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R N Singh - One of the best experts on this subject based on the ideXlab platform.
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delayed hydride cracking of zr 2 5 nb Pressure Tube material due to partially constrained precipitates
Journal of Nuclear Materials, 2019Co-Authors: R N Singh, Narayana T Murty, Per StahleAbstract:Abstract Formation of partially constrained precipitates such as hydride blisters and oxide nodules have been reported on surfaces of Zr-alloy components of pressurised heavy water reactors and is associated with a large increase in volume. Such a change in volume imposes large stresses in the material surrounding the precipitate and may facilitate stable crack growth through delayed hydride cracking. In this work, the stress field of the partially constrained precipitates with different depth and aspect ratio has been computed using a finite element method. The computed stress field is used to predict the region in the matrix in which radial hydride is likely to form and fracture, by taking into consideration grain-size, texture and multi-axial state of stress. For a hypothetical crack just below the precipitate, stress intensity factors are estimated using material properties for both unirradiated and irradiated Pressure Tube materials. The results are compared with the threshold stress intensity factors required for crack growth due to delayed hydride cracking.
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Effect of radial hydride fraction on fracture toughness of CWSR Zr-2.5% Nb Pressure Tube material between ambient and 300 degrees C temperatures
'Wiley', 2018Co-Authors: Rk Sharma, R N Singh, A.k. Bind, Avinash G, Tewari A, B P KashyapAbstract:Pressure Tube spools fabricated from Zr-2.5% Nb alloy, were subjected to gaseous hydrogen charging and stress reorientation treatment to form radial hydrides in a specially designed fixture. Curved Compact Tension specimens of 17mm width containing 100 wppm of hydrogen were made using electro discharge machining. Some of the samples were annealed at 300, 325 and 350 degrees C followed by furnace cooling. Metallographic examination of the samples revealed 100% radial hydride in the samples to which stress reorientation treatment was imparted. The radial hydride fraction decreased with increase in the temperature of annealing that followed stress reorientation. Fracture toughness of samples with variable radial hydride fraction was evaluated as per ASTM E1820-13 procedure between 25 and 300 degrees C. All the samples exhibited a sharp ductile to brittle transition behavior and the transition temperature was observed to increase with increase in the radial hydride fraction. (C) 2018 Elsevier B.V. All rights reserved
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effect of specimen thickness on dhc velocity for zr 2 5nb alloy Pressure Tube material
Journal of Nuclear Materials, 2015Co-Authors: Sai Sunil, H K Khandelwal, R N Singh, Amit Kumar Bind, J.k. ChakravarttyAbstract:Abstract Zr-2.5Nb alloy Pressure Tubes used in pressurized heavy water reactors (PHWR) are susceptible to failure by Delayed Hydride Cracking (DHC), which is a form of localized hydride-embrittlement phenomenon manifested in the presence of a hydrostatic stress gradient and hydrogen concentration above a threshold limit as a sub-critical crack growth process in hydride forming metals. The DHC parameters used for safety assessment are DHC velocity (VDHC) and a threshold stress intensity factor (KIH). In this work DHC velocity was determined for Zr-2.5Nb Pressure Tube material at 250 and 280 °C using specimens of thickness between 1 mm and 4.5 mm. The DHC velocity was found to increase with increase in specimen thickness at 250 °C and 280 °C. Significant amount of tunneling of the crack was observed for 1-mm and 2-mm thick specimens at 250 °C and 280 °C, with the degree of tunneling increasing with decrease in thickness and increase in test temperature.
