bluuebunny/crossref_metadata_2025 huggingface dataset:
Hydrogen Content, Preoxidation, and Cooling Scenario Effects on Post-Quench Microstructure and Mechanical Properties of Zircaloy-4 and M5® Alloys in LOCA Conditions
Authors: Brachet J. -C. , Vandenberghe-Maillot V. , Portier L. , Gilbon D. , Lesbros A. , Waeckel N. , Mardon J. -P.
Published: September 2009

Previous papers pointed out the influence of long-term service exposures on the thermal-mechanical behavior of Zr alloys in LOCA conditions and, especially, the impact of in-service hydrogen pick-up on post-quench mechanical properties. Moreover, the oxide layer grown under in-service conditions was occasionally expected to have a protective effect against high temperature oxidation. Finally, the oxygen and hydrogen distributions within the prior-β layer appear as a key parameter with regard to the residual ductility of the alloy, especially as a function of the cooling scenario. The objective of the study presented here was to further investigate the influence of these parameters on the post-quench mechanical properties. Unirradiated Zircaloy-4 and M5® cladding tubes were consequently hydrided up to different concentration levels, then oxidized at high temperature (1000–1200°C) up to at least 10 % measured equivalent cladding reacted (ECR) and directly quenched to room temperature (RT). Ring compression tests (RCT), 3-point bending tests (3PBT) at RT and 135°C, as well as impact tests at RT were then performed to determine the evolution of the post-quench mechanical properties of Zircaloy-4 and M5® alloys with H content. Similarly, specimens preoxidized out-of-pile were also submitted to high temperature oxidation and direct quench, as well as to post-quench ring compression tests. Along with calculations of oxygen diffusion in the metal, results from those tests allowed us to estimate the assumed protective effect of the pretransient oxide layer. Finally, using specimens in the as-received condition or hydrided to typical end-of-life H contents, the effect of temperature history after oxidation at 1200°C was studied, i.e., at the end of the high temperature isothermal oxidation, samples were either submitted to direct quenching to RT or to slow cooling to different final quenching temperatures. It was thus demonstrated that the cooling scenario has a significant impact on the post-quench mechanical properties. All test samples were investigated by means of fractographic examinations to assess the type of failure mode. Moreover, a deep metallurgical analysis has been performed: SEM and image analysis were used for accurate phase thickness measurements, nuclear and electron microprobes for quantitative mapping of hydrogen and oxygen. It proved that the oxygen and hydrogen contents and their distribution in the prior-β layer have a first-order influence on the residual ductility. From all the results obtained on as-received and hydrided samples directly quenched from the oxidation temperature, it was then possible to derive a relationship between structural parameters, i.e., oxygen and hydrogen contents and thickness of the prior-β layer, and the post-quench impact properties at RT.
In-Pile Criteria for the Initiation of Massive Hydriding of Zr in Steam-Hydrogen Environment
Authors: Evdokimov I. A. , Likhanskii V. V.
Published: September 2009

It is shown that irradiation of oxide film on the surface of zirconium alloy by fission fragments may promote the breakdown of oxide protective properties and lead to massive hydriding of zirconium in a steam-hydrogen environment far beyond the condition of “oxygen starvation.” Massive hydriding of zirconium claddings may cause severe secondary degradation of failed fuel rods in commercial nuclear reactors. The effect of irradiation is concerned with two factors. The first one is radiolysis of steam-hydrogen mixture and chemical interaction of highly reactive radicals with the oxide surface. The second factor is intense generation of radiation-induced defects in crystal lattice of ZrO2 which enhance the rate of oxide reduction. A model is proposed for assessment of the threshold conditions for the onset of massive hydriding depending on composition of steam-hydrogen mixture and irradiation dose rate.
Role of Twinning and Slip in Deformation of a Zr-2.5Nb Tube
Authors: Kim Young S.
Published: September 2009

Tensile tests were conducted from RT to 400°C on tensile specimens taken from the tangential direction (TD) and the longitudinal direction (LD) of a Zr-2.5Nb tube with a strong tangential texture. When tensile stresses were applied to the TD with the tensile axis normal to the basal plane, {1012} twinning was activated. With the tensile stresses applied to the LD, however, ‹a› slip only was activated. The activation of twinning and slip in the TD and the LD, respectively, caused the tensile properties of the Zr-2.5Nb tube to become anisotropic, leading to higher yield and tensile strengths, and lower uniform and tensile elongations in the TD when compared to those in the LD. Furthermore, twinning promoted softening soon after yielding only in the TD, and localized deformation, while the ‹a› slip on the prism plane caused a larger strain hardening after yielding. With increasing temperature, the primary deformation mode changed from twinning to ‹c+a› slip in the TD and ‹a› slip to ‹c+a› slip in the LD, resulting in negative temperature dependences for the extent of strain hardening and tensile elongation of the Zr-2.5Nb tube in both directions. The yield stress plateau is discussed in view of the activation of twinning or ‹c+a› slip with a higher critical resolved shear stress that is strongly temperature dependent.
Measurement of Rates of Delayed Hydride Cracking (DHC) in Zr-2.5Nb Alloys—An IAEA Coordinated Research Project
Authors: Coleman Christopher E. , Inozemtsev Victor V.
Published: September 2009

