fuel end capping (A study on the weldability of Zircaloy-4)

    Energy Technology Data Exchange (ETDEWEB)

    Kho, Jin Hyun; Sung, Ho Hyun; Hyun, Yong Kyu; Suh, Hee Kang [Korea University of Technology and Education, Cheonan (Korea)

    1998-03-01

    The integrity of nuclear fuel end cap welds is essential to the nuclear fuel performance and safety as well as the usability of power plant. The first aim of this project is to obtain experimental data on the nuclear fuel cladding materials of Zircaloy-4 with welding processes such as plasma arc, gas tungsten arc and laser beam welding. the data obtained in this study will be applicable to the nuclear fuel design, fabrication and nuclear fuel quality control. In addition, the welding processes applicable to the Zircaloy-4 welding were compared and contrasted. The weldability of Zircaloy-4 was evaluated from the metallurgical and mechanical standpoints. 88 refs., 57 figs., 16 tabs. (Author)

  6. Investigation of flow condition on the oxidation of Zircaloy-4 in air at 850 and 1100 .deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Maeng, Yun Hwan; Lee, Jae Young [Hangdong Global University, Pohang (Korea, Republic of); Park, Sang Gil [ACT Co. Ltd, Daejeon (Korea, Republic of)

    2016-05-15

    An oxidation behavior of the Zircaloy-4 was experimentally studied by varying a flow rate and partial pressure of air. Tests were conducted at two distinct temperatures in which a kinetic transition was occurred, or not: 850 .deg. C and 1100 .deg. C. The effects of flow rate and partial pressure of air was studied by a measurement of mass gain using thermogravimetric analyzer (TGA). After experiments, samples were observed with macrophotography and metallography using optical microscopy. The effect of flow rate and partial pressure of air were qualitatively analyzed with those methods. The effects of flow conditions on the oxidation kinetics of Zircaloy-4 samples were qualitatively studied. The flow rate and the partial pressure of air were changed and their effects was different when the temperature was changed.

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

    International Nuclear Information System (INIS)

    Pascard, R.

    1977-01-01

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

  8. Measurements of delayed hydride cracking propagation rate in the radial direction of Zircaloy-2 cladding tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan); Uchikoshi, H. [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The delayed hydride cracking (DHC) velocity of Zircaloy-2 was measured. Black-Right-Pointing-Pointer The velocity followed the Arrhenius law up to 270 Degree-Sign C. Activation energy was 49 kJ/mol. Black-Right-Pointing-Pointer The threshold stress intensity factor for the DHC was from 4 to 6 MPa m{sup 1/2}. Black-Right-Pointing-Pointer An increase in material strength accelerated the DHC. Black-Right-Pointing-Pointer Precipitation and fracture of hydrides at a crack tip is responsible for the DHC. - Abstract: Delayed hydride cracking (DHC) tests of Zircaloy-2 cladding tubes were performed in the chamber of a scanning electron microscope (SEM) to directly observe the crack propagation and measure the crack velocity in the radial direction of the tubes. Pre-cracks were produced at the outer surfaces of the tubes. Hydrogen contents of the tubes were from 90 ppm to 130 ppm and test temperatures were from 225 Degree-Sign C to 300 Degree-Sign C. The crack velocity followed the Arrhenius law at temperatures lower than about 270 Degree-Sign C with apparent activation energy of about 49 kJ/mol. The upper temperature limit for DHC, above which DHC did not occur, was about 280 Degree-Sign C. The threshold stress intensity factor for the initiation of the crack propagation, K{sub IH}, was from about 4 MPa m{sup 1/2} to 6 MPa m{sup 1/2}, almost independent of temperature. An increase in 0.2% offset yield stress of the material accelerated the crack velocity and slightly decreased K{sub IH}. Detailed observations of crack tip movement showed that cracks propagated in an intermittent fashion and the propagation gradually approached the steady state as the crack depth increased. The SEM observations also showed that hydrides were formed at a crack tip and a number of micro-cracks were found in the hydrides. It was presumed from these observations that the repetition of precipitation and fracture of hydrides at the crack tip would be

  9. 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

  10. 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

  11. Effect of the oxidation front penetration on in-clad hydrogen migration

    Science.gov (United States)

    Feria, F.; Herranz, L. E.

    2018-03-01

    In LWR fuel claddings the embrittlement due to hydrogen precipitates (i.e., hydrides) is a degrading mechanism that concerns in nuclear safety, particularly in dry storage. A relevant factor is the radial distribution of the hydrogen absorbed, especially the hydride rim formed. Thus, a reliable assessment of fuel performance should account for hydrogen migration. Based on the current state of modelling of hydrogen dynamics in the cladding, a 1D radial model has been derived and coupled with the FRAPCON code. The model includes the effect of the oxidation front progression on in-clad hydrogen migration, based on experimental observations found (i.e., dissolution/diffusion/re-precipitation of the hydrogen in the matrix ahead of the oxidation front). A remarkable quantitative impact of this new contribution has been shown by analyzing the hydrogen profile across the cladding of several high burnup fuel scenarios (>60 GW d/tU); other potential contributions like thermodiffusion and diffusion in the hydride phase hardly make any difference. Comparisons against PIE measurements allow concluding that the model accuracy notably increases when the effect of the oxidation front is accounted for in the hydride rim formation. In spite of the promising results, further validation would be needed.

  12. Aerosol material release rates from zircaloy-4 at temperatures from 2000 to 22000C

    International Nuclear Information System (INIS)

    Mulpuru, S.R.; Wren, D.J.; Rondeau, R.K.

    1987-01-01

    During some postulated severe accidents involving loss of coolant and loss of emergency coolant injection, the temperatures in a CANDU reactor fuel channel become high enough to cause failure and melting of the Zircaloy fuel cladding. At such high temperatures, vapors of fission products and structural (fuel and cladding) materials will be released into the coolant steam and hydrogen mixture. These vapors will condense as cooler conditions are encountered downstream. The vapors from structural materials are relatively involatile; therefore, they will condense readily into aerosol particles. These particles, in turn, will provide sites for the condensation of the more volatile fission products. The aerosol transport of fission products in the primary heat transport system (PHTS) will thus be influenced by the structural material release rates. As part of an ongoing program to develop predictive tools for aerosol and associated fission product transport through the PHTS, experiments have been conducted to measure the vapor mass release rates of the alloying elements from Zircaloy-4 at high temperatures. The paper presents the results and analysis of these experiments

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  14. 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)

  15. 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.

  16. UK experience on fuel and cladding interaction in oxide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom); Findlay, J R [AERE, Harwell, Didcot, Oxon (United Kingdom)

    1977-04-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.

  17. UK experience on fuel and cladding interaction in oxide fuels

    International Nuclear Information System (INIS)

    Batey, W.; Findlay, J.R.

    1977-01-01

    The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor (DFR). Supporting information has been obtained from irradiation programmes in Materials Testing Reactors (MTR). Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed

  18. High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Suzuki, Motoye

    1982-03-01

    In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

  19. Behavior and failure of fresh, hydrided and irradiated Zircaloy-4 fuel claddings under RIA conditions

    International Nuclear Information System (INIS)

    Le Saux, M.

    2008-01-01

    The purpose of this study is to characterize and simulate the mechanical behaviour and failure of fresh, hydrided and irradiated (in pressurized water reactors) cold-worked stress relieved Zircaloy-4 fuel claddings under reactivity initiated accident conditions. A model is proposed to describe the anisotropic viscoplastic mechanical behavior of the material as a function of temperature (from 20 C up to 1100 C), strain rate (from 3.10 -4 s -1 up to 5 s -1 ), fluence (from 0 up to 1026 n.m -2 ) and irradiation conditions. Axial tensile, hoop tensile, expansion due to compression and hoop plane strain tensile tests are performed at 25 C, 350 C and 480 C in order to analyse the anisotropic plastic and failure properties of the non-irradiated material hydrided up to 1200 ppm. Material strength and strain hardening depend on temperature and hydrogen in solid solution and precipitated hydride contents. Plastic anisotropy is not significantly modified by hydrogen. The material is embrittled by hydrides at room temperature. The plastic strain that leads to hydride cracking decreases with increasing hydrogen content. The material ductility, which increases with increasing temperature, is not deteriorated by hydrogen at 350 C and 480 C. Macroscopic fracture modes and damage mechanisms depend on specimen geometry, temperature and hydrogen content. A Gurson type model is finally proposed to describe both the anisotropic viscoplastic behavior and the ductile fracture of the material as a function of temperature and hydrogen content. (author) [fr

  20. Characterization of Zircaloy-4 oxide layers by impedance spectroscopy

    International Nuclear Information System (INIS)

    Barberis, P.

    1999-01-01

    Two Zircaloy-4 type alloys with different tin contents (0.5 and 1.2 wt%) have been oxidized in autoclave (400 C in steam) for several durations (1-140 days). The film has been characterized by electrochemical impedance spectroscopy (EIS). Several soaking times have been investigated (up to 40 days). The Cole-Cole representation has been used to display and study the data. A simple electrical model has been derived from the observed spectra: the electrical circuit includes two RC loops in series, whose capacitances are frequency dispersed. It is thoroughly related to the layer structure. It has been shown that even before the kinetic transition, the film is constituted of three parts: an inner layer which is compact, an outer layer subdivided in an external region immediately soaked by the electrolyte, and an internal one in which electrolyte diffusion processes can take place. The kinetic transition is interpreted in terms of an abrupt 'compacity' change, both layers degrading at this point. The alloy with high tin content exhibits higher dispersive properties of the oxide layer formed on it, in correlation with its faster oxidation kinetics. (orig.)

  1. 75 FR 12312 - South Carolina Electric and Gas Company; Virgil C. Summer Nuclear Station, Unit 1; Exemption

    Science.gov (United States)

    2010-03-15

    ... cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation. The... of energy release, hydrogen concentration, and cladding oxidation from the metal-water reaction shall... pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or...

  2. Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4; Influence de l'orientation des hydrures sur les modes de deformation, d'endommagement et de rupture du zircaloy-4 hydrure

    Energy Technology Data Exchange (ETDEWEB)

    Racine, A

    2005-09-15

    In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

  3. The effect of small specimen volume on the deformation of Zircaloy-4 PWR cladding under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Oakden, M.M.; Reynolds, A.E.

    1983-01-01

    Creep rupture tests were performed in flowing steam on single rod specimens of 17x17 type PWR Zircaloy-4 cladding 460 mm long. They were tested at temperatures between 640 deg.C and 985 deg.C with internal pressures in the range 1.00-9.65 MPa (gauge) (145-1400 lb/in 2 ). The internal free volume was limited to 2.9 ml. Axially extended 'carrot' shaped deformations were produced, the range of temperatures and pressures over which these occurred was found to be similar to that observed in previous tests conducted with a much larger free volume. The main result of limiting the internal volume was that straining of the specimens was accompanied by a more rapid drop in the internal pressure than occurred previously, and which reduced the extent of the deformation compared with that seen in the earlier work. However, diametral strains in excess of 33% were observed which would result in mechanical interaction of neighbouring bulges if this occurred in a multi-rod array. (author)

  4. Chemical decontamination and melt densification

    International Nuclear Information System (INIS)

    Dillon, R.L.; Griggs, B.; Kemper, R.S.; Nelson, R.G.

    1976-01-01

    Preliminary studies on the chemical decontamination and densification of Zircaloy, stainless steel, and Inconel undissolved residues remaining after dissolution of the UO 2 --PuO 2 spent fuel material from sheared fuel bundles are reported. The studies were made on cold or very small samples to demonstrate the feasibility of the processes developed before proceeding to hot cell demonstrations with kg level of the sources. A promising aqueous decontamination method for Zr alloy cladding was developed in which oxidized surfaces are conditioned with HF prior to leaching with ammonium oxalate, ammonium citrate, ammonium fluoride, and hydrogen peroxide. Feasibility of molten salt decontamination of oxidized Zircaloy was demonstrated. A low melting alloy of Zircaloy, stainless steel, and Inconel was obtained in induction heated graphite crucibles. Segregated Zircaloy cladding sections were directly melted by the inductoslag process to yield a metal ingot suitable for storage. Both Zircaloy and Zircaloy--stainless steel--Inconel alloys proved to be highly satisfactory getters and sinks for recovered tritium

  5. Oxidation of Zircaloy-4 in steam-nitrogen mixtures at 600-1200 °C

    Science.gov (United States)

    Steinbrueck, Martin; da Silva, Fabio Oliveira; Grosse, Mirco

    2017-07-01

    High-temperature oxidation of zirconium alloys in steam-nitrogen atmospheres may be relevant during various nuclear accident scenarios. Therefore, isothermal oxidation tests with Zircaloy-4 in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200 °C using thermogravimetry. The gas compositions were varied between 0 and 100 vol% nitrogen including 0.1 and 90 vol%. The strong effect of nitrogen on the oxidation kinetics of zirconium alloys was confirmed in these tests in mixed steam-nitrogen atmospheres. Even very low concentrations of nitrogen (starting from less than 1 vol%) strongly increase reaction kinetics. Nitrogen reduces transition time from protective to non-protective oxide scale (breakaway). The formation of zirconium nitride, ZrN, and its re-oxidation is the main reason for the highly porous oxide scales after transition. The results of this study have shown the safety relevant role of nitrogen during severe accidents and, more generally, suggest the need of using well controlled gas atmospheres for experiments on oxidation of zirconium alloys.

  6. Superficial characterization and zircaloy-2 electrochemistry with hydrothermal deposit of platinum; Caracterizacion superficial y electroquimica de zircaloy-2 con deposito hidrotermal de platino

    Energy Technology Data Exchange (ETDEWEB)

    Contreras R, A.; Arganis J, C. R.; Medina A, A. L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Gris C, M. M., E-mail: aida.contreras@inin.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2011-11-15

    The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy-2 tubes that contain in their interior UO{sub 2} pellets. With the objective of mitigating the speed of crack growth by IGSCC to a minimum negative impact on the BWR operation, General Electric developed the noble metals chemical addition (NMCA), in where noble metals particles as Pt, Pd, and Rh, are deposited on the surface of the metal to catalyze the recombination of H{sub 2} and O{sub 2}. Hydrogen is also injected to have it in excess and to favor this recombination (HWC) and zinc to reduce dose. In this work was oxidized zircaloy-2 low similar conditions to the HWC, platinum was deposited starting from a solution of Na{sub 2}Pt(OH){sub 6} with 30 ppm of Pt, in refined samples and without polishing, they were characterized by scanning electron microscopy, energy dispersed spectroscopy, XPS and electrochemistry, by means of Tafel curves and cyclical polarization. On the zircaloy surface was found a ZrO{sub 2} layer that remains under the different study conditions. Under HWC conditions is the oxides formation, possibly complex oxides of zirconium, iron and tin. After the platinum deposit these oxides decrease forming the sub-oxides: Zr{sub 2}O, Zr O, Zr{sub 2}O{sub 3}. The Tafel curves indicates the reduction of the oxygen of the sample with platinum and the cyclical polarization curves show that the reactions that happen on the zircaloy electrodes are not dur to located corrosion. (Author)

  7. A pellet-clad interaction failure criterion

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-23

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

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

    International Nuclear Information System (INIS)

    Esteves, A.M.