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influence of hydrogen content on fracture toughness of cwsr zr 2 5nb Pressure Tube alloy
Journal of Nuclear Materials, 2013Co-Authors: R N Singh, A.k. Bind, N S Srinivasan, Per StahleAbstract:In this work, influence of hydrogen and temperature on the fracture toughness parameters of unirradiated, cold worked and stress relieved (CWSR) Zr-2.5Nb Pressure Tube alloys used in Indian Pressurized Heavy Water Reactor is reported. The fracture toughness tests were carried out using 17 mm width curved compact tension specimens machined from gaseously hydrogen charged Tube-sections. Metallography of the samples revealed that hydrides were predominantly oriented along axial-circumferential plane of the Tube. Fracture toughness tests were carried out in the temperature range of 30-300 degrees C as per ASTM standard E-1820-06, with the crack length measured using direct current potential drop (DCPD) technique. The fracture toughness parameters (J(Q), J(Max) and dJ/da), were determined. The critical crack length (CCL) for catastrophic failure was determined using a numerical method. It was observed that for a given test temperature, the fracture toughness parameters representing crack initiation (J(Q)) and crack propagation (J(Max), and dJ/da) is practically unaffected by hydrogen content. Also, for given hydrogen content, all the aforementioned fracture toughness parameters increased with temperature to a saturation value. (c) 2012 Elsevier B.V. All rights reserved. (Less)
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threshold stress intensity factor for delayed hydride cracking in zr 2 5 nb Pressure Tube alloy
Materials Science and Engineering A-structural Materials Properties Microstructure and Processing, 2009Co-Authors: R N Singh, J.k. Chakravartty, Per Stahle, A A ShmakovAbstract:Delayed hydride cracking (DHC) velocity was determined at 203,227,250 and 283 degrees C using 17 mm width curved compact toughness specimens machined from an unirradiated Zr-2.5 wt.% Nb Pressure Tube spool, gaseously charged with 60 ppm of hydrogen by weight. Single CT specimen was used to determine DHC velocity at a constant temperature for a range of stress intensity factor (K-1) obtained by load drop method. For a given temperature and K-1 > 15 Mpa m(1/2), DHC velocity was found to be practically independent of K-1. For 15 > K-1 > 10 MPa m(1/2), DHC velocity decreased significantly with decrease in stress intensity factor and extrapolation of the data suggested the threshold stress intensity factor to be about 9-11 MPa m(1/2) in the aforementioned temperature range. The activation energy associated with DHC was observed to be 35.1 kJ/mol. (C) 2009 Elsevier B.V. All rights reserved.
Per Stahle - One of the best experts on this subject based on the ideXlab platform.
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delayed hydride cracking of zr 2 5 nb Pressure Tube material due to partially constrained precipitates
Journal of Nuclear Materials, 2019Co-Authors: R N Singh, Narayana T Murty, Per StahleAbstract:Abstract Formation of partially constrained precipitates such as hydride blisters and oxide nodules have been reported on surfaces of Zr-alloy components of pressurised heavy water reactors and is associated with a large increase in volume. Such a change in volume imposes large stresses in the material surrounding the precipitate and may facilitate stable crack growth through delayed hydride cracking. In this work, the stress field of the partially constrained precipitates with different depth and aspect ratio has been computed using a finite element method. The computed stress field is used to predict the region in the matrix in which radial hydride is likely to form and fracture, by taking into consideration grain-size, texture and multi-axial state of stress. For a hypothetical crack just below the precipitate, stress intensity factors are estimated using material properties for both unirradiated and irradiated Pressure Tube materials. The results are compared with the threshold stress intensity factors required for crack growth due to delayed hydride cracking.
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effect of direction of approach of test temperature on fracture toughness of zr 2 5nb Pressure Tube material
Materials Science and Engineering A-structural Materials Properties Microstructure and Processing, 2015Co-Authors: H K Khandelwal, S. Sunil, A.k. Bind, Ramniwas Singh, B N Rath, Per StahleAbstract:Previous work by one of the authors reported the stress-field for a fully-constrained hydride embedded in matrix computed using finite element method to develop an understanding of the temperature dependence of susceptibility of Zr-alloys to hydride embrittlement. It was observed that the nature and magnitude of stress field in the matrix and hydride depended on whether the hydride is expanding (experienced while cooling from a peak temperature) or contracting (experienced while heating to the test temperature). On the basis of the observed dependence of the nature and magnitude of stress components in the matrix and hydride, it was suggested that the fracture toughness of hydrided Zr-alloys may depend on the direction of approach of the test temperature. In order to verify this, fracture tests were carried out using samples machined from hydrogen charged Zr-2.5Nb alloy Pressure Tube material as per ASTM standard E1820-11 in which the test temperature was attained by either heating to the test temperature or cooling from a peak temperature. The direction of approach of the test temperature did not affect the fracture toughness in the lower and upper shelf temperature regime, it indeed had an effect in the transition regime. The computed stress field could provide explanation for the observed axial splits on the fracture surfaces for both the heating and cooling cases. (C) 2014 Elsevier B.V. All rights reserved.