Values of DHC rates in Zr alloys are sensitive to measurement procedures. A standard method has been developed at the laboratories of AECL and evaluated in a round-robin involving ten IAEA member states. Two test materials were used—Zr-2.5Nb pressure tubes in the cold-worked (CANDU™) and heat-treated (RBMK) conditions. Cracks were grown from fatigued starter cracks in the axial direction on the axial-radial plane of the original tubes. To obtain the maximum value of Vc, specimens were heated to dissolve all their hydrogen, then cooled at 1 to 3°C/min to the test temperature before loading at 15 MPa√m. Although the start of cracking was detected by potential drop, the extent of cracking was measured directly on the crack faces. The values of incubation time to the start of cracking were highly variable but Vc was well behaved. The values of Vc were normally distributed with a range varying from a factor of 1.2 to 5.2. At 250°C the mean value of Vc from 80 specimens of cold-worked material was 8.9(±1.12)×10−8 m/s and from 41 specimens of heat-treated material the mean value of Vc was3.3(±0.64)×10−8 m/s. Tests were also done at six other temperatures between 144 and 283°C, using up to 22 specimens at each temperature. Both materials had an Arrhenius-type temperature dependence, Vc=Aexp(−Q/RT). The use of strictly defined and coordinated experimental procedures gave a consistent set of Vc values, allowing effective comparison of results obtained in different national laboratories and resulting in good correlations between the DHC velocity values and differences in strength, crystallographic texture, and distribution of β-phase in the test materials.
Effects of Pt Surface Coverage on Oxidation of Zr and Other Materials
Authors: Anghel Clara , Hultquist Gunnar , Limbäch Magnus , Szakalos Peter
Published: September 2009

Certain elements, including noble metals, are identified to influence corrosion behavior of many metals in high-temperature water/steam and O2. We have previously reported effects of porous Pt coatings on the thermal oxidation of Zr, Fe, Ni, Cr, and GaAs in O2. Effects of Pt on oxidation of Zircaloy-2 in H2O have also been observed at temperatures near 400°C. An enhanced oxidation rate is observed in all of the studied systems upon a sufficiently high surface Pt-particle density. Even more interesting, low Pt-particle density in most cases leads to a decreased oxidation rate. In the case of Zr and Zircaloy-2 the beneficial effect of Pt is stronger when hydrogen is present in the metal substrate and in the oxide layer. From 18O/SIMS experiments it is concluded that, for a sufficiently high Pt-particle density on the surface, enhanced transport of dissociated oxygen, On− (n=0;2), towards the oxide/metal interface occurs in all of the studied systems. An oxygen spillover from Pt particles to the adjacent surface is a known phenomenon in catalysis and partly explains the results observed also in the current study of oxidation of Zr-based materials. Such a spillover involves a high surface diffusivity of dissociated oxygen, On−, which results in an increased gradient of On− across the oxide scale. In turn, the high oxygen gradient enhances the transport of oxygen towards the substrate/oxide-interface. This shows that not only diffusivity but also an effective activity of dissociated oxygen at the external oxide surface influences the oxidation rate. Naturally, the effects of Pt are maximized when small (nm-sized) and evenly distributed Pt particles are present on the surface. In our presentation a summary of obtained results of Pt additions on different materials will be given with suggested interpretations. Especially, the observed beneficial combination of Pt and hydrogen in the oxidation of Zr-based materials is discussed with a possible mechanistic explanation. Parallels are also drawn to the potential effect of Noble Metal Chemical Application (NMCA), used in a large fraction of the U.S. BWRs, on the corrosion resistance of Zr-based claddings.
In Situ EIS Measurements of Irradiated Zircaloy-4 Post-Transition Corrosion Kinetic Behavior
Authors: Rishel D. M. , Eklund K. L. , Kammenzind B. F.
Published: September 2009

Under irradiation the post-transition corrosion rates of Zircaloy are enhanced relative to out-of-flux rates. In an effort to better understand the effects of different mixed-field radiation environments and the variables that may contribute to this enhancement a series of in situ electrochemical impedance spectroscopy (EIS) tests were conducted in the Halden Test Reactor under PWR type conditions. The EIS tests were conducted using five identically fabricated electrode cells, which employed Zircaloy-4 tipped working electrodes. Approximately 30 μn thick oxides were prefilmed on the electrode tips prior to test insertion. The electrode test train was designed in a manner that enabled a wide range of gamma and neutron radiation flux conditions to be evaluated. Gamma fluxes (E ≥ MeV) ranged from 1.7×1013 to 1.3×1014 cm−2s−1 while the fast neutron fluxes (E ≥1 MeV) ranged from 2.9×10010 to 3.2×1013 cm−2s−1. In addition, the EIS tests were conducted over temperatures ranging from 250 to 335°C. Electrode sets were situated into two basic radiation environments, which were differentiated by the magnitude and relative proportion of the mixed radiation field. In one environment, where the fast neutron flux was low, the relative gamma radiation component was high (i.e., high gamma-to-neutron ratio environment). In the other environment, where the fast neutron flux was high, the relative gamma radiation component was low (i.e., low gamma-to-neutron ratio environment). Based on the low frequency limits, the impedance moduli and inferred corrosion rates were found to be a complex function of temperature, and the relative proportion of gamma and neutron radiation. At low temperatures of 250°C it was found that the electrode sets which were exposed under high gamma-to-neutron ratios exhibited low frequency impedances that were almost 2.5 times lower than the case for the electrode sets which were placed in low gamma-to-neutron ratio. This occurred even though the absolute magnitude of the gamma and neutron flux fields decreased. This would imply that the corrosion rate is higher under high gamma-to-neutron ratios than under low gamma-to- neutron ratios. Conversely, at high temperatures, 335°C, only very small differences in impedance were noted as a function of radiation environment. Post-test destructive examination of the Zircaloy tipped electrodes indicates that the growth of the oxide formed during the in-pile phase of testing is consistent with the in situ EIS measurements obtained in-pile. The porous nature of the post-transition oxide coupled with the apparent correlation with gamma radiation suggests that heterogeneous radiolysis within the corrosion product film plays an important role in enhancing post-transition corrosion kinetics.
The Effect of Hydrogen on the Transition Behavior of the Corrosion Rate of Zirconium Alloys
Authors: Harada Makoto , Wakamatsu Ryuji
Published: September 2009