    1991-01-01

    The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. (author)

  11. Dependence of laser assisted cleaning of clad surfaces on the laser fluence

    International Nuclear Information System (INIS)

    Nilaya, J.P.; Raote, P.; Sai Prasad, M.B.; Biswas, D.J.; Aniruddha Kumar

    2005-01-01

    The decontamination factor is studied as a function of laser fluence for three kinds of clad surfaces viz., plain zircaloy, autoclaved zircaloy and SS with cesium as the test contamination. It has been found that the decontamination factor exhibits a maximal behaviour with the laser fluence and its maximum value occurs at different laser fluences in the three cases. The maximal behaviour is attributed to reduced coupling of energy from the laser beam to the substrate due to the initiation of surface-assisted optical breakdown. The results obtained in the experiment carried out in helium environment qualitatively support this explanation (author)

  12. Influence of γ-irradiation on the transport kinetics of hydrogen in pre-transition oxidized Zircaloy-4 at room temperature

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Frantz A., E-mail: frantz.martin@cea.fr [CEA, DEN, DPC, SCCME, Laboratoire d' Etude de la Corrosion Aqueuse, F-91191 Gif-sur-Yvette (France); Dauvois, Vincent, E-mail: vincent.dauvois@cea.fr [CEA, DEN, DPC, SECR, Laboratoire de Radiolyse et de la Matière Organique, F-91191 Gif-sur-Yvette (France); Esnouf, Stéphane, E-mail: stephane.esnouf@cea.fr [CEA, DEN, DPC, SECR, Laboratoire de Radiolyse et de la Matière Organique, F-91191 Gif-sur-Yvette (France); CEA, DSM, IRAMIS, LIDYL, PCR, F-91191 Gif-sur-Yvette (France); Fourdrin, Chloé, E-mail: chloe.fourdrin@culture.gouv.fr [CEA, DEN, DPC, SECR, Laboratoire de Radiolyse et de la Matière Organique, F-91191 Gif-sur-Yvette (France); Jomard, François, E-mail: francois.jomard@uvsq.fr [CNRS/UVSQ, UMR 8635, GEMAC, 45 avenue des Etats Unis – Bâtiment Fermat, F-78035 Versailles (France); Chêne, Jacques, E-mail: chene_jacques@orange.fr [CNRS/CEA, UMR 8587, LECA – CEA, Saclay, F-91191 Gif-sur-Yvette (France)

    2015-10-15

    In a context of nuclear fuel reprocessing, the free-of-fuel hulls and ends of cladding tubes are compacted. The possibility of hydrogen degasing or absorption from/into these tubes has been studied with and without gamma irradiation at 293 K by means of deuterium as isotopic tracer for hydrogen. Under irradiation, as without, the hydrides present in the Zircaloy-4 hulls seem stable. The oxide layer present at the surface of the hulls allows a slow diffusion of deuterium, and the irradiation appeared to have no specific effect on the diffusion process: in both cases, hydrogen diffusion coefficients of the order of 5·10{sup −18} cm{sup 2} s{sup −1} have been found. The subsurface activity of deuterium is increased by one order of magnitude at least under irradiation, probably due to an activation of the dissociation/absorption kinetic steps. - Highlights: • Hydrogen diffusion coefficients at RT in zirconia grain boundaries were determined. • γ irradiation increases the hydrogen subsurface activity when exposed to H{sub 2} gas. • The dense thin zirconia film is not altered much by HNO{sub 3} exposure at 90 °C for 24 h. • Hydrogen desorption from Zr hydrides was studied at room temperature under γ rays.

  13. 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)

  14. Modelling of zircaloy-4 degradation in oxygen and nitrogen mixtures at high temperature

    International Nuclear Information System (INIS)

    Lasserre-Gagnaire, Marina

    2013-01-01

    Zircaloy-4 claddings provide the first containment of UO 2 fuel in Pressurised Water Reactors. It has been demonstrated that the fuel assemblies cladding could be exposed to air at high temperature in several accidental situations such as a loss of cooling accident in a spent fuel storage When mixed to oxygen at high temperature, the nitrogen, usually used as an inert gas, causes the accelerated corrosion of the cladding. The kinetic curves obtained by thermogravimetry reveal two stages: a pre-transition and a post-transition one. The pre-transition stage corresponds to the growth of a protective dense oxide layer: the kinetic rate decreases with time and is controlled by oxygen vacancy diffusion in the oxide layer. In the post-transition stage, the oxide layer is no longer protective and the kinetic rate increases with time. Images obtained by optical microscopy of a sample in the post-transition stage reveal the presence of corroded zones characterized by a porous scale with zirconium nitride precipitates at metal - oxide interface. Corrosion of Zy4 plates at 850 deg. C under mixed oxygen - nitrogen atmospheres has been studied during the post-transition stage. A sequence of three reactions is proposed to explain the mechanism of nitrogen-enhanced corrosion and the porosity of the corroded regions. The accelerating effect of nitrogen in the corrosion scale can therefore be described on the basis of an autocatalytic effect of the zirconium nitride precipitates. Then, it is demonstrated that the steady-state approximation as well as the existence of an elementary step controlling the growth process are valid during the post-transition stage. Thanks to the study of the variations of the surface rate of growth with the oxygen and nitrogen partial pressure, the rate-determining step is identified as the external interface reaction step of the oxidation of the zirconium nitride precipitates. Finally, a nucleation and growth model used for thermal reactions in powders

  15. Eddy-current testing of nuclear fuel cladding tubes using tilted encircling coil system, 1

    International Nuclear Information System (INIS)

    Yin, Renzhong; Sekine, Kazuyoshi; Shimizu, Hisaji; Tsukui, Kazushige; Urata, Megumu.

    1989-01-01

    The eddy current testing method with external encircling-coils has been widely used as a standard technique for inspection of defects in irradiated zircaloy cladding tubes. In this inspection, the systematic procedure to reliably characterize defects is required. This paper describes the newly developed external tilted encircling-coil system, in which the coil axis is inclined by an angle α to the sample tube axis, for reliable determination of the sort, location and size of defects. As the results of experimental work concerning some kinds of artificial defects in zircaloy cladding tubes using newly designed tilted coil system, an adaptable general-procedure for characterization of defects has been proposed. Furthermore, it has been confirmed that in the case of smaller tilt angles of coil, the signal-to noise ratio for defect response in this coil system is approximately equal to that of ordinary encircling coil system. (author)

  16. Superficial characterization by XP S of silver nanoparticles and their hydrothermal deposit over zircaloy; Caracterizacion superficial por XPS de nanoparticulas de plata y su deposito hidrotermal sobre zircaloy

    Energy Technology Data Exchange (ETDEWEB)

    Contreras R, A.; Gutierrez W, C.; Martinez M, I.; Medina A, A. L., E-mail: aida.contreras@inin.gob.mx [ININ, Departamento de Tecnologia de Materiales, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The analysis technique of X-ray photoelectron spectroscopy (XP S) is sensitive exclusively to the first layers of the solids surface, which allows obtaining information about the chemical, physical and electronic properties of them. The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy pipes that contain in their interior pellets or uranium dioxide. In this work is studied the zircaloy surface, oxidized zircaloy under similar conditions to those of a reactor BWR type and oxidized zircaloy with a hydrothermal deposit of silver nanoparticles and zinc. The silver deposit is a proposal of the Materials Technology Department of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, which has the same objective that the noble metals deposit (Pt, Pd, and Rh) that is practiced in some of the reactors BWR, in order to mitigating the speed of crack growth for IGSCC in stainless steels 304 Ss. (Author)

  17. Cladding hull decontamination and densification process. Part 2. Densification by inductoslag melting

    International Nuclear Information System (INIS)

    Nelson, R.G.; Montgomery, D.R.

    1980-04-01

    The Inductoslag melting process was developed to densify Zircaloy-4 cladding hulls. It is a cold crucible process that uses induction heating, a segmented water-cooled copper crucible, and a calcium fluoride flux. Metal and flux are fed into the furnace through the crucible, located at the top of the furnace, and the finished ingot is withdrawn from the bottom of the furnace. Melting rates of 40 to 50 kg/h are achieved, using 100 to 110 kW at an average energy use of 2.5 kWh/kg. The quality of ingots produced from factory supplied cladding tubing is sufficient to satisfy nuclear grade standards. An ingot of Zircaloy-4, made from melted cladding tubing that had been autoclaved to near reactor exposure and then descaled by the hydrogen fluoride decontamination process prior to Inductoslag melting, did not meet nuclear grade standards because the hydrogen, nitrogen, and hardness levels were too high. Melting development work is described that could possibly be used to test the capability of the Inductoslag process to satisfactorily melt a variety and mix of materials from LWR reprocessing, decontamination, and storage options. Results of experiments are also presented that could be used to improve remote operation of the melting process

  18. Effect of the anodization variables in the corrosion resistence of the zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Figueiredo, M.E.

    1981-02-01

    The anodization effect in the oxidation of the zircaloy-4 in steam atmosphere at 10,06MPa was investigated. It was also studied how the voltage and the types of electrolytes at several values of pH affect the growing of the anodic oxide film and the performance of the zircaloy-4 in relation to corrosion. Anodizations of zircaloy-4 tubes have been made with voltages ranging from zero to 280V and using electrolytic solutions of Na 2 B 4 O 7 , CH 3 COOH and NaOH in the concentrations of 1,0N, 0,1N and 0,01N. After anodization, the tubes were oxidized in autoclave under steam at 400 0 C and 10,06 MPa during 3 and 14 days. The results show that the anodization inhibit the oxidation process of zircaloy-4, and that this protection increases with the voltage applied for film formation. The relationship between the weight gain after oxidation in autoclave and the anodization voltage is of the exponential type: (σM/A) sub(AC) = Ce sup(-DV). The observed relationship between the applied voltage and the weight gain due to anodization is of the linear type: (σM/A) sub(AN) = aV. Concerning the influence of different electrolytes, it was observed a similar behaviour between them with respect to the thickness of the anodic oxide and the weight gain of zircaloy-4 after the autoclave test. (Author) [pt

  19. A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tahk, Y. W.; Oh, J. Y.; Lee, B. H.; Seo, C. G.; Chae, H. T.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    U{sub 3}Si{sub 2}-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

  20. SSMS near surface analysis of B in irradiated Zircaloy-2: ion implantation standards as a calibration technique

    International Nuclear Information System (INIS)

    Christie, W.H.; Carter, J.A.; Eby, R.E.; Landau, L.; Musick, W.R.

    1980-01-01

    Purpose of this study was to determine the amount of 10 B contamination on the surface of Zircaloy-2 clad irradiated fuel elements that had been stored in an aqueous solution containing 5000 wt. ppM enriched B. SMSS indicated that the contamination was less than 0.06 μg/cm 2

  1. Oxidation of Zircaloy-4 in steam-nitrogen mixtures at 600–1200 °C

    Energy Technology Data Exchange (ETDEWEB)

    Steinbrueck, Martin, E-mail: martin.steinbrueck@kit.edu; Oliveira da Silva, Fabio; Grosse, Mirco

    2017-07-15

    High-temperature oxidation of zirconium alloys in steam-nitrogen atmospheres may be relevant during various nuclear accident scenarios. Therefore, isothermal oxidation tests with Zircaloy-4 in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200 °C using thermogravimetry. The gas compositions were varied between 0 and 100 vol% nitrogen including 0.1 and 90 vol%. The strong effect of nitrogen on the oxidation kinetics of zirconium alloys was confirmed in these tests in mixed steam-nitrogen atmospheres. Even very low concentrations of nitrogen (starting from less than 1 vol%) strongly increase reaction kinetics. Nitrogen reduces transition time from protective to non-protective oxide scale (breakaway). The formation of zirconium nitride, ZrN, and its re-oxidation is the main reason for the highly porous oxide scales after transition. The results of this study have shown the safety relevant role of nitrogen during severe accidents and, more generally, suggest the need of using well controlled gas atmospheres for experiments on oxidation of zirconium alloys.

  2. Chemical interaction between the oxide and the clad in PHENIX fuel at burnup up to 60,000 MWd/t

    International Nuclear Information System (INIS)

    Conte, M.; Marcon, J.P.

    1977-01-01

    In every fuel element there is a potential problem of chemical interaction between the fissile portion and the clad. As a matter of fact, even if the choice of materials is made after having established a satisfactory chemical compatibility between the fuel- (UO 2 (U,Pu)O 2 , (U,Pu) C, . . .) and the clad (stainless steel, zircaloy, . . . ) out of pile, it is difficult to guarantee this compatibility after operation in the reactor due, on one hand, to the presence of fission products and, on the other hand, to impurities which are always present in the fuel to a greater or lesser degree. The fuel element currently chosen for the sodium-cooled fast reactors ((U,Pu)O 2 in stainless steel clad) does not avoid this problem, in particular because of the relatively high temperatures envisioned for this type of reactor - the clad temperature is about 650 deg. C. Since it is considered as a demonstration reactor, Phenix should be able to provide additional information on this phenomenon, and one will see that we have been able to shed light on some points which the experiments or irradiations made to date have been unable to explain. However, before presenting the experimental results obtained with Phenix fuel end drawing conclusions, we shall give a brief resume of the expected behavior of this fuel with respect to the phenomenon of interest. (author)

  3. Conditioning and storage of spent fuel cladding hulls by rolling and embedding in concrete

    International Nuclear Information System (INIS)

    Spenk, G.; Frotscher, H.; Graebner, H.; Kapulla, H.

    1981-01-01

    Under a contract with the European Atomic Energy Community the Kernforschungszentrum Karlsruhe, KfK, developed a conditioning process for LWR cladding waste. After compaction of the hulls by rolling they are embedded in a concrete matrix. In addition to basic data of the cladding waste, the compaction process, consisting of a dosage system and a rolling mill, is described. Several embedding techniques are possible, but a final selection has still to be made. Best results will probably be achieved by a vacuum technique. To characterize the waste product, leach tests have been started. The compression strength of compacted hulls embedded in concrete was determined to 2300 N.cm -2 . Hydrogen release due to radiolyses lies around 3 μl.g -1 sub(concrete).Mrad -1 which corresponds to the values expected on account of the water content of the samples. Less hydrogen was determined in samples with Zircaloy added. The tritium release of tritiated Zircaloy hulls embedded in concrete is greatly dependent on temperature and irradiation. At 100 0 C and with γ-irradiation the tritium release is about two orders of magnitude higher compared with experiments without irradiation. The thermal conductivity of samples of Zircaloy hulls embedded in concrete was determined to be 1.4W.m -1 .K -1 . (author)

  4. Effect of Thickness on Oxidation Behavior of Cr coated Zircaloy-4 using Arc Ion Plating

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Jung; Kim, Sun Jin [Hanyang University, Seoul (Korea, Republic of); Park, Jung Hwan; Kim, Hyun Gil; Jung, Yang Il; Park, Dong Jun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Ever since the Fukushima accident, accident tolerant fuel (ATF) has been widely studied. To increase the life time and safety of nuclear claddings, there are increasing demands for protective coatings exhibiting excellent oxidation resistance. Many metal and oxide films are produced by using this method because of the high kinetic energy of the ions, ionization efficiency and deposition rate. Candidate materials for a protective layer have higher thermal neutron absorption cross sections than Zr. However, there is no systematic study of thickness effect on oxidation resistance of protective layer. In this study, Cr films with different thickness (from 1 μm to 50 μm) were deposited on the cladding surfaces by AIP. The high temperature oxidation resistance of Cr films with different thicknesses has been investigated. Uniform oxide layer with nanoporous structures have been fabricated on the surface of Zr-Nb-Sn alloy. Oxidation behavior of the pristine Zr-Nb-Sn alloy and the Zr-Nb-Sn alloy with nanostructured oxide layer evaluated by measuring weight gain (TGA).

  5. Glow discharge mass spectrometry study of chemical impurities diffusion in zirconium oxide layers

    International Nuclear Information System (INIS)

    Actis-Dato, L.O.