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influence of hydrogen content on fracture toughness of cwsr zr 2 5nb Pressure Tube alloy
Journal of Nuclear Materials, 2013Co-Authors: R N Singh, A.k. Bind, N S Srinivasan, Per StahleAbstract:In this work, influence of hydrogen and temperature on the fracture toughness parameters of unirradiated, cold worked and stress relieved (CWSR) Zr-2.5Nb Pressure Tube alloys used in Indian Pressurized Heavy Water Reactor is reported. The fracture toughness tests were carried out using 17 mm width curved compact tension specimens machined from gaseously hydrogen charged Tube-sections. Metallography of the samples revealed that hydrides were predominantly oriented along axial-circumferential plane of the Tube. Fracture toughness tests were carried out in the temperature range of 30-300 degrees C as per ASTM standard E-1820-06, with the crack length measured using direct current potential drop (DCPD) technique. The fracture toughness parameters (J(Q), J(Max) and dJ/da), were determined. The critical crack length (CCL) for catastrophic failure was determined using a numerical method. It was observed that for a given test temperature, the fracture toughness parameters representing crack initiation (J(Q)) and crack propagation (J(Max), and dJ/da) is practically unaffected by hydrogen content. Also, for given hydrogen content, all the aforementioned fracture toughness parameters increased with temperature to a saturation value. (c) 2012 Elsevier B.V. All rights reserved. (Less)
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threshold stress intensity factor for delayed hydride cracking in zr 2 5 nb Pressure Tube alloy
Materials Science and Engineering A-structural Materials Properties Microstructure and Processing, 2009Co-Authors: R N Singh, J.k. Chakravartty, Per Stahle, A A ShmakovAbstract:Delayed hydride cracking (DHC) velocity was determined at 203,227,250 and 283 degrees C using 17 mm width curved compact toughness specimens machined from an unirradiated Zr-2.5 wt.% Nb Pressure Tube spool, gaseously charged with 60 ppm of hydrogen by weight. Single CT specimen was used to determine DHC velocity at a constant temperature for a range of stress intensity factor (K-1) obtained by load drop method. For a given temperature and K-1 > 15 Mpa m(1/2), DHC velocity was found to be practically independent of K-1. For 15 > K-1 > 10 MPa m(1/2), DHC velocity decreased significantly with decrease in stress intensity factor and extrapolation of the data suggested the threshold stress intensity factor to be about 9-11 MPa m(1/2) in the aforementioned temperature range. The activation energy associated with DHC was observed to be 35.1 kJ/mol. (C) 2009 Elsevier B.V. All rights reserved.
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influence of temperature on threshold stress for reorientation of hydrides and residual stress variation across thickness of zr 2 5nb alloy Pressure Tube
Journal of Nuclear Materials, 2006Co-Authors: R N Singh, Lala R Mikin, G K Dey, D N Sah, I S Batra, Per StahleAbstract:Threshold stress, σth, for reorientation of hydrides in cold worked and stress-relieved (CWSR) Zr–2.5Nb Pressure Tube material was determined in the temperature range of 523–673 K. Using tapered gage tensile specimen, mean value of σth was experimentally determined by two methods, half thickness method and area compensation method. The difference between local values of σth measured across the thickness of the Tube and the mean σth values yielded the residual stress variation across the Tube thickness. It was observed that both the mean threshold stress and residual stress decrease with increase in reorientation temperature. Also, the maximum value of residual stresses was observed near the midsection of the Tube.