Three zirconium alloys (Zry-4, Zry-2, and NDA) were corrosion tested at 633 K in pure water and in mixed gas of Ar and oxygen. In both environments, their corrosion rate showed cyclic changes. The time and the weight gain at the transition point changed with their corrosion resistance. In the pure water corrosion test, one test cycle was 600 h, and the cumulative total test duration was as much as 7200 h. After each test cycle, the weight changes and the hydrogen contents were analyzed. The hydrogen pick-up ratio was calculated for each test cycle based on the weight change and the increase of the hydrogen content. The calculated hydrogen pick-up ratios were not constant. They showed periodic changes. In the pre-transition region, the corrosion rate showed a gradual decrease; however, the calculated hydrogen pick-up rate gradually increased. After the transition, in contrast to the increase in the corrosion rate, the calculated hydrogen pick-up rate decreased. The changes of the corrosion rate and the calculated hydrogen pick-up rate for the test interval exhibited an inverse relationship. Transmission electron microscopy (TEM), scanning TEM, energy dispersive x-ray (EDX) spectroscopy, and electron diffraction (ED) pattern analysis of the oxide/metal interface showed the existence of an interface layer that consists of distorted ZrO2 and a sub-oxide. Before the transition, the interface layer was clearly observed. On the other hand, the interface layer disappeared after the transition. The cyclic changes in the corrosion rate were closely related to the interface barrier layer at the oxide/metal interface. The interface layer suppresses the corrosion rate, and the degradation of its barrier property increases the corrosion rate. By comparing the morphology of the oxide/metal interface between the water corrosion samples and the mixed gas oxidation samples, it can be seen that the generated hydrogen in the oxide may have two opposite effects. One is to stabilize the interface layer and to suppress the transport of the oxygen ion, and the other is to degrade the interface layer at the transition and accelerate the generation of the uniform oxide. The effect of the chemical composition on the properties of the oxide-metal interface and the mechanism of the corrosion transition will be discussed.
Corrosion and Oxide Properties of HANA Alloys
Authors: Park Jeong-Yong , Choi Byung-Kwon , Yoo Seung Jo , Jeong Yong Hwan
Published: September 2009

The corrosion behavior of HANA alloys has been investigated at 360°C in a pure water autoclave as well as in a pressurized water reactor (PWR)-simulating loop. The grain morphology and the incorporated precipitates of the oxide were characterized to elucidate the corrosion mechanism of HANA alloys. HANA alloys showed a much lower corrosion rate when compared to Zircaloy-4 in the investigated corrosion environments with HANA-6 (Zr-1.1Nb-0.05Cu), revealing a lower corrosion rate than HANA-4 (Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr). The corrosion rate was increased with an increase of the final annealing temperature of the alloys and the dissolved oxygen content in the corrosion environment. The oxide grain morphology of the HANA alloys was changed periodically from equiaxed grains to columnar ones during its growth to form a layered structure of the oxide. The change of the oxide grain morphology correlated well with the transition of the corrosion kinetics. It was revealed that the fully developed layer was thicker and the columnar grain size was larger in the protective oxide with a lower corrosion rate. The precipitates of HANA alloys were more finely distributed and oxidized more slowly in the oxide when compared to those of Zircaloy-4. Such oxidation characteristics of the precipitates along with the reduction of the Sn content were considered to be more favorable for growing an oxide to a greater thickness before transition and increasing the columnar grain size of the oxide, thereby improving the corrosion resistance of HANA alloys.
Microstructural Characterization of Oxides Formed on Model Zr Alloys Using Synchrotron Radiation
Authors: Motta A. T. , Gomes da Silva M. J. , Yilmazbayhan A. , Comstock R. J. , Cai Z. , Lai B.
Published: September 2009