    2000-02-01

    In the Pressurised Water Reactors (PWR) the primary cooling system is water at 350 deg C and 16 Mpa. In these extreme conditions the corrosion of the out-of-pile components of the reactor and in particular of the Zircaloy cladding containing the nuclear fuel pellets is accelerated. The formation of a growing oxide layer is observed on the elements of the nuclear reactor core. When the thickness of the oxide layer reaches a critical value, problems like structural malfunction and material failure can occur. At this stage the danger of the release of radioactivity in the coolant becomes effective. In this work a glow discharge mass spectrometer is used to study the diffusion of impurities like lithium, zinc and boron in oxide layers growing on Zircaloy samples. After a brief description of the different technique used the preparation and analysis of Zircaloy samples on which oxide layers of different thickness have been produced will be described. The analysis of these samples will allow the determination and evaluation the depth profiles of the impurities absorbed by the oxide. The analysis of the shape of the crater produced during the mass spectrometric analysis will give additional informations on the quality of the results obtained. The technique developed will finally be applied to the study of samples produced during reactor operation under real irradiation conditions. The results obtained show a lithium intake in the oxide layer and confirm the beneficial effect on the corrosion produced by boron. The influence of zinc on the corrosion behaviour of Zircaloy has not been established. The technique developed has also shown interesting capabilities concerning the analysis of irradiated samples. (author)

  6. Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shi, Xingwei; Cao, Xinrong; Liu, Zhengzhi

    2013-01-01

    Highlights: • A new verified oxidation model of cladding has been added in Severe Accident Program (SAP). • A coupled analysis method utilizing RELAP5 and SAP codes has been developed and applied to analyze a SA caused by LBLOCA. • Analysis of cladding oxidation under a SA for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) has been performed by SAP. • Estimation of the production of hydrogen has been achieved by coupled codes. - Abstract: Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding

  7. Conversion of fuel hulls to zirconate ion exchangers for stabilization of wastes from the thorium fuel cycle

    International Nuclear Information System (INIS)

    Levine, H.S.

    1978-01-01

    A conceptual reprocessing and waste management scheme for Zircaloy clad ThO 2 fuel was formulated to eliminate problems associated with concurrent dissolution of fuel and cladding in the conventional chop-leach headend step. These problems are avoided by use of a modified headend step to form oxide fuel and cladding process streams. A chlorinating agent then converts the cladding hulls and adhering fuel into volatile and nonvolatile chloride fractions. The former product is processed, by use of the Zircaloy conversion process, to form an inorganic ion exchange material and combined with HLLW from subsequent fuel reprocessing to form a stable and refractory waste form. The nonvolatile chloride fraction would be recovered, processed to remove chloride ions, and recombined with the main oxide fuel process stream for further treatment by use of the Thorex process

  8. Corrosion Characteristics and Kinetics of Zircaloys and Aluminium Alloys

    International Nuclear Information System (INIS)

    Sugondo; Chaidir, A

    1998-01-01

    Corrosion rate characterization of cladding materials has been done by dynamic method. The materials are zircaloy-2,zircaloy-4,AIMg2,and AIMgSi.The zircaloy alloys are characterized in the electrolytes of boric ion,iodide ion,lithium ion and cesium ion with a pH variation.The aluminum alloys are characterized in the cooling water of RSG-GAS reactor in different temperatures and Ph values .The results, show that corrosion product of iodine on zircaloy is not passivated, meanwhile the corrosion product of cesium undergoes passivation. However, the deposited substance in the surface of the specimens as indicated using WDX-SEM shows the same deposition rate.it is concluded therefore that iodine is diffused into the materials without getting resistance from the deposited substances on the surface. The effect of pH to corrosion rate of iodine on the zircaloy fluctuates meanwhile the cesium has the minimum corrosion rate at pH 7.5 At the concentration of 0.1 gram/1,cesium ion is more reactive than iodine but at higher concentration the reactivity becomes competitive . Furthermore , the interaction between zircaloy and boric ion at concentration of 300 ppm and lithium ion at 10 ppm shows an outstanding corrosion rate, i.e. 0.1 mpy. if both substances are mixed then the corrosion rate decreases drastically in the order of 10 -2 mpy.The reason of such a decrease may be due to the formation of complexes of boron lithium on the electrode surface. The arrhenius activation energies for such reaction have been found to be 37629.322 joule/mole 0 K for Al Mg 2 and 41609.822 joule /mole 0 K for AIMgSi ,respectively. This underlies the argument that AI Mg 2 is more reactive than AI Mg Si besides , AI Mg 2 is more reactive under acid condition meanwhile AI Mg Si more reactive under basic condition. Both alloys over come the minimum corrosion rate at the pH in between 4.7 to 7.5 and the level of the corrosion rate in the pH interval was outstanding

  9. Analysis of atomic distribution in as-fabricated Zircaloy-2 claddings by atom probe tomography under high-energy pulsed laser

    Energy Technology Data Exchange (ETDEWEB)

    Sawabe, T., E-mail: sawabe@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Sonoda, T.; Kitajima, S. [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Kameyama, T. [Tokai University, Department of Nuclear Engineering, Kitakaname 4-1-1, Hiratsuka, Kanagawa 259-1292 (Japan)

    2013-11-15

    The properties of second-phase particles (SPPs) in Zircaloy-2 claddings are key factors influencing the corrosion resistance of the alloy. The chemical compositions of Zr (Fe, Cr){sub 2} and Zr{sub 2}(Fe, Ni) SPPs were investigated by means of pulsed laser atom probe tomography. In order to prevent specimen fracture and to analyse wide regions of the specimen, the pulsed laser energy was increased to 2.0 nJ. This gave a high yield of average of 3 × 10{sup 7} ions per specimen. The Zr (Fe, Cr){sub 2} SPPs contained small amounts of Ni and Si atoms, while in Zr{sub 2}(Fe, Ni) SPPs almost all the Si was concentrated and the ratio of Zr: (Fe + Ni + Si) was 2:1. Atomic concentrations of the Zr-matrix and the SPPs were identified by two approaches: the first by using all the visible peaks of the mass spectrum and the second using the representative peaks with the natural abundance of the corresponding atoms. It was found that the change in the concentration between the Zr-matrix and the SPPs can be estimated more accurately by the second method, although Sn concentration in the Zr{sub 2}(Fe, Ni) SPPs is slightly overestimated.

  10. 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

  11. 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.

  12. Oxidation Kinetics of Ferritic Alloys in High-Temperature Steam Environments

    Science.gov (United States)

    Parker, Stephen S.; White, Josh; Hosemann, Peter; Nelson, Andrew

    2018-02-01

    High-temperature isothermal steam oxidation kinetic parameters of several ferritic alloys were determined by thermogravimetric analysis. The oxidation kinetic constant ( k) was measured as a function of temperature from 900°C to 1200°C. The results show a marked increase in oxidation resistance compared to reference Zircaloy-2, with kinetic constants 3-5 orders of magnitude lower across the experimental temperature range. The results of this investigation supplement previous findings on the properties of ferritic alloys for use as candidate cladding materials and extend kinetic parameter measurements to high-temperature steam environments suitable for assessing accident tolerance for light water reactor applications.

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

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1979-08-01

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

  14. Surface analytical investigations of the thermal behaviour of passivated Zircaloy-4 surfaces and of the reaction behaviour of iodine with Zircaloy-4 surfaces

    International Nuclear Information System (INIS)

    Kaufmann, R.

    1988-07-01

    In the first part of the present work the thermal behaviour of atmospherically oxidized Zircaloy-4 samples was investigated at various temperatures. In a next step the amount of iodine adsorbed at the metallic surface was determined as well at room temperature with varying iodine exposures as for constant exposure but varying temperatures. Furthermore, the zirconium iodide species resulting from the interaction of iodine with the Zircaloy-4 and desorbed at higher temperatures were identified by means of residual gas analysis. During these studies it was found that the oxidic overlayer of the passivated Zircaloy-4 samples is decomposed at temperatures above 200 0 C. The iodine uptake at metallic surfaces (cleaned by Ar-ion sputtering) at room temperature slows markedly down after formation of a closed zirconium-iodide overlayer and consequently the further reaction proceeds diffusion-controlled. At 200 0 C ZrI 4 is formed being the thermodynamically most stable Zr-iodide. During desorption experiments using iodine exposed Zircaloy-4 samples the release of ZrI 4 was proved. The results obtained from the various experiments are finally discussed with respect to the iodine-induced stress corrosion cracking process and the underlying basic mechanisms and a transport mechanism for the SCC in nuclear fuel rods is proposed. (orig./RB) [de

  15. Oxidation behaviour of Zr-Ce alloys. Kinetic and microstructure aspects

    International Nuclear Information System (INIS)

    Rouillon, Ludovic

    1996-01-01

    As Zircaloy alloys are used for fuel rods in pressurized water nuclear reactors, this research thesis aims at studying and improving corrosion resistance of zirconium alloys while maintaining their mechanical properties. It more precisely deals with the kinetic and microstructure aspects of the external corrosion of the cladding by the coolant. In the case of Zircaloys, this corrosion is characterized by a kinetic transition from an initially parabolic to a linear regime. This research aims at intervening on this transition by elaborating zirconium alloys containing an element which stabilizes zirconia, in this case cerium. After having reported a bibliographical study on sheath oxidation, on parameters which influence sheath oxidation kinetics, on zirconia stabilization by doping elements, on the interest of lanthanide oxides, the author reports a feasibility study on the use of cerium (choice and preparation, sintered ceramic characterization, annealing of stabilized zirconia), reports a metallurgical study of Zr-Ce alloys, reports the study of the oxidation behaviour of these alloys (in autoclave, in presence of oxygen, under oxygen and then water) and the characterization of the microstructures of the oxide layers. He finally discusses the relationship between microstructure and oxidation kinetics, the role of cerium in the oxidation process, and the role of water in the oxidation process [fr

  16. An attempt for a unified description of mechanical testing on Zircaloy-4 cladding subjected to simulated LOCA transient

    Directory of Open Access Journals (Sweden)

    Desquines Jean

    2016-01-01

    Full Text Available During a Loss Of Coolant Accident (LOCA, an important safety requirement is that the reflooding of the core by the emergency core cooling system should not lead to a complete rupture of the fuel rods. Several types of mechanical tests are usually performed in the industry to determine the degree of cladding embrittlement, such as ring compression tests or four-point bending of rodlets. Many other tests can be found in the open literature. However, there is presently no real intrinsic understanding of the failure conditions in these tests which would allow translation of the results from one kind of mechanical testing to another. The present study is an attempt to provide a unified description of the failure not directly depending on the tested geometry. This effort aims at providing a better understanding of the link between several existing safety criteria relying on very different mechanical testing. To achieve this objective, the failure mechanisms of pre-oxidized and pre-hydrided cladding samples are characterized by comparing the behavior of two different mechanical tests: Axial Tensile (AT test and “C”-shaped Ring Compression Test (CCT. The failure of samples in both cases can be described by usual linear elastic fracture mechanics theory. Using interrupted mechanical tests, metallographic examinations have evidenced that a set of parallel cracks are nucleated at the inner and outer surface of the samples just before failure, crossing both the oxide layer and the oxygen rich alpha layer. The stress intensity factors for multiple crack geometry are determined for both AT and CCT samples using finite element calculations. After each mechanical test performed on high temperature steam oxidized samples, metallography is then used to individually determine the crack depth and crack spacing. Using these two important parameters and considering the applied load at fracture, the stress intensity factor at failure is derived for each tested

  17. Plastic strain accumulation during asymmetric cyclic loading of Zircaloy-2 at room temperature

    International Nuclear Information System (INIS)

    Rajpurohit, R.S.; Santhi Srinivas, N.C.; Singh, Vakil

    2016-01-01

    Asymmetric cyclic loading leads to accumulation of cyclic plastic strain and reduces the fatigue life of components. This phenomenon is known as ratcheting fatigue. Zircaloy-2 is a important structural material in nuclear reactors and used as pressure tubes and fuel cladding in pressurized light and heavy water nuclear reactors. Due to power fluctuations, these components experience plastic strain cycles in the reactor and their life is reduced due to strain cycles. Power fluctuations also cause asymmetric straining of the material and leads to accumulation of plastic strain. The present investigation deals with the effect of the magnitude of mean stress, stress amplitude and stress rate on hardening/softening behavior of Zircaloy-2 under asymmetric cyclic loading, at room temperature. It was observed that plastic strain accumulation increased with mean stress and stress amplitude; however, it decreased with stress rate. (author)

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

    International Nuclear Information System (INIS)

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

    2012-04-01

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

  19. Hydride effect on crack instability of Zircaloy cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tseng, Che-Chung, E-mail: cctseng@iner.gov.tw [Institute of Nuclear Energy Research, No. 1000, Wunhua Road, Jiaan Village, Lungtan, Township, Taoyuan County 32546, Taiwan (China); Sun, Ming-Hung [Institute of Nuclear Energy Research, No. 1000, Wunhua Road, Jiaan Village, Lungtan, Township, Taoyuan County 32546, Taiwan (China); Chao, Ching-Kong [Department of Mechanical Engineering, National Taiwan University of Science and Technology, 43 Keelung Road, Section 4, Taipei 106, Taiwan (China)

    2014-04-01

    Highlights: • Radial hydrides near the crack tip had a significant effect on crack propagation. • For radial hydrides off the crack line vertically, the effect on crack propagation was notably reduced. • The longer hydride platelet resulted in a remarkable effect on crack propagation. • A long split in the radial hydride precipitate would enhance crack propagation. • The presence of circumferential hydride among radial hydrides may play an important role in crack propagation. - Abstract: A methodology was proposed to investigate the effect of hydride on the crack propagation in fuel cladding. The analysis was modeled based on an outside-in crack with radial hydrides located near its crack tip. The finite element method was used in the calculation; both stress intensity factor K{sub I} and J integral were applied to evaluate the crack stability. The parameters employed in the analysis included the location of radial hydride, hydride dimensions, number of hydrides, and the presence of circumferential hydride, etc. According to our study, the effective distance between a radial hydride and the assumed cladding surface crack for the enhancement of crack propagation proved to be no greater than 0.06 mm. For a hydride not on the crack line, it would induce a relatively minor effect on crack propagation if the vertical distance was beyond 0.05 mm. However, a longer hydride precipitate as well as double radial hydrides could have a remarkable effect on crack propagation. A combined effect of radial and circumferential hydrides was also discussed.