Blair P Bromley - One of the best experts on this subject based on the ideXlab platform.
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physics characteristics of internally cooled annular fuel for potential application in Pressure Tube heavy water reactors
Annals of Nuclear Energy, 2019Co-Authors: Blair P Bromley, Ashlea V Colton, K Groves, Sourena GolesorkhiAbstract:Abstract This paper summarizes the results of exploratory lattice physics studies of alternative, advanced fuel bundle concepts that could potentially be implemented in Pressure Tube heavy water reactors (PT-HWRs). The lattice physics code WIMS-AECL was used to analyze the physics performance and operational characteristics of an 18-element and a 12-element internally cooled annular fuel (ICAF) fuel bundle, made with (LEU,Th)O2 fuel, with both low-burnup and high-burnup options. Such fuel bundles with annular fuel elements may be able to operate at higher bundle power levels and with higher linear element (LER) ratings than fuel bundles with conventional solid cylindrical fuel elements. In addition, the use of thorium mixed with LEU can help extend uranium resources, exploit the energy potential in thorium, and also reduce the production of plutonium and minor actinides, due to the smaller fraction of 238U in the fuel. The influence of improvements in the neutron economy of these lattices was also investigated, by incorporating higher-purity heavy water moderator and coolant (99.90 at.% D2O) in the models, along with enriched zirconium (95 wt% 90Zr/Zr) for the zirconium alloys used in the structural components. The results were compared with those for a more conventional 37-element PT-HWR fuel bundle using natural uranium (NU) fuel. Results show that annular fuels could be very attractive, being able to achieve higher burnup, comparable or better fissile utilization, reduced coolant void reactivity, and comparable or more negative fuel temperature coefficients.
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physics characteristics of internally cooled annular fuel for potential application in Pressure Tube heavy water reactors
Annals of Nuclear Energy, 2019Co-Authors: Blair P Bromley, Ashlea V Colton, K Groves, Sourena GolesorkhiAbstract:Abstract This paper summarizes the results of exploratory lattice physics studies of alternative, advanced fuel bundle concepts that could potentially be implemented in Pressure Tube heavy water reactors (PT-HWRs). The lattice physics code WIMS-AECL was used to analyze the physics performance and operational characteristics of an 18-element and a 12-element internally cooled annular fuel (ICAF) fuel bundle, made with (LEU,Th)O2 fuel, with both low-burnup and high-burnup options. Such fuel bundles with annular fuel elements may be able to operate at higher bundle power levels and with higher linear element (LER) ratings than fuel bundles with conventional solid cylindrical fuel elements. In addition, the use of thorium mixed with LEU can help extend uranium resources, exploit the energy potential in thorium, and also reduce the production of plutonium and minor actinides, due to the smaller fraction of 238U in the fuel. The influence of improvements in the neutron economy of these lattices was also investigated, by incorporating higher-purity heavy water moderator and coolant (99.90 at.% D2O) in the models, along with enriched zirconium (95 wt% 90Zr/Zr) for the zirconium alloys used in the structural components. The results were compared with those for a more conventional 37-element PT-HWR fuel bundle using natural uranium (NU) fuel. Results show that annular fuels could be very attractive, being able to achieve higher burnup, comparable or better fissile utilization, reduced coolant void reactivity, and comparable or more negative fuel temperature coefficients.
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Modeling and mitigation of bundle end power peaking in Pressure Tube heavy water reactor advanced fuels using thorium dioxide
Annals of Nuclear Energy, 2018Co-Authors: Clifford Dugal, Ashlea V Colton, Sourena Golesorkhi, Blair P BromleyAbstract:Abstract Fuel bundle end power peaking in Pressure Tube heavy water reactors (PT-HWRs), particularly considering advanced reactor fuels (such as thorium-based fuels), was evaluated and a mitigation method was assessed. For bundles containing all the same fuel pellets, power is greatest at the bundle ends due to higher thermal neutron flux there. A method to achieve a flatter axial power profile along a bundle by downblending the fissile content in the end pellets with thorium dioxide was evaluated. The method was shown to be effective, and only a slight reduction in fresh fuel reactivity was observed.