To understand how alloy chemistry and microstructure impact corrosion performance, oxide layers formed at different stages of corrosion on various model zirconium alloys (Zr-xFe-yCr, Zr-xCu-yMo, for various x, y) and control materials (pure Zr, Zircaloy-4) were examined to determine their structure and the connection of such structure to corrosion kinetics and oxide stability. Microbeam synchrotron radiation diffraction and fluorescence of oxide cross sections were used to determine the oxide phases present, grain size, and orientation relationships as a function of distance from the oxide-metal interface. The results show a wide variation of corrosion behavior among the alloys, in terms of the pretransition corrosion kinetics and in terms of the oxide susceptibility to breakaway corrosion. The alloys that exhibited protective behavior at 500°C also were protective during 360°C corrosion testing. The Zr-0.4Fe-0.2Cr model ternary alloy showed protective behavior and stable oxide growth throughout the test. The results of the examination of the oxide layers with microbeam X-ray diffraction show clear differences in the structure of protective and nonprotective oxides both at the oxide-metal interface and in the bulk of the oxide layer. The nonprotective oxide interfaces show a smooth transition from metal to oxide with metal diffraction peaks disappearing as the monoclinic oxide peaks appear. In contrast, the protective oxides showed a complex structure near the oxide-metal interface, showing peaks from Zr3O suboxide and a highly oriented tetragonal oxide phase with specific orientation relationships with the monoclinic oxide and the base metal. The same interfacial structures are observed through their diffraction signals in protective oxide layers formed during both 360°C and 500°C corrosion testing. These diffraction peaks showed much higher intensities in the samples from 500°C testing. The results for the various model alloys are discussed to help elucidate the role of individual alloying elements in oxide formation and the influence of oxide microstructure on the corrosion mechanism.
Manufacturing Variability, Microstructure, and Deformation of Zr-2.5Nb Pressure Tubes
Authors: Bickel G. A. , Griffiths M.
Published: September 2009

The time-dependent deformation behavior of pressure tubes in CANDU™ (CANada Deuterium Uranium) reactors is an important property that has to be predicted for reactor life management. Measurements accumulated over many years have shown that there is considerable variability in deformation rates between different tubes. The deformation behavior not only varies from tube to tube but also varies along the length of each tube; this axial variation in itself is unique to each tube. The deformation behavior is determined by the microstructure and this is a function of the manufacturing history. An exhaustive study has been conducted to collect and compile deformation and manufacturing data for tubes in many different reactors. The manufacturing data have been analyzed with respect to the measured in-reactor performance of the pressure tubes. Data from pressure tubes fabricated over a 30-year period have been analyzed. During this time the manufacturing parameters have evolved to improve the workability and mechanical properties and also reduce tube-to-tube variability. The effect of manufacturing variables on in-reactor performance has been assessed for tubes that have been in service for many years and approaching the end of their design life. The analysis shows that the two most important factors that affect the deformation behavior of pressure tubes are the material source (ingot) and the extrusion conditions. These are related by the microstructure (texture, grain size, and dislocation density) to the deformation. The results of the statistical analyses will be presented and discussed in terms of the manufacturing conditions that bring about specific microstructures.
Effect of Irradiation Damage on the Deformation Properties of Zr-2.5Nb Pressure Tubes
Authors: Griffiths M. , Wang N. , Buyers A. , Donohue S. A.
Published: September 2009

The diametral expansion, elongation, and sag rates of Zr-2.5Nb pressure tubes in CANDU® (CANada Deuterium Uranium) nuclear reactors are important properties that limit their useful life and the maximum power level for reactor operation. As a result irradiation creep models are needed to predict the deformation behavior of the core components over the reactor life. It is important to know the creep behavior as a function of neutron flux in order to develop creep models over the range of operating conditions in the reactor core. At the edge of the reactor core, the neutron flux is decreasing very rapidly and there is a complex transition in creep behavior from irradiation-dominated creep to thermal-dominated creep. Also, mechanical properties such as tensile strength, fracture toughness, and delayed hydride-cracking are changing in the transition from thermal to irradiation conditions at the edge of the reactor core. Detailed studies have been completed on a Zr-2.5Nb tube irradiated in the NRU materials test reactor at Chalk River Laboratories. Pressure tube 601 was operating for a period of 66 950 h at temperatures ranging from about 547 K at the inlet and 571 K at the outlet. After the tube was removed in 1988 samples were taken for retrospective dosimetry to determine the fast neutron flux along the assembly. It was determined that the tube had been irradiated to a peak fluence of about 6×1025 n·m−2 corresponding to a fast neutron flux of about 2×1017 n·m−2·s−1. The flux profile was mapped and it was clear that the flux dropped rapidly to negligible values at about 0.5 m from the ends of the fueled zone. Samples of pressure tubes were taken for hardness testing and characterization by TEM and XRD analysis at various locations corresponding with different operating conditions (neutron flux and temperature) but at the same time. The creep behavior during operation was obtained by periodic gaging of the pressure tube internal diameter. The results of the microstructure characterization are presented and discussed in relation to the measured mechanical properties (creep and hardness). The microstructure and mechanical properties change significantly in the transition from the unirradiated state up to fluxes of about 1×1017 n·m−2·s−1.
Determination and Interpretation of Texture Evolution during Deformation of a Zirconium Alloy
Authors: Allen V. M. , Quinta da Fonseca J. , Preuss M. , Robson J. D. , Daymond M. , Comstock R. J.
Published: September 2009