  20. Thin polycrystalline diamond films protecting zirconium alloys surfaces: From technology to layer analysis and application in nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Ashcheulov, P. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Škoda, R.; Škarohlíd, J. [Czech Technical University in Prague, Faculty of Mechanical Engineering, Technická 4, Prague 6, CZ-160 07 (Czech Republic); Taylor, A.; Fekete, L.; Fendrych, F. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Vega, R.; Shao, L. [Texas A& M University, Department of Nuclear Engineering TAMU-3133, College Station, TX TX 77843 (United States); Kalvoda, L.; Vratislav, S. [Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, Brehova 7, CZ-115 19, Prague 1 (Czech Republic); Cháb, V.; Horáková, K.; Kůsová, K.; Klimša, L.; Kopeček, J. [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Sajdl, P.; Macák, J. [University of Chemistry and Technology, Power Engineering Department, Technická 3, Prague 6, CZ-166 28 (Czech Republic); Johnson, S. [Nuclear Fuel Division, Westinghouse Electric Company, 5801 Bluff Road, Hopkins, SC 29209 (United States); Kratochvílová, I., E-mail: krat@fzu.cz [Institute of Physics, Academy of Sciences Czech Republic v.v.i, Na Slovance 2, CZ-182 21, Prague 8 (Czech Republic); Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, Brehova 7, CZ-115 19, Prague 1 (Czech Republic)

    2015-12-30

    Graphical abstract: - Highlights: • In this work we showed that films prepared by MW-LA-PECVD technology can be used as anticorrosion protective layer for Zircaloy2 nuclear fuel claddings at elevated temperatures (950 °C) when α phase of zirconium changes to β phase (more opened for oxygen/hydrogen diffusion). Quality of PCD films was examined by Raman spectroscopy, XPS, SEM, AFM and SIMS analysis. • The polycrystalline diamond films were of high quality - without defects and contaminations. After hot steam oxidation (950 °C) a high level of structural integrity of PCD layer was observed. Both sp{sup 2} and sp{sup 3} C phases were present in the protective PCD layer. Higher resistance and a lower degree of impedance dispersion was found in the hot steam oxidized PCD coated Zircaloy2 samples, which may suggest better protection of the Zircaloy2 surface. The PCD layer blocks the hydrogen diffusion into the Zircaloy2 surface thus protecting the material from degradation. • Hot steam oxidation tests confirmed that PCD coated Zircaloy2 surfaces were effectively protected against corrosion. Presented results demonstrate that the PCD anticorrosion protection can significantly prolong service life of Zircaloy2 nuclear fuel claddings in nuclear reactors even at elevated temperatures. - Abstract: Zirconium alloys can be effectively protected against corrosion by polycrystalline diamond (PCD) layers grown in microwave plasma enhanced linear antenna chemical vapor deposition apparatus. Standard and hot steam oxidized PCD layers grown on Zircaloy2 surfaces were examined and the specific impact of polycrystalline Zr substrate surface on PCD layer properties was investigated. It was found that the presence of the PCD coating blocks hydrogen diffusion into the Zircaloy2 surface and protects Zircaloy2 material from degradation. PCD anticorrosion protection of Zircaloy2 can significantly prolong life of Zircaloy2 material in nuclear reactors even at temperatures above Zr

  1. Development of nuclear fuel for the future -Development of performance improvement of the cladding by ion beam-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Byung Hoh; Jung, Moon Kyoo; Jung, Kee Suk; Kim, Wan; Lee, Jae Hyung; Song, Tae Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Han, Jun Kun [Sung Kyoon Kwan Univ., Seoul (Korea, Republic of); Kwon, Hyuk Sang [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1995-07-01

    In this research we analyzed the state of art related to the surface treatment method of nuclear fuel cladding for the development of the surface treatment technique of nuclear fuel cladding by ion beam while investigating major causes of the leakage of fuel rods. Ion implantation simulation code called TRIM-95 was used to decide basic parameters of ion beams and setup an appropriate process for ion implantation. Performance of the ion beam extraction was measured after adding the needed vacuum and cooling system to the existing gas and metal ion implanters. Target system for the ion implantation of fuel cladding improved and a plasma accelerator was installed on the target chamber of the metal ion implanter. The plasma accelerator is used to produce low energy, high current ion beams. The mechanical and chemical properties of the implanted Zircaloy-4 such as micro hardness, wear resistance, fretting wear, friction coefficient and corrosion resistance was measured under the room temperature and atmosphere. A micro structure and composition analysis of Zircaloy-4 sample was performed before and after the implantation to study the cause of the improvement in the mechanical and chemical characteristics. 94 figs, 11 tabs, 51 refs. (Author).

  2. 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.

  3. Superficial characterization by XP S of silver nanoparticles and their hydrothermal deposit over zircaloy

    International Nuclear Information System (INIS)

    Contreras R, A.; Gutierrez W, C.; Martinez M, I.; Medina A, A. L.

    2012-10-01

    The analysis technique of X-ray photoelectron spectroscopy (XP S) is sensitive exclusively to the first layers of the solids surface, which allows obtaining information about the chemical, physical and electronic properties of them. The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy pipes that contain in their interior pellets or uranium dioxide. In this work is studied the zircaloy surface, oxidized zircaloy under similar conditions to those of a reactor BWR type and oxidized zircaloy with a hydrothermal deposit of silver nanoparticles and zinc. The silver deposit is a proposal of the Materials Technology Department of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, which has the same objective that the noble metals deposit (Pt, Pd, and Rh) that is practiced in some of the reactors BWR, in order to mitigating the speed of crack growth for IGSCC in stainless steels 304 Ss. (Author)

  4. Determinations of the temperature of terminal solid solubility in dissolution and precipitation of hydrogen/deuterium in irradiated Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Vizcaino, P [CNEA-CONICET, Centro Atomico Ezeiza (Argentina)

    2012-07-01

    The proposed plan is an approach to the metallurgical consequences of the high neutron fluencies (10''2''2 n/cm''2) on the hydrogen behavior in zirconium based alloys, based on the significance of the microstructural behavior of the high burn up fuel claddings during the dry storage period. The studies are focused on Zircaloy-4, concerning to two processes: Neutron irradiation damage; Hydrogen pick up. The Zircaloy-4 was taken from cooling channels of the PHWR Atucha 1. These components remained more than 10 years in service, reaching neutron fluencies up to 10''2''2 n/cm''2. In the last recent years, measurements of the hydride dissolution temperatures have shown that hydrogen solubility is affected by the neutron irradiation, increasing it respect to the unirradiated Zircaloy solubility. In addition, in this material the amorphization/dissolution of the second phase particles (SPPs) was observed, being proposed an interaction between the hydrogen atoms, the SPPs and the irradiation defects as a possible explanation of the observed behavior. For the present case, attention will be focused on the hydride precipitation process, since it is strongly related with delay hydrogen cracking initiation, a problem of direct concern for the dry storage. The goal of the present proposal is to make an approach to the source of the observed effect, applying several specific techniques as differential scanning calorimetry (DSC), high resolution x-ray diffraction and transmission electron microscopy. The objectives can be divided as follows: Determination of the temperatures of terminal solid solubility in dissolution (TTSSd) and in precipitation (TTSSp) in high fluency irradiated Zircaloy-4, reproducing the temperatures at which the Zircaloy fuel claddings remain during dry storage by an annealing program during the DSC experiments; Observations by optical and transmission electron microscopy of the hydride distribution before (as received material) and after high temperature

  5. The steam pressure effect on high temperature corrosion of zircaloy-4

    International Nuclear Information System (INIS)

    Kim, K. P.; Park, G. H.

    1998-01-01

    To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The degree of high temperature oxidation of zircaloy-4 was measured at three different conditions, high pressure steam, high pressure Ar gas with small amount of steam, and 1 atm steam. All the measurements were done at 750 deg C. The oxide thickness is much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. No effect was observed in the case of high pressure Ar containing small amount of steam. Many cracks exist on the surface of specimens oxidized at high pressure steam, which come from the enhanced tetragonal to monoclinic phase transformation due to high pressure steam. The enhanced oxidation seems to oxide cracking

  6. A phenomenological model for iodine stress corrosion cracking of zircaloy

    International Nuclear Information System (INIS)

    Miller, A.K.; Tasooji, A.

    1981-01-01

    To predict the response of Zircaloy tubing in iodine environments under conditions where either crack initiation or crack propagation predominates, a unified model of the SCC process has been developed based on the local conditions (the local stress, local strain, and local iodine concentration) within a small volume of material at the cladding inner surface or the crack tip. The methodology used permits computation of these values from simple equations. A nonuniform distribution of local stress and strain results once a crack has initiated. The local stress can be increased due to plastic constraint and triaxiality at the crack tip. Iodine penetration is assumed to be a surface diffusion-controlled process. Experimental data are used to derive criteria for intergranular failure, transgranular failure, and ductile rupture in terms of the local conditions. The same failure criteria are used for both crack initiation and crack propagation. Irradiation effects are included in the model by changing the value of constants in the equation governing iodine penetration and by changing the values used to represent the mechanical properties of the Zircaloy. (orig./HP)

  7. 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

  8. 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)

  9. Studies of irradiated zircaloy fuel sheathing using XPS

    Energy Technology Data Exchange (ETDEWEB)

    Chan, P K; Irving, K G [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Hocking, W H; Duclos, A M; Gerwing, A F [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO{sub 2}) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs.

  10. Studies of irradiated zircaloy fuel sheathing using XPS

    International Nuclear Information System (INIS)

    Chan, P.K.; Irving, K.G.; Hocking, W.H.; Duclos, A.M.; Gerwing, A.F.

    1995-01-01

    The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO 2 ) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs

  11. Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins

    International Nuclear Information System (INIS)

    Makenas, B.J.; Lawrence, L.A.; Jensen, B.W.

    1982-05-01

    Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 (E > .1 MeV) and 8.8 x 10 22 n/cm 2 (E > .1 MeV) respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %

  12. Zircaloy-4 stress corrosion by iodine: crack kinetics and influence of irradiation on the crack initiation

    International Nuclear Information System (INIS)

    Serres, A.

    2008-01-01

    During the PWR power transients, iodine-induced stress corrosion cracking (I-SCC) is one of the potential failure modes of Zircaloy-4 fuel claddings under Pellet-Cladding Interaction conditions. The primary objective of this study is to distinguish the parameters that contribute to the I-SCC phenomenon in iodized methanol solutions at ambient temperature, on notched tensile specimens, using crack growth rate measurements provided by Direct Current Potential Drop. The results show that for a KI lower than 20 MPa.m 1/2 , the IG and mixed IG/TG velocity of propagation is a linear function of KI, regardless of the propagation mode. Between 20 and 25 MPa.m 1/2 , the TG crack growth rate also depends linearly on KI, but increases at a faster rate with respect to KI than during the IG and mixed IG/TG propagation steps. The crack propagation direction and plane (LT and TL) have an impact on the propagation modes, but no impact on the kinetics. The increase of iodine content induces an increase of the crack growth rate for a given KI, and a decrease of the KI, threshold, allowing the crack propagation. This work enables us to quantify the effect of iodine content and of KI on the crack propagation step, propose a propagation law taking into accounts these parameters, and improve the I-SCC description for models. During operation, a zirconium cladding is neutron-irradiated, modifying its microstructure and deformation modes. The second objective of the study is therefore to investigate the impact of these modifications on I-SCC. For that purpose, smooth specimens in recrystallized Zircaloy-4 are proton-irradiated to 2 dpa at 305 C, the microstructure and deformation modes of unirradiated and irradiated Zircaloy-4 are characterized by TEM and SEM, and the influence of these radiation-induced modifications on the I-SCC susceptibility is studied. The Laves phases precipitates are slightly modified by irradiation. The formation of P -type dislocation loops correlated with

  13. Evolution of processing of GE fuel clad tubing for corrosion resistance in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Williams, C.D. [GE Nuclear Energy, Wilmington, NC (United States); Adamson, R.B. [GE Nuclear Energy, Wilmington, NC (United States); Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Plaza-Meyer, E. [GE Nuclear Energy, Wilmington, NC (United States); Proebstle, R.A. [GE Nuclear Energy, Wilmington, NC (United States); White, D.W. [GE Nuclear Energy, Wilmington, NC (United States)

    1996-05-01

    The current modification of the primary GE in-process solution-quench heat treatment, an (alpha+beta) solution-quench carried out at a tube diameter requiring only two subsequent reduction and anneal cycles, is applicable to Zr barrier fuel clad tubing, to non-barrier fuel clad tubing, and to the TRICLAD tubing product. A combination of good in-reactor corrosion performance and degradation resistance is anticipated for these products, based on knowledge of metallurgical characteristics and supported by the demonstrated performance capability of the Zircaloy-2 materials used. (orig.)

  14. Studies of Corrosion of Cladding Materials in Simulated BWR-environment Using Impedance Measurements. Part I: Measurements in the Pre-transition Region

    International Nuclear Information System (INIS)

    Forsberg, Stefan; Ahlberg, Elisabet; Andersson, Ulf

    2004-09-01

    The corrosion of three Zircaloy 2 cladding materials, LK2, LK2+ and LK3, have been studied in-situ in an autoclave using electrochemical impedance spectroscopy. Measurements were performed in simulated BWR water at temperatures up to 288 deg C. The impedance spectra were successfully modelled using equivalent circuits. When the oxide grew thicker during the experiments, a change-over from one to two time constants was seen, showing that a layered structure was formed. Oxide thickness, oxide conductivity and effective donor density were evaluated from the impedance data. The calculated oxide thickness at the end of the experiments was consistent with the value obtained from SEM. It was shown that the difference in oxide growth rate between the investigated materials is small in the pre-transition region. The effective donor density, which is a measure of electronic conductivity, was found to be lower for the LK3 material compared to the other two materials

  15. Oxidation-induced deformation of zircaloy-4 tubing in steam in the temperature range 600-1000 degree C

    International Nuclear Information System (INIS)

    Aly, A.E.; Hussein, A.G.; EL-Raghy, S.M.; EL-Sayed, A.A.; EL-Banna, O.A.

    1992-01-01

    The oxidation-induced deformation of zircaloy-4 (zry-4) tubing in steam has been studied in the temperature range 600 to 1000 degree C. The induced deformation has been measured in both radial and axial directions of the tube. The effect of hydrogen addition to steam was also investigated. The oxidation-induced deformation has been characterized by uniform and non-uniform (distortion) strain period. During the uniform strain period the radial strain kinetics were found in general, to be parallel to the oxidation kinetics. The axial strain (δA) induced by oxidation was found to be always lower than the radial strain (εR). The addition of 5% by volume hydrogen to steam leads to an increase in the oxidation rate and to a decrease in the degree of anisotropy between radial and axial strains

  16. A model for hydrogen pickup for BWR cladding materials

    International Nuclear Information System (INIS)

    Hede, G.; Kaiser, U.

    2001-01-01

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

  17. Development of Cr cold spray–coated fuel cladding with enhanced accident tolerance

    Directory of Open Access Journals (Sweden)

    Martin Ševeček

    2018-03-01

    Full Text Available Accident-tolerant fuels (ATFs are currently of high interest to researchers in the nuclear industry and in governmental and international organizations. One widely studied accident-tolerant fuel concept is multilayer cladding (also known as coated cladding. This concept is based on a traditional Zr-based alloy (Zircaloy-4, M5, E110, ZIRLO etc. serving as a substrate. Different protective materials are applied to the substrate surface by various techniques, thus enhancing the accident tolerance of the fuel. This study focuses on the results of testing of Zircaloy-4 coated with pure chromium metal using the cold spray (CS technique. In comparison with other deposition methods, e.g., Physical vapor deposition (PVD, laser coating, or Chemical vapor deposition techniques (CVD, the CS technique is more cost efficient due to lower energy consumption and high deposition rates, making it more suitable for industry-scale production. The Cr-coated samples were tested at different conditions (500°C steam, 1200°C steam, and Pressurized water reactor (PWR pressurization test and were precharacterized and postcharacterized by various techniques, such as scanning electron microscopy, Energy-dispersive X-ray spectroscopy (EDX, or nanoindentation; results are discussed. Results of the steady-state fuel performance simulations using the Bison code predicted the concept's feasibility. It is concluded that CS Cr coating has high potential benefits but requires further optimization and out-of-pile and in-pile testing. Keywords: Accident-Tolerant Fuel, Chromium, Cladding, Coating, Cold Spray, Nuclear Fuel

  18. Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25 C and 480 C under various stress states, including RIA loading conditions

    International Nuclear Information System (INIS)

    Le Saux, M.; Carassou, S.; Averty, X.; Le Saux, M.; Besson, J.; Poussard, C.

    2010-01-01

    The anisotropic plastic behavior and the fracture of as-received and hydrided Cold-Worked Stress Relieved Zircaloy-4 cladding tubes are investigated under thermal-mechanical loading conditions representative of Pellet-Clad Mechanical Interaction during Reactivity Initiated Accidents in Pressurized Water Reactors. In order to study the combined effects of temperature, hydrogen content, loading direction and stress state, Axial Tensile, Hoop Tensile, Expansion Due to Compression and hoop Plane Strain Tensile tests are performed at room temperature, 350 C and 480 C on the material containing various hydrogen contents up to 1200 wt. ppm (hydrides are circumferential and homogeneously distributed). These tests are combined with digital image correlation and metallographic and fractographic observations at different scales. The flow stress of the material decreases with increasing temperature. The material is either strengthened or softened by hydrogen depending on temperature and hydrogen content. Plastic anisotropy depends on temperature but not on hydrogen content. The ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking. The plastic strain that leads to hydride fracture at room temperature decreases with increasing hydrogen content. The influence of stress triaxiality on hydride cracking is negligible in the studied range. The influence of hydrogen on material ductility is negligible at 350 C and 480 C since hydrides do not crack at these temperatures. The ductility of the material increases with increasing temperature. The evolution of material ductility is associated with a change in both the macroscopic fracture mode of the specimens and the microscopic failure mechanisms. (authors)

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

    Directory of Open Access Journals (Sweden)

    de Souza Gomes Daniel

    2016-01-01

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

  20. Development of a deformation and failure model for Zircaloy at high temperatures for light water reactor loss-of-coolant-accident investigations

    International Nuclear Information System (INIS)

    Raff, S.