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comparisons between rfsp and mcnp for modeling Pressure Tube heavy water reactor cores with thorium based fuels
Annals of Nuclear Energy, 2018Co-Authors: Huiping Yan, Blair P Bromley, Ashlea V Colton, Clifford Dugal, Sourena GolesorkhiAbstract:Abstract Thorium-based fuels are recognized to hold significant promise as an option for achieving a long-term, sustainable nuclear fuel cycle and energy security. Pressure Tube heavy water reactors (PT-HWRs) are well suited to exploit the energy potential of thorium. Deterministic reactor physics codes are often used in exploratory studies to evaluate the performance and operational characteristics of various fuel bundle, lattice and core concepts with thorium based fuels in PT-HWRs. Because of the approximations inherent in deterministic codes, they are often considered less accurate than stochastic codes. In order to enhance confidence in deterministic code-based predictions, these codes are often benchmarked against stochastic codes, when experimental data is not available for code validation. Code-to-code comparisons of core physics calculations were made between the deterministic reactor physics toolset WIMS-AECL/WIMS-Utilities/RFSP and the stochastic neutron transport code MCNP for a series of core configurations with mixed oxide fuels containing thorium in PT-HWRs. The core neutron multiplication factors (keff) appear to have a difference (RFSP-MCNP) ranging between −2.4 mk and +4.0 mk. The MCNP full-core calculations confirm that that thorium-based fuels have a lower coolant void reactivity (CVR), ranging from +8.3 mk to +11.3 mk (versus 14 mk for NU fuel). The core cases with NU fuel have a small difference (RFSP-MCNP) in peak bundle power (ranging between −0.13% and 0.65%). Cores with LEU at 1.2 wt% 235U/U, Pu/Th, and LEU/Th (LEU at 5 wt% 235U/U) fuel have higher differences in peak bundle power (ranging between −7% and −12%). All these core cases have peak channel power differences between −2.1% and −8.9%. Core with 233U fuel has the smallest peak bundle difference (−0.05%) and smallest peak channel differences (−0.58%) which represent the best agreement between MCNP and RFSP simulations. The performed code-to-code comparisons have demonstrated that the core physics parameters estimated by RFSP calculations are consistent with MCNP simulations, especially for fuel where the main fissile component are 235U-based and 233U-based fuel.
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lattice physics evaluation of 35 element mixed oxide thorium based fuels for use in Pressure Tube heavy water reactors
Annals of Nuclear Energy, 2018Co-Authors: Ashlea V Colton, Blair P BromleyAbstract:Abstract A series of 2-D lattice physics calculations with depletion were carried with WIMS-AECL Version 3.1 out as part of exploratory scoping studies to evaluate various thorium-based fuel bundle concepts for potential application in Pressure Tube heavy water reactors (PT-HWRs). Fuel bundles concepts investigated consisted of a cluster of 35 fuel elements arranged in two rings (14 + 21), and surrounding a central graphite displacer rod. The fuel is comprised of thorium dioxide mixed with a fissile driver of reactor-grade plutonium (∼67 wt% Pu fissile /Pu; 3.5–4.5 wt% PuO 2 /(Pu,Th)O 2 ), low enriched uranium (5 wt% 235 U/U; 40–50 wt% LEUO 2 /(LEU,Th)O 2 ) or uranium–233 (1.8 wt% 233 UO 2 /( 233 U,Th)O 2 ). Estimates of burnup-averaged fuel temperature coefficients (FTC) and coolant void reactivity (CVR) were found to be lower than those for conventional natural uranium dioxide (NUO 2 ) PT-HWR fuel in a 37-element bundle. A low-burnup option for using (LEU,Th)O 2 fuel in a PT-HWR is found to be attractive as a means for extracting energy from thorium, while also generating stockpiles of 233 U, and demonstrating enhanced safety characteristics with reduced CVR and FTC relative to NUO 2 .