Worldwide, crystal plasticity models are currently developed to predict texture development during processing of material. Such models require a precise knowledge of the active deformation mechanisms. The activation energy for certain deformation modes will change with temperature and also depend on the chemistry of the alloy as well as the microstructure. Deformation mechanisms were studied in ZIRLO™ during room and high-temperature uniaxial compression testing. Materials with a strong crystallographic basal texture and a more random texture due to β-quenching were investigated with the aim of establishing the effect of temperature, microstructure, and texture on the active deformation modes during the initial stages of deformation. First, specimens were strained at room temperature, 180°C and 300°C to 2 % and 5 % or 10 % total strain and subsequently analyzed by Electron Back Scatter Diffraction (EBSD) to determine the texture evolution. It was found that a dramatic texture change was observed for all testing temperatures in the strongly textured specimen after only 5 % total strain, which can only be understood in terms of tensile twinning of {101¯2} ⟨1¯011⟩ being active mainly at room temperature and compressive twinning of {112¯2} ⟨1¯1¯23⟩ being operational at room and elevated temperature. The β-quenched specimens did not show any evidence of texture change when strained to 10 %. In-situ intergranular strains were measured by time-of-flight neutron diffraction during continuous compressive loading. This information enabled the development of a crystal plasticity finite element model (CPFEM), which was subsequently used to predict the stress state in individual grains. It was found that in the strongly textured material the spread of intergranular strain in the {0002} grain family (normal pointing towards the ND direction) results in some grains being in compression even though the mean stresses are tensile, which could explain the activation of the observed compressive twinning. The crystal plasticity model also demonstrated that the observed texture changes in the strongly textured material, including those at high temperature, cannot be explained by slip alone even when ⟨c+a⟩ slip is considered. In addition, the model showed that the dramatic difference in yield strength of the two conditions studied here cannot be solely attributed to the difference in texture but that grain size plays an important role.
Irradiation-Induced Growth and Microstructure of Recrystallized, Cold Worked and Quenched Zircaloy-2, NSF, and E635 Alloys
Authors: Kobylyansky G. P. , Novoselov A. E. , Ostrovsky Z. E. , Obukhov A. V. , Shishin V. Yu. , Shishov V. N. , Nikulina A. V. , Peregud M. M. , Mahmood S. T. , White D. W. , Lin Y -P. , Dubecky M. A.
Published: September 2009

This paper is devoted to the study of the effect of the texture, phase composition, and microstructure on the irradiation-induced growth strain (GS) of zirconium-based alloys. GS measurements and TEM microstructural examinations were performed on Zry-2, NSF, and E635 samples in the recrystallized, beta quenched and cold-worked (CW) conditions. The samples were irradiated in the BOR-60 reactor in the temperature range of 315–325°C up to a neutron fluence level of 1.1 × 1026 n/m2 (E>1MeV), i.e., up to a damage dose of 23 dpa. Growth strains of NSF and E635 alloys in all states and in the longitudinal and transverse directions are lower as compared to those of Zry-2, and do not exceed 0.2 % even at the maximum fluence level. As for recrystallized Zry-2, the GS kinetics are characterized by the appearance of the accelerated growth stage. A combination of a certain amount of Nb, Fe, and Sn in the matrix content plays a key role in GS kinetics. The higher the degree of CW, the higher the irradiation growth but its rate of increase with increasing fluence is different for alloys of different compositions. The maximum GS, reaching 0.72 %, is observed in the 20 % CW Zry-2 samples. Texture, along with the alloy composition, is one of the main GS-determining factors. Irradiation growth of the transversal samples is lower as compared to the longitudinal ones because of texture. As for quenched alloys, the texture is practically isotropic and GS values are low, independent of the alloy composition. In CW materials, the density of ‹c›- dislocations greatly affects the irradiation growth strain. Particles of Zr(Fe,Cr)2 and Zr2(Fe,Ni) phases in Zry-2 as well as Zr(Nb,Fe)2 in NSF and E635 are depleted in iron under irradiation. The Fe goes into the matrix and modifies its properties. The HCP lattice structure in the Laves phases in NSF and E635 changes into BCC (β-Nb-type). FCC (Zr,Nb)2Fe precipitates preserve on the whole their composition and structure; no amorphization of the Nb-containing precipitates is observed. The Zr2(Fe,Ni) precipitates with a BCT lattice remain crystalline, and HCP Zr(Cr,Fe)2 precipitates undergo amorphization. The average particle size in the irradiated alloys is larger and the concentration is a little lower as compared to the unirradiated ones. Irradiation-induced fine dispersed precipitates about 3 nm in size, probably enriched in niobium, appear in NSF and E635. The observed changes of microhardness are discussed from the viewpoint of generation of radiation defects (clusters, dislocation loops), evolution of the initial dislocation structure, and matrix composition (enrichment in Fe, Cr, and, probably, Nb).
Toward a Better Understanding of Dimensional Changes in Zircaloy-4: What is the Impact Induced by Hydrides and Oxide Layer?
Authors: Blat-Yrieix M. , Ambard A. , Foct F. , Miquet A. , Beguin S. , Cayet N.
Published: September 2009