    1982-11-01

    To describe Zircaloy-4 deformation and failure behaviour at high temperatures (600 to 1400 0 C), the phenomenological model NORA was developed and verified against numerous experimental results. The model can be applied to the calculation of fuel rod cladding deformation during small and large break loss-of-coolant-accidents. (orig./RW) [de

  1. Compatibility studies on Mo-coating systems for nuclear fuel cladding applications

    Science.gov (United States)

    Koh, Huan Chin; Hosemann, Peter; Glaeser, Andreas M.; Cionea, Cristian

    2017-12-01

    To improve the safety factor of nuclear power plants in accident scenarios, molybdenum (Mo), with its high-temperature strength, is proposed as a potential fuel-cladding candidate. However, Mo undergoes rapid oxidation and sublimation at elevated temperatures in oxygen-rich environments. Thus, it is necessary to coat Mo with a protective layer. The diffusional interactions in two systems, namely, Zircaloy-2 (Zr2) on a Mo tube, and iron-chromium-aluminum (FeCrAl) on a Mo rod, were studied by aging coated Mo substrates in high vacuum at temperatures ranging from 650 °C to 1000° for 1000 h. The specimens were characterized using scanning electron microscopy (SEM), energy-dispersive spectrometry (EDS) and nanoindentation. In both systems, pores in the coating increased in size and number with increasing temperature over time, and cracks were also observed; intermetallic phases formed between the Mo and its coatings.

  2. A study on the reaction of Zircaloy-4 tube with hydrogen/steam mixture

    Science.gov (United States)

    Lee, Ji-Min; Kook, Dong-Hak; Cho, Il-Je; Kim, Yong-Soo

    2017-08-01

    In order to fundamentally understand the secondary hydriding mechanism of zirconium alloy cladding, the reaction of commercial Zircaloy-4 tubes with hydrogen and steam mixture was studied using a thermo-gravimetric analyser with two variables, H2/H2O ratio and temperature. Phenomenological analysis revealed that in the steam starvation condition, i.e., when the H2/H2O ratio is greater than 104, hydriding is the dominant reaction and the weight gain increases linearly after a short incubation time. On the other hand, when the gas ratio is 5 × 102 or 103, both hydriding and oxidation reactions take place simultaneously, leading to three distinct regimes: primary hydriding, enhanced oxidation, and massive hydriding. Microstructural changes of oxide demonstrate that when the weight gain exceeds a certain critical value, massive hydriding takes place due to the significant localized crack development within the oxide, which possibly simulates the secondary hydriding failure in a defective fuel operation. This study reveals that the steam starvation condition above the critical H2/H2O ratio is only a necessary condition for the secondary hydriding failure and, as a sufficient condition, oxide needs to grow sufficiently to reach the critical thickness that produces substantial crack development. In other words, in a real defective fuel operation incident, the secondary failure is initiated only when both steam starvation and oxide degradation conditions are simultaneously met. Therefore, it is concluded that the indispensable time for the critical oxide growth primarily determines the triggering time of massive hydriding failure.

  3. Method for decontaminating stainless cladding tubes

    International Nuclear Information System (INIS)

    Komatsu, Fumiaki.

    1986-01-01

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

  4. Mechanical properties and structure of Zircaloy attached by UO2+x and fission products

    International Nuclear Information System (INIS)

    Holub, F.

    1987-08-01

    The aim of this project was to determine the combined long-term effect of simulated fission products and hyperstoichiometric uranium dioxide on the mechanical properties and structure of Zircaloy. Three groups of fission product elements or compounds were defined: The rare earth oxides CeO 2 , La 2 O 3 , Nd 2 O 3 , Y 2 O 3 ; The metals No, Ru, Ag; The low melting elements Te, Sb and Cd. Each of these groups of fission products was mixed with UO 2+x in proportion related for burnups of 5, 10 and 30%. The simulated fuel mixtures were filled into tubular Zircaloy casings, plugged and welded. These specimens were annealed at 350, 500 and 700 deg. C up to 17,500 hours. The test results indicate different kinds of action of the simulated fuel constituents. Mixtures of rare earth oxides and UO 2+x embrittle Zircaloy drastically at higher temperatures. There exists a mutual intensifying effect of rare earth oxides and UO 2+x . UO 2+x and (Mo + Ru + Ag) and their mixtures act very similar on Zircaloy. The low melting fission products (Te + Sb + Cd) influence the ductility of Zircaloy in an advantageous manner, compared to pure UO 2+x fuel. The layer of zirconium tellurides seems to protect the Zircaloy metal against the embrittling attack of oxygen from UO 2+x . The most important events of tensile tests at 400 deg. C are the high values of the elongation of specimens which are brittled at room temperature. It should guarantee the integrity of fuel elements, which have been attacked chemically by fission products at temperatures of 400 deg. C and higher

  5. Comparison of high temperature steam oxidation behavior of Zircaloy-4 versus austenitic and ferritic steels under light water reactor safety aspects

    International Nuclear Information System (INIS)

    Leistikow, S.; Schanz, G.; Zurek, Z.

    1985-12-01

    A comparative study of the oxidation behavior of Zy-4 versus steel No. 1.4914 and steel No. 1.4970 was performed in high temperature steam. Reactor typical tube sections of all three materials were exposed on both sides to superheated steam at temperatures ranging from 600 to 1300 0 C for up to 6 h. The specimens were evaluated by gravimetry, metallography, and other methods. The results are presented in terms of weight gain, corresponding metal (wall) penetration and consumption as function of time and temperature. Concerning the corrosion resistance the ranking position of Zy-4 was between the austenitic and the ferritic steel. Because of the chosen wall dimensions Zy-4 and the austenitic steel behaved similarly in that the faster oxidation of the thicker Zy-4 cladding consumed the total wall thickness in a time equivalent to the slower oxidation of the thinner austenitic steel cladding. The ferritic steel cladding however was faster consumed because of the lower oxidation resistance and the thinner wall thickness compared to the austenitic steel. So besides oxide scale formation, oxygen diffusion into the bulk of the metal forming various oxygen-containing phases were evaluated - also in respect to their influence on mechanical cladding properties and the dimensional changes. (orig./HP) [de

  6. 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.

  7. 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

  8. Conditioning of high activity solid waste: fuel claddings and dissolution residues

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    This chapter reports on experimental studies of embedding into matrix material, the melting and conversion of zircaloy, and waste properties and characterization. Methods are developed for embedding the waste scrap into a solid and resistant matrix material in order to confine the radioactivity and to prevent it from dispersion. The matrix materials investigated are lead alloys, ceramics and compacted graphite or aluminium powder. The treatment of fuel hulls by melting or chemical conversion is described. Cemented hulls are characterized and the pyrophoricity of zircaloy fines is investigated. Topics considered include the embedding of hulls into graphite and aluminium, the embedding of hulls and dissolution residues into alumino-ceramics, the solidification of alpha-bearing wastes into a ceramic matrix, and the conditioning of cladding waste by eutectoidic melting and by embedding in glass

  9. On the corrosion behavior of zircaloy-4 in spent fuel pools under accidental conditions

    International Nuclear Information System (INIS)

    Lavigne, O.; Shoji, T.; Sakaguchi, K.

    2012-01-01

    Highlights: ► Corrosion behavior of oxidized Zr-4 in alkaline media in presence of chloride and radical forms. ► Generation of radical forms by sonolysis of water. ► Limited increase of the passive current densities under polarization with the increase of pH and the presence of radicals. ► Decrease of the passive range of oxidized Zr-4 with presence of Cl − (E pit ∼ 0.6 V SCE ). ► Decrease of the pitting potential when oxide layer is scratched or damaged (E pit ∼ 0.16 V SCE ). - Abstract: After zircaloy cladding tubes have been subjected to irradiation in the reactor core, they are stored temporarily in spent fuel pools. In case of an accident, the integrity of the pool may be affected and the composition of the coolant may change drastically. This was the case in Fukushima Daiichi in March 2011. Successive incidents have led to an increase in the pH of the coolant and to chloride contamination. Moreover, water radiolysis may occur owing to the remnant radioactivity of the spent fuel. In this study, we propose to evaluate the corrosion behavior of oxidized Zr-4 (in autoclave at 288 °C for 32 days) in function of the pH and the presence of chloride and radical forms. The generation of radicals is achieved by the sonolysis of the solution. It appears that the increase in pH and the presence of radicals lead to an increase in current densities. However, the current densities remain quite low (depending on the conditions, between 1 and 10 μA cm −2 ). The critical parameter is the presence of chloride ions. The chloride ions widely decrease the passive range of the oxidized samples (the pitting potential is measured around +0.6 V (vs. SCE)). Moreover, if the oxide layer is scratched or damaged (which is likely under accidental conditions), the pitting potential of the oxidized sample reaches the pitting potential of the non-oxidized sample (around +0.16 V (vs. SCE)), leaving a shorter stable passive range for the Zr-4 cladding tubes.

  10. Adsorption and diffusion of hydrogen in Zircaloy-4

    International Nuclear Information System (INIS)

    Torres, E.; Desquines, J.; Baietto, M.C.; Coret, M.; Wehling, F.; Blat-Yrieix, M.; Ambard, A.

    2015-01-01

    Hydrogen in zirconium alloys is considered in many nuclear safety issues. Below 500 Celsius degrees, rather limited knowledge is available on the combined hydrogen adsorption at the sample surface and diffusion in the metal. A modeling of hydrogen gaseous charging has been established starting with a set of relevant laws and parameters derived from open literature. Simulating the hydrogen charging process requires simultaneous analysis of gaseous surface adsorption, hydrogen solid-solution diffusion and precipitation, when exceeding the material solubility limit. The modeling has been extended to reproduce the solid-gas exchange. Gaseous charging experiments have been performed at 420 C. degrees on Stress Relieved Annealed (SRA) Zircaloy-4 cladding samples to validate the model. The sample hydrogen content has been systematically measured after charging and compared to the calculated value thus providing a validation of the adsorption modeling. Complementary tests have been carried out on Recrystallized Annealed (RXA) Zircaloy-4 rods to characterize the combined diffusion and adsorption process. The hydrogen concentration distribution has been characterized using an inverse technique based on destructive analyses of the samples. This additional set of data was relevant for the validation of the hydrogen combined adsorption/diffusion modeling up to 420 C. degrees. (authors)

  11. Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

    International Nuclear Information System (INIS)

    Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

    1979-01-01

    This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

  12. Mechanical properties of zircaloy-4 tubes for CAREM 25 fuel rods

    International Nuclear Information System (INIS)

    Juarez, G; Bianchi, D; Flores, A; Vizcaino, P

    2012-01-01

    The aim of the present work was giving support to the development of Zircaloy-4 fuel claddings for the CAREM 25 reactor through microstructural and mechanical properties studies along the manufacturing process. The manufacturing route was defined in 4 cold rolling stages and two thermal treatments, one at the middle and one after the last rolling stage. The first two rolling stages were performed in FAESA and the last two in PPFAE-CNEA using the rolling machine HPTR 8-15. The reference values for the evaluation were those indicated in the technical specification CAREM25 F ET-3-B0610. In this context, four tubes were received from FAESA. To these tubes mechanical properties determinations were performed to characterize the material in each step performed in PPFAE. The mechanical properties of the cladding tubes also achieve the standard values (σ 0.2 = 450 MPa, e = 15%), being σ 0.2 = 530 MPa and 18% the elongation (author)

  13. Measurement of dose rate and estimation of beta activity in zircaloy hull drum

    International Nuclear Information System (INIS)

    Pandey, J.P.N.; Kumar, Pankaj; Shinde, A.M.; Purohit, R.G.; Sarkar, P.K.

    2012-01-01

    Fuel Reprocessing Plant is designed for the processing of spent fuel from reactor for the recovery of plutonium and uranium as PuO 2 and U 3 O 8 respectively. Zircaloy is used as cladding material of natural uranium fuel pins used in the reactors. In reprocessing plants chop and leach method is used to remove the zircaloy clad from the fuel matrix during Head End Treatment. Initially spent fuel bundles are chopped into pieces and collected in perforated baskets kept in dissolvers. All chopped pieces are dissolved in HNO 3 in the dissolvers followed by heating and boiling. Dissolved solutions are transferred to Filtrate Tank (FT) leaving behind un-dissolved zircoloy hull pieces in the dissolver baskets. Un-dissolved and almost dry hull pieces are transferred in hull drum from the dissolver baskets using the Hull Tilting Facility. Hull drums are made of stainless steel having 500 litre capacity and two third of its volume is filled with zircoloy pieces. Hull drums filled with hull pieces are loaded in Hull Removal Cask (HRC) and transported to SWMF (Solid Waste Management Facility) site for interim storage/disposal in tile holes. Hull pieces are high active solid wastes which contain significant amount of fission products. Radiation levels on hull drums are in the range of few hundreds of mGy/h which has high potential of external hazards if not handled properly. Therefore hull drums are handled remotely in specially designed lead shielded cask

  14. Characterization of hydrogen, nitrogen, oxygen, carbon and sulfur in nuclear fuel (UO2) and cladding nuclear rod materials

    International Nuclear Information System (INIS)

    Crewe, Maria Teresa I.; Lopes, Paula Corain; Moura, Sergio C.; Sampaio, Jessica A.G.; Bustillos, Oscar V.

    2011-01-01

    The importance of Hydrogen, Nitrogen, Oxygen, Carbon and Sulfur gases analysis in nuclear fuels such as UO 2 , U 3 O 8 , U 3 Si 2 and in the fuel cladding such as Zircaloy, is a well known as a quality control in nuclear industry. In UO 2 pellets, the Hydrogen molecule fragilizes the metal lattice causing the material cracking. In Zircaloy material the H2 molecules cause the boiling of the cladding. Other gases like Nitrogen, Oxygen, Carbon and Sulfur affect in the lattice structure change. In this way these chemical compounds have to be measure within specify parameters, these measurement are part of the quality control of the nuclear industry. The analytical procedure has to be well established by a convention of the quality assurance. Therefore, the Oxygen, Carbon, Sulfur and Hydrogen are measured by infrared absorption (IR) and the nitrogen will be measured by thermal conductivity (TC). The gas/metal analyzer made by LECO Co. model TCHEN-600 is Hydrogen, Oxygen and Nitrogen analyzer in a variety of metals, refractory and other inorganic materials, using the principle of fusion by inert gas, infrared and thermo-coupled detector. The Carbon and Sulfur compounds are measure by LECO Co. model CS-400. A sample is first weighed and placed in a high purity graphite crucible and is casted on a stream of helium gas, enough to release the oxygen, nitrogen and hydrogen. During the fusion, the oxygen present in the sample combines with the carbon crucible to form carbon monoxide. Then, the nitrogen present in the sample is analyzed and released as molecular nitrogen and the hydrogen is released as gas. The hydrogen gas is measured by infrared absorption, and the sample gases pass through a trap of copper oxide which converts CO to CO 2 and hydrogen into water. The gases enter the cell where infrared water content is then converted making the measurement of total hydrogen present in the sample. The Hydrogen detection limits for the nuclear fuel is 1 μg/g for the Nitrogen

  15. UO2 - Zr chemical interaction of PHWR fuel pins under high temperature

    International Nuclear Information System (INIS)

    Majumdar, P.; Mukhopadhyay, D.; Gupta, S.K.