A A Shmakov - One of the best experts on this subject based on the ideXlab platform.
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threshold stress intensity factor for delayed hydride cracking in zr 2 5 nb Pressure Tube alloy
Materials Science and Engineering A-structural Materials Properties Microstructure and Processing, 2009Co-Authors: R N Singh, J.k. Chakravartty, Per Stahle, A A ShmakovAbstract:Delayed hydride cracking (DHC) velocity was determined at 203,227,250 and 283 degrees C using 17 mm width curved compact toughness specimens machined from an unirradiated Zr-2.5 wt.% Nb Pressure Tube spool, gaseously charged with 60 ppm of hydrogen by weight. Single CT specimen was used to determine DHC velocity at a constant temperature for a range of stress intensity factor (K-1) obtained by load drop method. For a given temperature and K-1 > 15 Mpa m(1/2), DHC velocity was found to be practically independent of K-1. For 15 > K-1 > 10 MPa m(1/2), DHC velocity decreased significantly with decrease in stress intensity factor and extrapolation of the data suggested the threshold stress intensity factor to be about 9-11 MPa m(1/2) in the aforementioned temperature range. The activation energy associated with DHC was observed to be 35.1 kJ/mol. (C) 2009 Elsevier B.V. All rights reserved.
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Threshold stress intensity factor for delayed hydride cracking in Zr-2.5%Nb Pressure Tube alloy
2009Co-Authors: Singh, Ram N, J.k. Chakravartty, Ståhle Per, A A ShmakovAbstract:Delayed hydride cracking (DHC) velocity was determined at 203, 227, 250 and 283 °C using 17 mm width curved compact toughness specimens machined from an unirradiated Zr–2.5 wt.% Nb Pressure Tube spool, gaseously charged with 60 ppm of hydrogen by weight. Single CT specimen was used to determine DHC velocity at a constant temperature for a range of stress intensity factor (KI) obtained by load drop method. For a given temperature and KI > 15 MPa m1/2, DHC velocity was found to be practically independent of KI. For 15 > KI > 10 MPa m1/2, DHC velocity decreased significantly with decrease in stress intensity factor and extrapolation of the data suggested the threshold stress intensity factor to be about 9–11 MPa m1/2 in the aforementioned temperature range. The activation energy associated with DHC was observed to be 35.1 kJ/mol
M R Daymond - One of the best experts on this subject based on the ideXlab platform.
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thermal creep behavior in heat treated and modified textured zr excel Pressure Tube material
Materials Science and Engineering A-structural Materials Properties Microstructure and Processing, 2017Co-Authors: Kazi F Ahmmed, Levente Balogh, Yasir Idrees, David Kerr, M R DaymondAbstract:Abstract Thermal creep properties of Zr-Excel alloy (Zr-3.5 wt% Sn-0.8 wt% Nb-0.8 wt% Mo) have been investigated in this study. The limits of existing experimental data on the current material and the search for more isotropic properties in Pressure Tube (PT) material motivates the current study. Heat-treatment on Excel PT materials leads to significant microstructural changes, where a strong texture of the As-Received (ASR) PT materials has been altered to a different degree of randomness at various solution temperatures. Microstructural alteration through heat-treatments should have an influence on the thermal creep behavior of the current material. It is observed that at a relatively high creep test temperature (>300 °C), steady-state creep rate of both ASR and heat-treated materials depends on the applied stress in a power law fashion and approaches stress exponent value of ~3. However, at a relatively low test temperature, n showed a high value in all treatments. The activation energy ( Q ) for the creep mechanism has been found to be microstructure-dependent. The ASR material has a Q value of ~73 to ~114 kJ/mol at the applied test temperature range (150–350 °C). However, martensitic structure formation and other microscopic changes (dislocation structures and elemental segregation) caused significant increase in Q values up to ~325 kJ/mol in the heat-treated materials. A significant decrease has been observed in creep anisotropy in the WQ-895 treatment for which texture has been moderately randomized.