In pressurized water reactors, new operating conditions (higher burnup, new chemistry, etc.) can have an effect on the dimensional stability of the fuel assembly skeleton. Previous studies have shown that the fuel assembly growth is, among others parameters (free growth, creep), strongly driven by corrosion. Oxide layer and hydrides precipitation could both induce an increase of the dimensional parameters. The scope of the present study is, regardless of irradiation effect, to quantify and to understand the separate effects of hydrogen and oxide layers on the Zircaloy-4 dimensional changes. Experimental works have been performed in laboratory on stress relieved annealed (SRA) and recrystallized (RXA) Zircaloy-4 strips. First, the hydrogen impact on dimensional changes has been studied without the effect of the oxide layer. The measurements were performed at room temperature on strips previously pre-hydrided by the gaseous charging method. The hydrogen content of the samples was between 100 ppm and 2000 ppm. Results indicate a linear correlation between hydrogen content and length variation. RXA material is more affected by the hydrogen effect than the SRA material. Nevertheless, in comparison with data issued from out-of-reactor measurement, the impact of hydrides is not sufficient (with irradiation growth) to explain the post-irradiation examinations (PIE) results. To understand these differences, the oxide layer contribution must be quantified. Second, the impact of the oxide layer was therefore studied on RXA Zircaloy-4 strips. Corrosion tests have been performed in autoclave at 360°C in primary water (2 ppm Li-1000 ppm B-H2) on as-received and pre-hydrided materials. To obtain thicker oxide layer within a shorter duration, samples have been also oxidized in furnace at 415°C. Moreover, as no significant hydriding occurs during oxidation in air, we are able to characterize properly the specific effect of the oxide layer. As for hydrides' effect, an increase of strain is observed as the oxide thickness becomes thicker. The contributions of hydrides and oxide layer are then discussed with regard to the metallurgical properties of the alloy. Finally, all these results are compared with PIE observations. Free growth, hydride precipitation, and oxide thickness seem to be the three main parameters to explain the dimensional changes in Zircaloy-4 observed in reactor.
CASTA DIVA®: Experiments and Modeling of Oxide-Induced Deformation in Nuclear Components
Authors: Barberis P. , Rebeyrolle V. , Vermoyal J. J. , Chabretou V. , Vassault J. P.
Published: September 2009

In-pile deformation of nuclear structural components remains a major preoccupation for fuel vendors and utilities, especially when introducing new alloys or extending burn-ups. The usual invoked causes of such deformations are free growth, hydriding (hydrides induce a volume expansion which translates into deformation), irradiation creep if the component is subjected to external loads (e.g., axial load for the fuel assembly), and flux gradients (for fuel channel bow). In this paper we show through out-of-pile experiments (autoclave corrosion and low temperature air oxidation) and in-pile data analysis that the oxide layer significantly contributes to the deformation. A model (CASTA DIVA®) taking into account all the above mentioned phenomena has been derived and correctly predicts the deformation behavior, the stresses in the oxide layer, as well as the external cracking of the oxide. The relative impacts of hydriding and stresses in the metal are computed. A comparison between Zircaloy-4 and M5® is made, showing the better behavior of M5.
Cladding Tube Deformation Test for Stress Reorientation of Hydrides
Authors: Alam Adil M. , Hellwig Christian
Published: September 2009

The phenomenon of stress reorientation of hydrides in fuel cladding tubes was studied with the help of a test technique named as the “cladding tube deformation test” (CTDT). Unirradiated Zircaloy-2 (Zry-2) cladding tube specimens charged with 250 ppm of hydrogen were tested. The test consisted of heating specimens to 400°C and then cooling them down to room temperature at cooling rates of 0.7°C/min and 2°C/min. During the cooling phase, a constant tensile load was applied to specimens with the help of two inner cylinder halves. The stress-strain fields developed in the tube specimen were calculated with the help of the finite element method (FEM). Post-testing metallographic observations revealed considerable amount of hydrides oriented radially as a result of loading. Lengths and relative position of all hydride bands were determined through a semi-automatic image processing technique. Mapping of FEM calculated stress field on the metallographic section helped to determine a stress threshold value of 72 MPa. Tests performed with a cooling rate 2°C/min showed a considerably lower extent of hydride reorientation. The above test technique was validated by performing internal pressurization tests using pre-hydrided Zry-2 cladding tubes. The external diameter of the tube was tapered by fine turning in order to exert different uniform hoop stress values on different cross sections of the tube. Internal pressurization tests were performed with a maximum temperature of 400°C and a cooling rate of 0.5°C/min. Post-testing metallographic observations on several cross sections of the specimen revealed a stress reorientation threshold of 65 MPa. Percentage of radially oriented hydrides as well as their lengths increased with the stress level until it reached a plateau. Further tests were carried out with the CTDT technique on specimens with H2 content ranging from 560 ppm to 750 ppm. These results revealed a reorientation stress threshold of around 120 MPa for higher hydrogen content.
Evaluation of Hydride Reorientation Behavior and Mechanical Properties for High-Burnup Fuel-Cladding Tubes in Interim Dry Storage
Authors: Aomi Masaki , Baba Toshikazu , Miyashita Toshiyasu , Kamimura Katsuichiro , Yasuda Takayoshi , Shinohara Yasunari , Takeda Toru
Published: September 2009