    2001-01-01

    At high temperature Zircaloy clad interacts with the UO 2 fuel as well as with the steam to produce oxide layer of a-Zr(O) and ZrO 2 . This layer formation significantly reduces the structural strength of the clad. A computer code SFDCPA/MOD1 has been developed to simulate the interaction and predict the oxide layer thickness for any accidental transient condition. It is well validated with published experimental data on the isothermal and transient temperature condition. The program is applied to Indian Pressurized Heavy Water Reactor (PHWR) fuel pin under certain severe transient condition where it experiences temperature above 1000 C. The study gives an idea of the un-oxidized thickness of Zircaloy, which is an important criterion for fuel integrity. (author)

  16. 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

  17. Nb effect on Zr-alloy oxidation under high pressure steam at high temperatures

    International Nuclear Information System (INIS)

    Park, Kwangheon; Yang, Sungwoo; Kim, Kyutae

    2005-01-01

    The high-pressure steam effects on the oxidation of Zircaloy-4 (Zry-4) and Zirlo (Zry-1%Nb) claddings at high temperature have been analyzed. Test temperature range was 700-900degC, and pressures were 1-150 bars. High pressure-steam enhances oxidation of Zry-4, and the dependency of enhancement looks exponential to steam pressure. The origin of the oxidation enhancement turned out to be the formation of cracks in oxide. The loss of tetragonal phase by high-pressure steam seems related to the crack formation. Addition of Nb as an alloying element to Zr alloy reduces significantly the steam pressure effects on oxidation. The higher compressive stresses and the smaller fraction of tetragonal oxides in Zry-1%Nb seem to be the diminished effect of high-pressure steam on oxidation. (author)

  18. Utilising DualEELS to probe the nanoscale mechanisms of the corrosion of Zircaloy-4 in 350 °C pressurised water

    Energy Technology Data Exchange (ETDEWEB)

    Annand, Kirsty J., E-mail: k.annand.1@research.gla.ac.uk [Materials and Condensed Matter Physics, School of Physics and Astronomy, University of Glasgow, Glasgow G12 8QQ (United Kingdom); MacLaren, Ian [Materials and Condensed Matter Physics, School of Physics and Astronomy, University of Glasgow, Glasgow G12 8QQ (United Kingdom); Gass, Mhairi [AMEC Foster Wheeler, Clean Energy, Walton House, Birchwood Park, WA3 6GA (United Kingdom)

    2015-10-15

    Characterisation of materials utilised for fuel cladding in nuclear reactors prior to service is integral in order to understand corrosion mechanisms which would take place in reactor. Zircaloy-4 is one such material of choice for nuclear fuel containment in Pressurised Water Reactors (PWRs). In particular, the metal-oxide interface has been a predominant focus of previous research, however, due to the complex oxidation process of zirconium cladding, there is still no clear understanding of what is present at the interface. Using Scanning Transmission Electron Microscopy (STEM) and Dual Electron Energy Loss Spectroscopy (DualEELS), we have studied the corrosion of this material under conditions similar to those that could be encountered in service. It is shown that under all conditions, whether during faster oxidation in the early stages, slow growth just prior to the transition to a new growth regime, or in the faster growth that happens after this transition, the surface of the metal below the scale is loaded with oxygen up to around 33 at%. Approaching transition, in conditions of slow growth and slow oxygen supply, an additional metastable suboxide is apparent with a thickness of tens of nm. By studying changes in both chemical composition and dielectric function of the material at the oxide scale – metal interface with nanometre resolution, quantitative mapping could be achieved, clearly showing that this is a suboxide composition of ZrO and a Zr oxidation state close to +2. - Highlights: • Metal-oxide interface evolution studied by few-nm resolution EELS mapping. • Low loss EELS is very effective for mapping phase evolution. • ZrO suboxide phase characterised using EELS, which fits theoretical predictions. • ZrO formation found to correlate to local growth rate of oxide scale.

  19. Microstructure and high temperature oxidation resistance of Ti-Ni gradient coating on TA2 titanium alloy fabricated by laser cladding

    Science.gov (United States)

    Liu, Fencheng; Mao, Yuqing; Lin, Xin; Zhou, Baosheng; Qian, Tao

    2016-09-01

    To improve the high temperature oxidation resistance of TA2 titanium alloy, a gradient Ni-Ti coating was laser cladded on the surface of the TA2 titanium alloy substrate, and the microstructure and oxidation behavior of the laser cladded coating were investigated experimentally. The gradient coating with a thickness of about 420-490 μm contains two different layers, e.g. a bright layer with coarse equiaxed grain and a dark layer with fine and columnar dendrites, and a transition layer with a thickness of about 10 μm exists between the substrate and the cladded coating. NiTi, NiTi2 and Ni3Ti intermetallic compounds are the main constructive phases of the laser cladded coating. The appearance of these phases enhances the microhardness, and the dense structure of the coating improves its oxidation resistance. The solidification procedure of the gradient coating is analyzed and different kinds of solidification processes occur due to the heat dissipation during the laser cladding process.

  20. Biaxial creep deformation of Zircaloy-4 in the high alpha phase temperature range

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The ballooning response of Zircaloy-4 fuel tubes during a postulated loss-of-coolant accident may be calculated from a knowledge of the thermal environment of the rods and the creep deformation characteristics of the cladding. In support of such calculations biaxial creep studies have been performed on fuel tubes supplied by Westinghouse, Wolverine and Sandvik of temperatures in the alpha phase range. This paper presents the results of an investigation of their respective creep behaviour which has resulted in the formulation of equations for use in LOCA fuel ballooning codes. (author)

  1. A Prediction Study on Oxidation of Aluminum Alloy Cladding of U{sub 3}Si{sub 2}-Al Fuel Plate

    Energy Technology Data Exchange (ETDEWEB)

    Tahk, Y.W.; Lee, B.H.; Oh, J.Y.; Park, J.H.; Yim, J.S. [Research Reactor Design and Engineering Div., Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of)

    2011-07-01

    U{sub 3}Si{sub 2}-Al dispersion fuel with aluminum alloy cladding will be used for the Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding undergoes corrosion at slow rates under operational status. This causes thinning of the cladding walls and impairs heat transfer to the coolant. Predictions of the aluminum oxide thickness of the fuel cladding and the maximum temperature difference across the oxide film are needed for reliability evaluation based on the design criteria and limits which prohibit spallation of oxide film. In this work, several oxide thickness prediction models were compared with the measured data of in-pile test results from RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model were performed for JRTR fuel. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, fresh fuel is discharged after 900 effective full power days (EFPD), which is too long a span to predict oxidation properly without an elaborate model. The latest model developed by Kim et al. is in good agreement with the recent in-pile test data as well as with the out-of-pile test data available in the literature, and is one of the best predictors for the oxidation of aluminum alloy cladding in various operating condition. Accordingly, this model was chosen for estimating the oxide film thickness. Through the preliminarily evaluation, water pH level is to be controlled lower than 6.2 for the conservativeness in the case of including the effect of anticipated operational occurrences and the spent fuel residence time in the storage rack after discharging. (author)

  2. Light water reactor mixed-oxide fuel irradiation experiment

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

    1998-01-01

    The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

  3. Fatigue characteristics of ODS surface treated Zircaloy-4

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sun Han; Jung, Yan gIl; Park, Dong Jun; Park, Jung Hwan; Kim, Hyun Gil; Yang, Jae Ho; Koo, Yang Hyun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Various accident tolerant fuel (ATF) cladding concepts are considered and have being developed to increase the oxidation resistance and ballooning/ rupture resistance of current Zr-based cladding material under accident conditions. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures. ODS treatment is a way of improve the high temperature- oxidation resistant and mechanical stress by disperse the hardened particles inside of metal to interrupt the movement of the electric potential. In this study, the accident tolerance improved zirconium alloy by the ODS surface treatment was evaluated for the fatigue characteristics which is one of the significant items of the integrity assessment.

  4. Characterization of Zircaloy-2 and Zircaloy-4 by X-Ray fluorescence

    International Nuclear Information System (INIS)

    Sato, I.M.; Imakuma, K.; Salvador, V.L.R.

    1981-03-01

    The analytical characterization of zircaloy-2 and zircaloy-4 is intimataly connected with the determination of Sn, Fe, Cr, Ni, O, N, H, and Hf. An analytical method developed in this laboratory is discribed for the determination of metallic elements like Sn, Fe, Cr and Ni using the technique of X-ray fluorescence. The samples are prepared in the form of double-layer pellets using boric acid as a binding agent. The zircaloy-4 is dissolved in hydrofluoric acid and the metallic elements are converted to fluorides. The standard samples used for calibration are prepared from synthetic materials. The elements are determined by measuring the characteristic first order K α lines. A Zircaloy-4 sample analysed yielded the following values: Sn=1.30+-0.03%, Fe=0.18+-0.01%, Cr=0.088+-0.004% and Ni=14+-3 ppm. The reproducibility, precision, as well as the theoretical limit of detection of the method are discussed. The determination of the elements O, N and H present as occluded gas in the zircaloy is nearing completion. These analyses are being carried out by a Mass Spectrometric technique where an aliquot of the released gas is analysed. (Author) [pt

  5. The maximum allowable temperature of zircaloy-2 fuel cladding under dry storage conditions

    International Nuclear Information System (INIS)

    Mayuzumi, M.; Yoshiki, S.; Yasuda, T.; Nakatsuka, M.

    1990-09-01

    Japan plans to reprocess and reutilise the spent nuclear fuel from nuclear power generation. However, the temporary storage of spent fuel is assuming increasing importance as a means of ensuring flexibility in the nuclear fuel cycle. Our investigations of various methods of storage have shown that casks are the most suitable means of storing small quantities of spent fuel of around 500 t, and research and development are in progress to establish dry storage technology for such casks. The soundness of fuel cladding is being investigated. The most important factor in evaluating soundness in storage under inert gas as currently envisaged is creep deformation and rupture, and a number of investigations have been made of the creep behaviour of cladding. The present study was conducted on the basis of existing in-house results in collaboration with Nippon Kakunenryo Kaihatsu KK (Nippon Nuclear Fuel Department Co.), which has hot lab facilities. Tests were run on the creep deformation behaviour of irradiated cladding, and the maximum allowable temperature during dry storage was investigated. (author)

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

    International Nuclear Information System (INIS)

    Cottone, G.

    1978-01-01

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

  7. Experimental study on coolability of particulate core-metal debris bed with oxidization, (2). Fragmentation and enhanced heat transfer in zircaloy debris bed

    International Nuclear Information System (INIS)

    Su, Guanghui; Sugiyama, Ken-ichiro; Aoki, Hiroomi; Kimura, Iichi

    2006-01-01

    The oxidization and coolability characteristics of the particulate Zircaloy debris bed, which is deposited under the hard debris and through which first vapor penetrates and then water penetrates, are studied in the present paper. In the vapor penetration experiments, it is found that Zircaloy debris particles are effectively broken into small pieces after making thick oxidized layer with deep clacks by rapid oxidization under the condition that vapor with 20 cm/s penetrates for 30 to 70 min at an initial debris bed temperature of 1,030degC. It is also confirmed in the water penetration experiments that the oxidized particle debris bed has potentially of high coolability when water penetrates through the fully oxidized particle bed because of a high capillary force originating from those particles with deep cracks on their surfaces. Based on the present study, a new scenario for the appearance and disappearance of the hot spot in the TMI-2 accident is possible. The particulate core-metal core-metal debris bed is first heated up by rapid oxidization with heat generation when vapor can penetrate through the debris bed with porosities. This corresponds to the appearance of the hot spot. The resultant oxidized particulate debris bed causes a high coolability due to its high capillary force when the water can touch the debris bed at wet condition. This corresponds to the disappearance of the hot spot. (author)

  8. Development of ODS (oxide dispersion strengthened) ferritic-martensitic steels for fast reactor fuel cladding

    International Nuclear Information System (INIS)

    Ukai, Shigeharu

    2000-01-01

    In order to attain higher burnup and higher coolant outlet temperature in fast reactor, oxide dispersion strengthened (ODS) ferritic-martensitic steels were developed as a long life fuel cladding. The improvement in formability and ductility, which are indispensable in the cold-rolling method for manufacturing cladding tube, were achieved by controlling the microstructure using techniques such as recrystallization heat-treatment and α to γ phase transformation. The ODS ferritic-martensitic cladding tubes manufactured using these techniques have the highest internal creep rupture strength in the world as ferritic stainless steels. Strength level approaches adequate value at 700degC, which meets the requirement for commercial fast reactors. (author)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  10. Control chart analysis of data regarding 0.2% yield strength (YS) and percent total circumferential elongation (%TCE) for zircaloy clad tubes for PHWR and BWR fuels

    International Nuclear Information System (INIS)

    Yadav, M.B.; Singh, Hari; Vaidyanathan, S.; Sood, D.D.; Raghavan, S.V.; Bandyopadhyay, A.K.; Kulkarni, P.G.

    1992-01-01

    Zircaloy cladding tubes for PHWR and BWR fuels are manufactured and tested at Nuclear Fuel Complex (NFC), Hyderabad. Atomic Fuels Division is carrying out the quality assurance of the fuels on behalf of Nuclear Power Corporation (NPC). In this paper an attempt has been made to assess whether the quality of the clad tubes has met the requirements specified for the two mechanical properties of the tubes namely 0.2% yield strength and percent total circumferential elongation using control chart technique. For this purpose data for about 100 lots in each case were used. Process means and process standard deviations for these properties and the control limits for the corresponding control charts were estimated. The main findings are: (i) In case of PHWR tubes the production quality level with respect to 0.2% YS is higher, while that in case of %TCE is lower causing rejection of lots. On the other hand in the case of BWR tubes the production quality levels with respect to both the properties are higher than the required one. (ii) With respect to 0.2% YS, in case of BWR tubes a change in the pattern of distribution is detected beyond the lot serial no.47. However in case of PHWR tubes, though the data falls into two groups, no such pattern is seen. A modification in the acceptance/rejection criterion of the lot has been suggested. It is also pointed out that to have a correct picture of the total variation it is necessary to study the within tube variation. (author). 4 figs, 2 tabs

  11. 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.

  12. 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.

  13. Embrittlement of pre-hydrided Zircaloy-4 by steam oxidation under simulated LOCA transients

    Energy Technology Data Exchange (ETDEWEB)

    Desquines, J., E-mail: jean.desquines@irsn.fr; Drouan, D.; Guilbert, S.; Lacote, P.

    2016-02-15

    During a Loss Of Coolant Accident (LOCA), the mechanical behavior of high temperature steam oxidized fuel rods is an important issue. In this study, as-received and pre-hydrided axial tensile samples were steam oxidized in a vertical furnace and water quenched in order to simulate a LOCA transient. The samples were then subjected to a mechanical test to determine the failure conditions. Two different rupture modes were evidenced; the first one associated to linear elastic fracture mechanics and the second one is associated to sample failure without applied load. The oxidized cladding fracture toughness was determined relying on intensive metallographic analysis. The sample failure conditions were then back predicted confirming that the main rupture parameters are well captured.

  14. Hydrogen generation monitoring and mass gain analysis during the steam oxidation for Zircaloy using hydrogen and oxygen sensors

    International Nuclear Information System (INIS)

    Fukumoto, Michihisa; Hara, Motoi; Kaneko, Hiroyuki; Sakuraba, Takuya

    2015-01-01

    The oxidation behavior of Zircaloy-4 at high temperatures in a flowing Ar-H_2O (saturated at 323 K) mixed gas was investigated using hydrogen and oxygen sensors installed at a gas outlet, and the utility of the gas sensing methods by using both sensors was examined. The generated amount of hydrogen was determined from the hydrogen partial pressure continuously measured by the hydrogen sensor, and the resultant calculated oxygen amount that reacted with the specimen was in close agreement with the mass gain gravimetrically measured after the experiment. This result demonstrated that the hydrogen partial pressure measurement using a hydrogen sensor is an effective method for examining the steam oxidation of this metal as well as monitoring the hydrogen evolution. The advantage of this method is that the oxidation rate of the metal at any time as a differential quantity is able to be obtained, compared to the oxygen amount gravimetrically measured as an integral quantity. When the temperature was periodically changed in the range of 1173 K to 1523 K, highly accurate measurements could be carried out using this gas monitoring method, although reasonable measurements were not gravimetrically performed due to the fluctuating thermo-buoyancy during the experiment. A change of the oxidation rate was clearly detected at a monoclinic tetragonal transition temperature of ZrO_2. From the calculation of the water vapor partial pressure during the thermal equilibrium condition using the hydrogen and oxygen partial pressures, it became clear that a thermal equilibrium state is maintained when the isothermal condition is maintained, but is not when the temperature increases or decreases with time. Based on these results, it was demonstrated that the gas monitoring system using hydrogen and oxygen sensors is very useful for investigating the oxidation process of the Zircaloy in steam. (author)

  15. 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

  16. 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.)