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microstructural evaluation and crystallographic texture modification of heat treated zirconium excel Pressure Tube material
Journal of Alloys and Compounds, 2016Co-Authors: Kazi F Ahmmed, M R Daymond, Michael A GharghouriAbstract:Abstract Dual phase (α − β) Zr-Excel (Zr-3.5 wt% Sn-0.8 wt% Mo-0.8 wt% Nb) Pressure Tube (PT) material was solution treated at a range of temperatures in the (α + β) Zr or β Zr -phase field, and cooled in water or air to generate various microstructures. TEM and neutron diffraction revealed significant alterations during the heat-treatments in both the microstructure and the initial strong transverse texture of the As-Received (ASR) PT material. The proportion of transformed β-phase (β → α) or martensitic/acicular α-phase increased progressively with increasing solution temperature. The morphologies of the transformed β products are different in the water-quenched material (martensitic plates) than in the air-cooled material (relatively wide, parallel plates). The room temperature texture of the heat-treated material shows that: the extent to which the initially strong transverse texture is modified (basal poles shifting towards the axial direction) increases with increasing solution temperature, yielding a relatively random texture at higher solution treatment temperature (quenching from ∼930 °C). This texture modification through heat-treatment is more effective by water-quenching than by air-cooling. Although, a randomized texture is developed in the α → β → α phase transformation, variant selection is observed. It is suggested that this selection occurs primarily during cooling (β → α), whereas no (or very weak) variant selection occurs during heating (α → β), as indicated by the identical room temperature β-phase (remnant) textures in the heat-treated and ASR materials.
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effect of neutron irradiation on deformation mechanisms operating during tensile testing of zr 2 5nb
Acta Materialia, 2016Co-Authors: Fei Long, Levente Balogh, Donald W Brown, Paula Mosbrucker, Travis Skippon, C D Judge, M R DaymondAbstract:Abstract In situ neutron diffraction has been carried out on fast neutron irradiated Zr–2.5Nb Pressure Tube and non-irradiated Zr–2.5Nb Pressure Tube samples, with tensile deformation applied along two texture directions. Through the evolution of intergranular strains with applied stress, as measured via the lattice strains of individual grain families, the deformation mechanism of the Pressure Tube material with and without fast neutron irradiation has been studied. Prismatic , basal and pyramidal has been found to operate in both irradiated and non-irradiated Zr–2.5Nb alloy at different stress levels. Furthermore, by line profile analysis, it is shown that the density of type dislocations in the irradiated sample changed very little during deformation, whereas the density of pyramidal dislocations increased continuously with plastic strain once plastic deformation started. Therefore, it is found that the neutron irradiation induced defects selectively harden prismatic and basal slip, and have very small impact on the operation of pyramidal slip system.
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aging response and characterization of precipitates in zr alloy excel Pressure Tube material
Journal of Nuclear Materials, 2014Co-Authors: M Sattari, R A Holt, M R DaymondAbstract:Abstract Precipitation hardening in the Zr-based alloy Excel (Zr–3.5 wt.% Sn–0.8 wt.% Mo–0.8 wt.% Nb) was studied using hardness testing and transmission electron microscopy (TEM). Solution treatment at 890 °C, in the αZr + βZr region, and 980 °C, in the βZr region, followed by water-quenching and aging resulted in an increase of hardness, of up to 47% compared to the annealed material. The optimum condition for aging heat treatment was found to be 450 °C for 1–2 h. The precipitates were observed only in the transformed βZr martensitic phase. Energy dispersive X-ray spectroscopy (EDS) in TEM showed the precipitate composition to be Zr–30 wt.% Mo–25 wt.% Nb–2 wt.% Fe. The crystal structure of the precipitates was determined to be hexagonal with a = 0.294 nm and c = 0.448 nm, i.e. c/a = 1.526.