The hydride stress reorientation behavior and the mechanical properties of irradiated cladding tubes were investigated to evaluate the high-burnup fuel-cladding tube properties in interim dry storage. As for the boiling water reactor (BWR) Zircaloy-2 (Zry-2) cladding, the hydride reorientation to the radial direction occurred at relatively low hoop stresses during the hydride reorientation treatment (HRT), such as less than 70 MPa. The increase of reorientation with hoop stress was not monotonic for the specimens in which a part of the hydrides remained precipitated at the HRT temperature, such as the case for 50GWd/t type cladding at a 300°C HRT. The degree of reorientation depended on the HRT solution temperature rather than on the estimated temperature at which the hydride precipitation occurred under the relatively moderate HRT conditions. In the relatively low cooling rate HRT, the hydride preferential precipitation in the Zr liner increased for Zr lined cladding compared to that in a relatively high cooling rate. The ductility of the specimens after the 300°C HRT showed relatively good correlation to the Polymax index which reflects the length or continuity of the hydrides regardless of their orientation. The ductility of the specimens after the 400°C, 0 MPa, 30°C/h HRT increased in ring compression testing at room temperature compared to no HRT (as-irradiated) specimens, and it indicated recovery of irradiation damage occurred at the 400°C annealing temperature and affected the ductility of the irradiated Zry-2 cladding. As for the pressurized water reactor Zircaloy-4 cladding, little increase in the radial hydride ratio occurred in a 100 MPa, 340°C or less HRT. On the other hand, the amount and the length of the hydride in the midwall area of the cladding depended on the temperature and the cooling rate from the HRT due to hydrogen migration from the hydride rim area. It is deduced that the ductility in ring compression deformation was affected by the orientation, amount, and length of hydride in the midwall area.
Experimental and Modeling Approach of Irradiation Defects Recovery in Zirconium Alloys: Impact of an Applied Stress
Authors: Ribis Joë , Onimus Fabien , Béchade Jean-Luc , Doriot Sylvie , Cappelaere Chantal , Lemaignan Clément , Barbu Alain , Rabouille Olivier
Published: September 2009

During neutron irradiation, both interstitial and vacancy loops are formed in high concentration in zirconium alloys. Due to this high density of loops, the material is considerably hardened, but the recovery of the radiation damage during a heat treatment leads to a progressive softening of the irradiated material. The recovery of the radiation induced hardening has been investigated using microhardness tests. Transmission electron microscopy (TEM) observations performed on irradiated foils have also shown that the loop density falls while the loop size increases during the thermal annealing. Furthermore, the TEM analysis has revealed that only vacancy loops are present in the material after long term annealing, the interstitial loops having entirely disappeared. A numerical cluster dynamic modeling has also been used in order to reproduce the material recovery for various annealing conditions. The microstructural evolution during mechanical testing with various loading conditions has also been studied. It has been shown that during a creep test with low applied stress (130 MPa) and high temperature (450°C), the microstructure evolution can essentially be explained by the thermal recovery of the loops leading to glide of dislocations as found for an non-irradiated material. At intermediate temperature (400°C), it is shown that for low stress level (130 MPa) the microstructure evolution can also be explained by the thermal recovery of loops, whereas for higher stress (250 MPa), sweeping of loops by gliding dislocations can also occur. In addition, for an applied stress of 130 MPa and a temperature of 400°C, dislocation density is higher in the irradiated material than in the non-irradiated material deformed in the same conditions. It is also shown that secondary slip systems are more activated in the irradiated material than in the non-irradiated material. From this detailed analysis, the mechanical behavior during creep is interpreted in terms of microscopic deformation mechanisms.
Investigation of Structural and Chemical Uniformity of Zr2.5% Nb and E635 Alloy by Radioactive Indicators
Authors: Arzhakova Valentine , Shikov Alexander , Kabanov Alexander , Belov Vladimir , Shtutsa Mikhail , Ziganshin Alexander , Kurapov Valeriy , Kurapova Liya
Published: September 2009

The performance and structure of final nuclear zirconium alloy components is formed at all processing steps, from ingot melting and subsequent forging, quenching, pilgering, and heat treatment. The subject of this investigation was the structure of zirconium alloy ingots made by double melting in a vacuum arc furnace, and investigated by the use of radioactive indicators (RAIs). The isotopes that were used were W185 and C14, which were added in very small quantities to the charge before melting the ingot. The RAI isotopes decorate the ingot structure, and provide an opportunity to investigate the behavior of impurity levels during solidification of the molten pool, and the structure in the various types of solidification zones within the ingot. The forged material and other products were investigated using the C14 isotope. Carbon distribution data were obtained for both Zr2.5Nb and for E635 structures on both macroscopic and microscopic levels. The structures and phases in which the carbon isotope was concentrated were discovered.
Contribution of Thermodynamic Calculations to Metallurgical Studies of Multi-Component Zirconium Based Alloys
Authors: Toffolon-Masclet C. , Brachet J. C. , Servant C. , Joubert J. M. , Barberis P. , Dupin N. , Zeller P.
Published: September 2009