  17. A study of the accelerated zircaloy-4 oxidation reaction with H2O/H2 mixture gas

    International Nuclear Information System (INIS)

    Kim, Y. S.; Cho, I. J.

    2001-01-01

    A study of the Zircaloy-4 reaction with H 2 O/H 2 mixture gas is carried out by using TGA (Thermo Gravimetric Apparatus) to estimate the hydrogen embrittlement which can possibly cause catastrophic nuclear fuel rod failure. Reaction rates are measured as a function of H 2 /H 2 O. In the experiments reaction temperature is set at 500 .deg. C and total pressure of the mixture gas is maintained at 1 atm. Experimental results reveal that hydriding and oxidation reaction are competing. In early stage, hydriding kinetics is faster than oxidation, however, oxidant in H 2 O forms oxide on the surface as steam environment is maintained, thus, this growing oxide begins to protect the zirconium base metal against hydrogen permeation. In this second stage, the total kinetic rate follows enhanced oxidation kinetics. In the final stage, it is observed that the oxide is broken down and massive hydriding takes place through the mechanical defects in the oxide, whose kinetics is similar to pure hydriding kinetics. These results are confirmed by SEM and EDX analysis along with hydrogen concentration measurements

  18. Modelling of zircaloy-4 corrosion in nitrogen and oxygen mixtures at high temperature

    International Nuclear Information System (INIS)

    Lasserre, M.; Peres, V.; Pijolat, M.; Coindreau, O.; Duriez, C.; Mardon, J.P.

    2015-01-01

    Previous studies of zircaloy-4 corrosion in air have shown accelerated corrosion in the 600-1000 Celsius degrees temperature range with Zr nitrides precipitating near the metal/oxide surface. The aim of this series of slides is to assess the influence of N 2 and O 2 partial pressures on the kinetic rate of growth of a new phase and to propose a kinetic modelling of zircaloy-4 corrosion

  19. Creep modeling of textured zircaloy under biaxial stressing

    International Nuclear Information System (INIS)

    Adams, B.L.; Murty, K.L.

    1984-01-01

    Anisotropic biaxial creep behavior of textured Zircaloy tubing was modeled using a crystal-plastic uniform strain-rate upper-bound and a uniform stress lower-bound approach. Power-law steady-state creep is considered to occur on each crystallite glide system by fixing the slip rate to be proportional to the resolved shear stress raised to a power. Prismatic, basal, and pyramidal slip modes were considered. The crystallographic texture is characterized using the orientation distribution function determined from a set of three pole-figures. This method is contrasted with a Von-Mises-Hill phenomenological model in comparison with experimental data obtained at 673 deg K. The resulting creep-dissipative loci show the importance of the basal slip mode on creep in heavily cold-worked cladding, whereas prismatic slip is more important for the recrystallized materials. (author)

  20. On-line ultrasonic inside-diameter control system for Zircaloy

    International Nuclear Information System (INIS)

    Tanaka, Y.; Fujii, N.; Komatsu, M.; Kubota, H.

    1984-01-01

    An ultrasonic inside-diameter (ID) control system was used during the final etching process for producing Zircaloy nuclear fuel cladding tubes. This results in establishing automatic inside-diameter control during etching with an automatic etching system. In this system, the inside-diameter at the center point in the length of each tube is continuously measured with the ultrasonic inside-diameter measuring equipment during the etching process and the etching is automatically stopped by a signal from the control equipment when the inside-diameter reaches the target value. This made the final etching process economical and suitable for large-scale production, having an equal or better level at the inside-diameter of tubes etched with this system than those made by a process controlled by an air-micrometer

  1. Elucidating the iodine stress corrosion cracking (SCC) process for zircaloy tubing

    International Nuclear Information System (INIS)

    Nagai, M.; Shimada, S.; Nishimura, S.; Amano, K.

    1984-01-01

    Several experimental investigations were made to enhance understanding of the iodine stress corrosion cracking (SCC) process for Zircaloy: (1) oxide penetration process, (2) crack initiation process, and (3) crack propagation process. Concerning the effect of the oxide layer produced by conventional steam-autoclaving, no significant difference was found between results for autoclaved and as-pickled samples. Tests with 15 species of metal iodides revealed that only those metal iodides which react thermodynamically with zirconium to produce zirconium tetraiodide (ZrI 4 ) caused SCC of Zircaloy. Detailed SEM examinations were made on the SCC fracture surface of irradiated specimens. The crack propagation rate was expressed with a da/dt=C Ksup(n) type equation by combining results of tests and calculations with a finite element method. (author)

  2. Oxidation-induced embrittlement and structural changes of Zircaloy-4 tubing in steam at 700-1000 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Ali, A E; Huessein, A G; El-Sayed, A A; El Banna, O A [Atomic Energy Authority, Cairo (Egypt); El Raghy, S M [Cairo Univ. (Egypt). Faculty of Engineering

    1997-02-01

    The oxidation-induced embrittlement and structural changes of Zircaloy-4 (KWU-Type) tubing was investigated under light water reactors (LWR) Loss-of-Coolant. Accident conditions (LOCA) in temperature range 700-1000 deg. C. The effect of hydrogen addition to steam was also investigated in the temperature range 800-1000 deg. C. The oxidation-induced embrittlement was found to be a function of both temperature and time. Fractography investigation of oxidized tubing showed a typical brittle fracture in the stabilized-alpha zone. The microhardness measurements revealed that the alpha-Zr is harder than that near the mid-wall position. The oxidation-induced embrittlement at 900 deg. C was found to be higher than at 1000 deg. C. The results also indicated that the addition of 5% by volume hydrogen to steam resulted in an increase in the degree of embrittlement. (author). 22 refs, 9 figs, 3 tabs.

  3. Control of degradation of spent LWR [light-water reactor] fuel during dry storage in an inert atmosphere

    International Nuclear Information System (INIS)

    Cunningham, M.E.; Simonen, E.P.; Allemann, R.T.; Levy, I.S.; Hazelton, R.F.

    1987-10-01

    Dry storage of Zircaloy-clad spent fuel in inert gas (referred to as inerted dry storage or IDS) is being developed as an alternative to water pool storage of spent fuel. The objectives of the activities described in this report are to identify potential Zircaloy degradation mechanisms and evaluate their applicability to cladding breach during IDS, develop models of the dominant Zircaloy degradation mechanisms, and recommend cladding temperature limits during IDS to control Zircaloy degradation. The principal potential Zircaloy cladding breach mechanisms during IDS have been identified as creep rupture, stress corrosion cracking (SCC), and delayed hydride cracking (DHC). Creep rupture is concluded to be the primary cladding breach mechanism during IDS. Deformation and fracture maps based on creep rupture were developed for Zircaloy. These maps were then used as the basis for developing spent fuel cladding temperature limits that would prevent cladding breach during a 40-year IDS period. The probability of cladding breach for spent fuel stored at the temperature limit is less than 0.5% per spent fuel rod. 52 refs., 7 figs., 1 tab

  4. Superior corrosion resistance properties of TiN-based coatings on Zircaloy tubes in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Khatkhatay, Fauzia [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Jiao, Liang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jian, Jie [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Zhang, Wenrui [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jiao, Zhijie [Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109-2104 (United States); Gan, Jian; Zhang, Hongbin [Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Zhang, Xinghang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Department of Mechanical Engineering, Texas A and M University, College Station, TX 77843-3123 (United States); Wang, Haiyan, E-mail: wangh@ece.tamu.edu [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States)

    2014-08-01

    Thin films of TiN and Ti{sub 0.35}Al{sub 0.65}N nanocomposite were deposited on polished Zircaloy-4 tubes. After exposure to supercritical water for 48 h, the coated tubes are remarkably intact, while the bare uncoated tube shows severe oxidation and breakaway corrosion. X-ray diffraction patterns, secondary electron images, backscattered electron images, and energy dispersive X-ray spectroscopy data from the tube surfaces and cross-sections show that a protective oxide, formed on the film surface, effectively prevents further oxidation and corrosion to the Zircaloy-4 tubes. This result demonstrates the effectiveness of thin film ceramics as protective coatings under extreme environments.

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

    International Nuclear Information System (INIS)

    Bruet, M.; Stora, J.P.

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Pool boiling CHF enhancement by micro/nanoscale modification of zircaloy-4 surface

    International Nuclear Information System (INIS)

    Ahn, Ho Seon; Lee, Chan; Kim, Hyungdae; Jo, HangJin; Kang, SoonHo; Kim, Joonwon; Shin, Jeongseob; Kim, Moo Hwan

    2010-01-01

    Consideration of the critical heat flux (CHF) requires difficult compromises between economy and safety in many types of thermal systems, including nuclear power plants. Much research has been directed towards enhancing the CHF, and many recent studies have revealed that the significant CHF enhancement in nanofluids is due to surface deposition of nanoparticles. The surface deposition of nanoparticles influenced various surface characteristics. This fact indicated that the surface wettability is a key parameter for CHF enhancement and so is the surface morphology. In this study, surface wettability of zircaloy-4 used as cladding material of fuel rods in nuclear power plants was modified using surface treatment technique (i.e. anodization). Pool boiling experiments of distilled water on the prepared surfaces was conducted at atmospheric and saturated conditions to examine effects of the surface modification on CHF. The experimental results showed that CHF of zircaloy-4 can be significantly enhanced by the improvement in surface wettability using the surface modification, but only the wettability effect cannot explain the CHF increase on the treated zircaloy-4 surfaces completely. It was found that below a critical value of contact angle (10 o ), micro/nanostructures created by the surface treatment increased spreadability of liquid on the surface, which could lead to further increase in CHF even beyond the prediction caused only by the wettability improvement. These micro/nanostructures with multiscale on heated surface induced more significant CHF enhancement than it based on the wettability effect, due to liquid spreadability.

  8. Rethinking the Zircaloy Embrittlement Criteria and Its Impact on Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Lee, You Ho; Kim, Bo Kyung; No, Hee Cheon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    These fuel rod failure modes include integral thermal shock fracture, and impact tests. It is quite remarkable to see that the proposed Zircaloy embrittlemt criteria attained from ring compression tests, in general, successfully assure structural integrity of fuel rods subject to relevant failure modes in accidents. This fact demonstrates that ductility of Zircaloy is the key metric to structural integrity of fuel rods. However, the Zircaloy embrittlement criteria set in 1970s inevitably pose limitations that have become increasingly important for today's nuclear fuel and reactor operations. In particular, the criteria do not take into account the steady-state hydrogen embrittlement with burnup. This may be understandable considering the markedly lower discharge burnup in 1970s compared to that of today. The revision of the rule has been already conducted by the U.S NRC to account for high burnup effects on ECR while the temperature limit remains unchanged. The newly proposed rule of the U.S NRC stick to the similar ring compression tests conducted in the early 1970s. In the monumental experimental investigation of Hobson and Rittenhouse in 1972 and 1973, the experimental evidence for the current 1204oC was first addressed. The study found a reasonably accurate correlation between zero ductility temperature and the sum of alpha and oxide layer thickness for the specimens oxidized below 2200oF (1204 .deg. C). However, in spite of the similar oxidation degree, specimens oxidized at 2400 .deg. F (1315 deg. C) were markedly more brittle than specimens oxidized at 2200 .deg. F (1204 .deg. C). The study explained this by the increase in solid-solution hardening due to a higher oxygen solubility at a higher temperature. Such a nice experimental correlation attained between the nil ductility temperature and the remaining beta layer thickness fraction below 1204 .deg. C has become a critical basis for the current temperature limit; at 1315 .deg. C- thecorrelation

  9. Ductility and failure behaviour of both unirradiated and irradiated zircaloy-4 cladding using plane strain tensile specimens

    International Nuclear Information System (INIS)

    Carassou, S.; Le Saux, M.; Pizzanelli, J.P.; Rabouille, O.; Averty, X.; Poussard, C.; Cazalis, B.; Desquines, J.; Bernaudat, C.

    2010-01-01

    In this work, eight PST (Plan Strain Tensile) tests machined from a Zircaloy-4 (Zy-4) cladding irradiated up to 5 annual cycles have been performed at 280, 350 and 480 Celsius degrees. The specimen displacements during the tests were filmed and digitally recorded to allow the use of a Digital Image Correlation (DIC) analysis technique to experimentally determine the local strains on the outer surface of the specimens. The plane strain conditions have been verified and prevail over a wide area between the notches of the specimen, as expected from full 3D FE numerical analysis performed in support of the tests. For the first time, the location of the onset of fracture for this geometry on irradiated material has been experimentally observed: at 280 C.degrees, crack initiates in the vicinity of the notches, in an area where plane strain conditions are not fulfilled, and for a local circumferential strain value of about 5%. At 350 C. degrees and 480 C. degrees, cracks initiate at a location where plane strain conditions prevail, for circumferential strain values respectively close to 10% and greater than 50%. These results have been compared to results obtained previously by similar test on fresh and hydrided material, as well as tests performed as support to the study. At 350 C. degrees, the homogeneous 700 ppm hydrided Zy-4 and the Zy-4 irradiated during 5 annual cycles exhibit similar fracture behaviour, for both fracture hoop strain values (10%) and fracture mode (through-wall slant fracture). For the irradiated material, it has clearly been established that at 350 C. degrees, a brittle fracture occurs at the outer surface in the hydride rim. The crack propagates subsequently toward the inner surface and the notches, where final fracture occurs

  10. Irradiation and lithium presence influence on the crystallographic nature of zirconia in the framework of PWR zircaloy 4 fuel cladding corrosion study

    International Nuclear Information System (INIS)

    Gibert, C.

    1999-01-01

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

  11. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Science.gov (United States)

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

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  12. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2008-08-31

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  13. High temperature oxidation test of oxide dispersion strengthened (ODS) steel claddings

    International Nuclear Information System (INIS)

    Narita, Takeshi; Ukai, Shigeharu; Kaito, Takeji; Ohtsuka, Satoshi; Matsuda, Yasushi

    2006-07-01

    In a feasibility study of ODS steel cladding, its high temperature oxidation resistance was evaluated. Although addition of Cr is effective for preventing high temperature oxidation, excessively higher amount of Cr leads to embrittlement due to the Cr-rich α' precipitate formation. In the ODS steel developed by the Japan Atomic Energy Agency (JAEA), the Cr content is controlled in 9Cr-ODS martensite and 12Cr-ODS ferrite. In this study, high temperature oxidation test was conducted for ODS steels, and their results were compared with that of conventional austenitic stainless steel and ferritic-martensitic stainless steel. Following results were obtained in this study. (1) 9Cr-ODS martensitic and 12Cr-ODS ferritic steel have superior high temperature oxidation resistance compared to 11mass%Cr PNC-FMS and even 17mass% SUS430 and equivalent to austenitic PNC316. (2) The superior oxidation resistance of ODS steel was attributed to earlier formation of the protective alpha-Cr 2 O 3 layer at the matrix and inner oxide scale interface. The grain size of ODS steel is finer than that of PNC-FMS, so the superior oxidation resistance of ODS steel can be attributed to the enhanced Cr-supplying rate throughout the accelerated grain boundary diffusion. Finely dispersed Y 2 O 3 oxide particles in the ODS steel matrix may also stabilized the adherence between the protective alpha-Cr 2 O 3 layer and the matrix. (author)

  14. ''C-ring'' stress corrosion cracking scoping experiment for Zircaloy spent fuel cladding

    International Nuclear Information System (INIS)

    Smith, H.D.