The Zircobase thermodynamic database for zirconium alloys coupled with Thermo-Calc software represents a powerful tool for prediction of thermodynamic and metallurgical data such as activities, formation enthalpies, phase transformation temperatures, solubility limits, existence temperature range, and chemical compositions of second phase precipitates. This database was built up with binary and ternary descriptions assessed according to the CALPHAD methodology. It is sometimes necessary to take into account new systems, but also new versions of binary descriptions as recently experienced. For example, the two binary systems Zr-Fe and Zr-Sn had to be updated in order to fit new experimental results believed to be more accurate than previously available. This paper aims at showing the improvements of the database taking into account new descriptions of binary systems, as also ternary description such as Zr-Fe-Cr, Zr-Fe-Ni, and Zr-Nb-Fe. For this last ternary system, new experimental data were necessary. New experimental study of the two ternary phases, hexagonal Zr(Nb,Fe)2 and cubic (Zr,Nb)4Fe2, allowed their crystal structures (P63/mmc and Fd3¯ m, respectively) to be checked. This was useful to build up a sublattice model giving account of the existence of a composition range for these two intermetallic phases. Moreover, several specific ZrNbFe alloys were fabricated and annealed for times ranging from 1000 h to 10,000 h at 550, 700, 800, and 900°C to determine the equilibrium binary and ternary phase domains as a function of the temperature. All these data were used to obtain an improved thermodynamic modelling of this system. Finally, we illustrate some thermodynamic predictions of the different phases evolutions as a function of the temperature on multi-alloyed industrial type alloys. These examples show quite good agreement between the thermodynamic predictions and the experimental data derived from calorimetric experiments and microstructural observations.
Tearing Crack Growth and Fracture Micro-Mechanisms Under Micro Segregation in Zr-2.5%Nb Pressure Tube Material
Authors: Kapoor K. , Saibaba N. , Kashyap B. P. , Ramana Rao A. V.
Published: September 2009

In the case of Zr-2.5%Nb pressure tube material, certain tramp elements (Cl, P, and C) have a deleterious effect on the fracture properties. In order to dramatically reduce the amount of these impurities, vacuum arc re-melting is adopted. The effect of the melting practice (double or quadruple) on the fracture properties of this material has been previously studied in detail. However, in these studies, the micro-mechanisms of fracture and the role of the trace impurities on crack initiation and propagation in the sub-critical regime were not investigated in detail. In the present study, the mechanisms operating during the three stages of crack growth, i.e., initiation, propagation and fracture, in the case of low toughness double melted material, are proposed based on a detailed study. In our observation, the tramp elements segregate in the form of fine stringers. In the regions away from the segregation, the material is found to be very ductile with the appearance of ligaments with high local plastic deformation before fracture. The presence of such ligaments on the fracture surface is an indication of fracture occurring in a transition zone (ductile to brittle). In the case of the failure in this transition region, both cleavage and ductile mechanisms can occur in the same specimen. In the transition region, near the upper shelf region, the initiation of the crack occurs by cleavage at a local discontinuity, but the toughness increases rapidly and crack propagation occurs in ductile manner by formation of microvoids. The crack front propagates finding the local discontinuities, leaving behind the unbroken regions with high toughness (ligaments). As the crack propagation continues and the crack face opens, the ligaments left well behind the crack tip rupture after the crack front moves further. The electron micro-beam analysis and X-ray mapping show a build-up of Cl concentration at the stringer sites. In the present analysis, the fracture behavior of double and quadruple melted Zr-2.5%Nb material is compared using a “tearing instability” criterion. Using this approach, an attempt is made to assess the enhanced safety margins, in terms of critical crack length, achieved by modification in the melting practice. First, an estimate from the small specimen J-R data was obtained. However, to obtain a realistic estimate for fracture under burst condition an appropriate scaling factor was applied.
Voluntary Motor Output Is Altered by Spike-Timing-Dependent Changes in the Human Corticospinal Pathway
Authors: Taylor Janet L. , Martin Peter G.
Published: September 2009

Repeated pairs of timed presynaptic and postsynaptic potentials cause lasting changes in efficacy of transmission at many synapses. The corticospinal tract is the major pathway controlling voluntary movement in humans, and corticospinal neurons have monosynaptic connections to motoneurons of many muscles. We hypothesized that corticospinal transmission in humans could be altered by delivering, to the corticospinal–motoneuronal synapses, timed pairs of presynaptic volleys (produced by cortical stimulation) and antidromic postsynaptic volleys (by peripheral nerve stimulation). To test corticospinal transmission, electrical cervicomedullary stimuli evoked motor responses [cervicomedullary motor-evoked potentials (CMEPs)] in biceps brachii before and for 1 h after conditioning with 50 paired cortical and peripheral nerve stimuli. Seven interstimulus intervals (ISIs) of conditioning stimulus pairs were tested on different days. With one ISI (+3 ms; cortical before peripheral nerve stimulation), CMEPs were significantly increased in size by 33 ± 30% (mean ± SD;n= 7) from 4 until 32 min after conditioning. With two other ISIs (−13 ms, +22 ms), CMEPs were decreased from ∼30 until 60 min after conditioning (by 25 ± 23% and 27 ± 32%;n= 8). The remaining ISIs produced no changes. In a second study, subjects performed weak bilateral voluntary elbow flexion contractions before and after conditioning of the right elbow flexors. Conditioning ISIs that increased or decreased CMEPs similarly increased or decreased voluntary force and EMG on the right. Thus, depending on their timing, repeated paired stimuli can potentiate or depress corticospinal transmission, and these changes are functionally relevant. We suggest that bidirectional spike-timing-dependent plasticity can be induced at corticospinal–motoneuronal synapses and can influence voluntary motor output.