    1986-03-01

    This document describes the purpose and execution of the stress corrosion cracking scoping experiment using ''C-ring'' cladding specimens. The design and operation of the ''C-ring'' stressing apparatus is described and discussed. The experimental procedures and post-experiment sample evaluation are described

  15. 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

  16. 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.)

  17. Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Y.; Qian, S.; Littrell, K.; Parish, C.M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Plummer, L.K. [University of Oregon, Eugene, OR 97403 (United States)

    2015-05-15

    A nondestructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless-steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Our study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount (≈20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor can be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for determining absolute hydrogen concentrations.

  18. Analysis of zircaloy oxide thickness data from PWRs

    International Nuclear Information System (INIS)

    Sheppard, K.D.; Speyer, D.M.; Chan, Y.Y.; Frankl, I.; Strasser, A.A.

    1990-02-01

    Prior EPRI funded research (Project 1250-1) resulted in a set of Zircaloy waterside corrosion models. These models were based principally on KWU reactor data. The objective of this study was to evaluate the ability of the KWU corrosion models to predict available domestic USA data for all domestic PWR vendors in order to further validate the models and to provide a consistent basis to judge the corrosion data of the domestic plants. A methodology for analyzing the large amount of data was developed and implemented in a single channel model. This model includes the capability, by a method described herein, of accounting for open core related effects (crossflow) and the effect of the immediately adjacent fuel rods, guide tubes, etc., on the coolant conditions around the fuel rods that were measured for oxide thickness. Data from the Arkansas Unit number-sign 2 (ANO-2) Combustion Engineering (C-E), Oconee Units 1 and 2 built by Babcock ampersand Wilcox (B ampersand W), and the Trojan reactor built by Westinghouse (W) were used in this study. The corrosion models previously developed, and the present single channel model methodology, were able to predict the corrosion data quite well. The maximum corrosion thickness was on the order of 20 to 40 microns in all plants studied. 13 refs., 11 figs., 5 tabs

  19. Modelling of oxidation and hydriding behaviour of Zircaloy-2 pressure tubes in PHWR

    International Nuclear Information System (INIS)

    Sah, D.N.; Sunil Kumar; Khan, K.B.

    2002-01-01

    A computer model named DOCTOR (Deuteriding of Coolant Tubes during Operation of Reactor) has been developed for predicting the axial profile of oxide thickness and hydrogen (Deuterium) concentration in PHWR pressure tubes. This model is applicable to single channel or full core analysis. The main source of hydrogen is considered to be oxidation of pressure tube on the i.d. surface by high temperature coolant water. Three stages of oxidation is considered namely, pre- transition, post transition and accelerated. Oxidation rate is considered to be dependent on channel power, axial power/flux distribution, coolant temperature and pre-existing oxide thickness at the location. The kinetics parameters for oxidation model are derived from the actual measurement of oxide thickness on a number of pressure tubes examined in PIE Division. The input data required for the model are: channel power, channel power factor, axial flux distribution, coolant inlet temperature, critical oxide thickness, hydrogen pick up fraction, initial hydrogen in the material and time of operation (efpy). The model calculates the oxide layer thickness on the inside surface of the pressure tube along the length. The amount of hydrogen picked up by the pressure tube is calculated from the oxide thickness using hydrogen pick up fraction determined from the PIE data. The pressure tube length is divided into a number of axial segments for calculation. The temperature and fast neutron flux assumed to be constant in a given segment. The axial temperature profile calculated from the axial power profile in the channel is used for calculating the oxidation rate at various locations in the pressure tube. The model has been validated with PIE data of hydrogen equivalent measurement on a number of irradiated Zircaloy-2 pressure tubes of various PHWRs. The performance of the model in predicting the axial profile of hydrogen in the pressure tubes has been found to be good. (author)

  20. Spectrochemical determination of impurities in zircaloy 2 and 4

    International Nuclear Information System (INIS)

    Paula Reino, L.C. de; Lordello, A.R.

    1987-06-01

    A method has been developed for the determination of Hf,Co,Mo,Pb,Ti,V,Al,Si,W,Cu,Mg,Mn,B and Cd in zircaloy 2 and 4. For hafnium determination 10% CuF 2 is added as spectrographic buffer on a previously oxidized zircaloy; the samples are loaded in a shallow cup electrode of Scribner Mullins type and excited in a direct current arc. The carrier distillation technique has been used for the other elements. Better results were obtained with 25% AgCl as carrier. The precision of the method varies from 4% for copper to 29% for boron but it does not exceed 17% for most elements. (Author) [pt

  1. Corrosion characteristics and oxide microstructures of Zircaloy-4 in aqueous alkali hydroxide solutions

    International Nuclear Information System (INIS)

    Jeong, Y.H.; Baek, J.H.; Kim, S.J.; Kim, H.G.

    1999-01-01

    The corrosion characteristics of Zircaloy-4 have been investigated in various aqueous solutions of LiOH, NaOH, KOH, RbOH and CsOH with equimolar M + and OH - at 350 C. The characterization of the oxides was performed using transmission electron microscope (TEM) and scanning electron microscope (SEM) on the samples which were prepared to have an equal oxide thickness in pre-transition and post-transition regimes. At a low concentration (4.3 mmol) of aqueous alkali hydroxide solutions, the corrosion rates decrease gradually as the ionic radius of cation increases. At a high concentration (32.5 mmol), the corrosion rate increases significantly in LiOH solution and slightly in NaOH solution, but in the other hydroxide solutions such as KOH, RbOH and CsOH, the corrosion rate is not accelerated. Even if the specimens have an equal oxide thickness in LiOH, NaOH and KOH solutions, the oxide microstructure formed in the LiOH solution is quite different from those formed in the NaOH or the KOH solutions. In the LiOH solution, the oxides grown in the pre-transition regime as well as in the post-transition regime have an equiaxed structure including many pores and open grain boundaries. The oxides grown in the NaOH solution have a protective columnar structure in the pre-transition regime but an equiaxed structure in the post-transition regime. On the other hand, in the KOH solution, the columnar structure is maintained from its pre-transition regime to the post-transition regime. On the basis of the above results, it can be suggested that the cation incorporation into zirconium oxide would control the oxide microstructure, the oxide growth mechanism at the metal-oxide interface and the corrosion rate in alkali hydroxide solutions. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun, E-mail: pdj@kaeri.re.kr; Kim, Hyun Gil; Jung, Yang Il; Park, Jung Hwan; Yang, Jae Ho; Koo, Yang Hyun

    2016-12-15

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

  3. Corrosion in nuclear systems. 4. Comparison of the PWR Cladding Corrosion Models for Test IFA-638.1-3

    International Nuclear Information System (INIS)

    Kim, Yong-Deog; Bae, Seong-Man; Lee, Chang-Sup

    2001-01-01

    The cladding corrosion test (IFA-638) is being performed to investigate the corrosion properties of different modern PWR cladding materials. The experimental results are evaluated by the corrosion models EPRI/KWU/CE, ESCORE, NEPLC, and COCHISE. When comparing the measured and the predicted oxide thickness, the following conclusions can be drawn: 1. Considering the fresh material parts (lower parts) of each rod, the oxide thickness calculations of all models under-predicted the measured values by up to 50% after 118 days of exposure. The NEPLC model, however, showed good agreement for 263 days of exposure, while the COCHISE and EPRI/KWU/CE-ESCORE models over-predicted (about +50%) and under-predicted (about -42%), respectively. 2. The oxide layer thickness on the pre-irradiated parts (upper parts) of each rod is well predicted by the COCHISE model after 118 days of exposure, but the other models over-predicted the thickness. All the models over-predicted the oxide thickness after 263 days of exposure, and the divergency between the measured and calculated oxide thickness became larger. 3. The differences in the calculated oxide thickness between the models at low burnup (fresh parts) are attributed to the different transition point determinations of the models. 4. Comparing the measurements with the calculations from the pre-irradiated parts of each rod, the overall over-prediction could be accounted for by the fact that the post-transition regime of all four models was calibrated for standard Zircaloy-4 materials. The differences between the models were attributed to empirical variables such as the frequency factor (k 2 , B) and the activation energy (Q 2 ) in Tables I, II, and III, which were calibrated with other experimental/plant data. (authors)

  4. Observations on deformation systems in zircaloy-2 deformed at room temperature

    International Nuclear Information System (INIS)

    Pettersson, K.; Bergqvist, H.

    1975-08-01

    Different polycrystalline samples of Zircaloy-2 with textures such that the c-axis of most of the grains are oriented near the sheet normal were subjected to loading conditions such that sheet thinning was accomplished. Metallography showed that no twinning was involved. Electron microscopy showed the presence of dislocations which were usually confined to deformation bands. With the help of stereo micrographs the most likely plane of slip was determined to be (1011). The possibility of slip as a means of breaking the oxide film in iodine induced stress corrosion cracking of Zircaloy-2 is briefly discussed. (author)

  5. Microstructure and crystallographic texture evolution during TIG welding of zircaloy-2 material

    International Nuclear Information System (INIS)

    Jha, S.K.; Singh, R.P.; Singh, V.K.; Ramanathan, R.; Samjdar, I.; Srivastava, D.; Tewari, R.; Dey, G.K.

    2005-01-01

    Zirconium and its alloys are extensively used as structural materials in nuclear reactors, because of better neutron economy, good corrosion resistance in water and good mechanical properties at operating temperature. Zircaloy-2 and zircaloy-4 are widely used in both pressurized water reactors (PWR) and boiling water reactors (BWR) as fuel cladding materials and as calandria tube and pressure tube materials in pressurized heavy water reactors (PHWR). The satisfactory performance and the life of the reactor components depend mainly upon their mechanical properties, corrosion properties and dimensional stability in the reactor condition, which are strong function of metallurgical parameters such as microstructure and texture. Therefore, for best performance of the reactor components these parameters are optimized during their fabrication. The microstructure and texture of the zircaloy-2 components are expected to get modified during the welding of the components. In this study the evolution of the microstructure and texture has been investigated as a function of the welding parameters. Heat input was varied the current and welding time. A variety of analytical techniques have been applied for the study on microstructure and texture of the welds. Optical microscopy and electron microscopy were used to evaluate the detailed microstructure. X-ray diffraction (XRD) was used investigate the crystallographic textures among the base metal, heat affected zone and fusion zone. Particular attention was focused on the determination of microtexture in weld by using electron backscatter diffraction (EBSD) technique. After that, an effort was put to compare the results of X-ray macro-texture and EBS-microtexture. (author)

  6. Demonstration of fuel resistant to pellet-cladding interaction. Second semiannual report, January--June 1978

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1978-09-01

    This program has as its ultimate objective the demonstration of an advanced fuel concept that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Since currently used fuel in the nuclear power industry is subject to the PCI failure mechanism, reactor operators limit the rates of power increases and thus reduce their capacity factors in order to protect the fuel. Two concepts are being developed for possible demonstration within this program: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as ''barrier fuels'') have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress and reactive fission products during reactor service. The demonstration of one of these concepts in a commercial power reactor is planned for PHASE 2 of this program. The current plans for the demonstration will involve approximately 132 bundles of PCI-resistant fuel

  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. 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

  9. Simulation of the chemical state of irradiated oxide fuel; influence of the internal corrosion on the mechanical properties of Zry-4 tubing

    International Nuclear Information System (INIS)

    Hofmann, P.

    1979-03-01

    Zircaloy is not compatible with oxide fuel nor with some fission product elements. Therefore, chemical interaction between the irradiated oxide fuel and the Zry cladding material take place, especially at temperatures that can be reached during reactor incidents (ATWS, LOCA). In order to find out which influence the chemical interaction between the fission products and the Zry cladding material have on the mechanical properties of Zry-4 tubing out-of-pile burst experiments and creep rupture tests have been performed at temperatures >=600 0 C with short tube specimens containing simulated fission products. First of all, assessments of the chemical state of irradiated oxide fuel were performed and a method is described for introducing simulated fission product species into fresh oxide fuel for irradiation tests. As the test results of the out-of-pile studies show, only iodine can lead to a low ductility failure of the Zry-tubing at temperatures >=600 0 C. However, the influence of iodine on the deformation behavior of Zry-tubing can be neglected above 850 0 C. (orig.) [de

  10. 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.)

  11. 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)

  12. CEA fuel pencil qualification under irradiation: from component conception to fuel assembly irradiation in a power reactor

    International Nuclear Information System (INIS)

    Marin, J.-F.; Pillet, Claude; Francois, Bernard; Morize, Pierre; Petitgrand, Sylvie; Atabek, R.-M.; Houdaille, Brigitte.

    1981-06-01

    Fabrication of fuel pins made of uranium oxide pellets and of a zircaloy 4 cladding is described. Irradiation experiment results are given. Thermomechanical behavior of the fuel pin in a power reactor is examined [fr

  13. 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

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

    International Nuclear Information System (INIS)

    Smith, E.

    1979-01-01

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

  15. Finite element modelling of the oxidation kinetics of Zircaloy-4 with a controlled metal-oxide interface and the influence of growth stress

    International Nuclear Information System (INIS)

    Zumpicchiat, Guillaume; Pascal, Serge; Tupin, Marc; Berdin-Méric, Clotilde

    2015-01-01

    Highlights: We developed two finite element models of zirconium-based alloy oxidation using the CEA Cast3M code to simulate the oxidation kinetics of Zircaloy-4: the diffuse interface model and the sharp interface model. We also studied the effect of stresses on the oxidation kinetics. The main results are: • Both models lead to parabolic oxidation kinetics in agreement with the Wagner’s theory. • The modellings enable to calculate the stress distribution in the oxide as well as in the metal. • A strong effect of the hydrostatic stress on the oxidation kinetics has been evidenced. • The stress gradient effect changes the parabolic kinetics into a sub-parabolic law closer to the experimental kinetics because of the stress gradient itself, but also because of the growth stress increase with the oxide thickness. - Abstract: Experimentally, zirconium-based alloys oxidation kinetics is sub-parabolic, by contrast with the Wagner theory which predicts a parabolic kinetics. Two finite element models have been developed to simulate this phenomenon: the diffuse interface model and the sharp interface model. Both simulate parabolic oxidation kinetics. The growth stress effects on oxygen diffusion are studied to try to explain the gap between theory and experience. Taking into account the influence of the hydrostatic stress and its gradient into the oxygen flux expression, sub-parabolic oxidation kinetics have been simulated. The sub-parabolic behaviour of the oxidation kinetics can be explained by a non-uniform compressive stress level into the oxide layer.

  16. Cladding using a 15 kW CO2 laser

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. Sensing characteristics of nanocrystalline bismuth oxide clad-modified fiber optic gas sensor

    Science.gov (United States)

    Manjula, M.; Karthikeyan, B.; Sastikumar, D.

    2017-08-01

    Gas sensing properties of nanocrystalline bismuth oxide clad - modified fiber optic sensor is reported for ammonia, ethanol, methanol and acetone gasses at room temperature. The output of sensor increases or decreases for certain gasses when the concentration of the gas is increased. The sensor exhibits high response and good selectivity to methanol gas. Time response characteristics of the sensor are also reported.

  18. The effect of hot spots upon swelling of Zircaloy cladding as modelled by the code CANSWEL-2

    International Nuclear Information System (INIS)

    Haste, T.J.; Gittus, J.H.

    1980-12-01

    The code CANSWEL-2 models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised-water reactor (PWR). It can treat azimuthal non-uniformities in cladding thickness and temperature, and model the mechanical restraint imposed by the nearest neighbouring rods, including situations where cladding is forced into non-circular shapes. The physical and mechanical models used in the code are presented. Applications of the code are described, both as a stand-alone version and as part of the PWR LOCA code MABEL-2. Comparison with a limited number of relevant out-of-reactor creep strain experiments has generally shown encouraging agreement with the data. (author)

  19. 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)

  20. 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)