Search results for: zircaloy

Authors: Bharat Kumar, Deepak Kumar, Vijay Chaudhry

Abstract:

Zirconium alloys are widely used for the core components of a pressurized heavy water reactor (PHWR) or Canada deuterium (CANDU) reactor due to their low neutron absorption cross-section and excellent mechanical properties. The components made of Zirconium alloys are subjected to flow-induced vibrations, resulting in fretting wear at the interface of; pressure tubes and bearing pads, pressure tubes and calandria tubes, and calandria tubes and Liquid injection shutdown system (LISS) nozzles. There is a need to explore the tribological response under such conditions. Present work simulates the contact between calandria tube and LISS nozzle of PHWR/CANDU reactor as cylinder-on-cylinder contact configuration. Reciprocating tribo-tests were conducted on Zircaloy-4 (Zr-4) under the self-mated condition at varying amplitude, frequency, and sliding time. To understand the active wear mechanism, worn surfaces were analyzed using Scanning Electron Microscopy (SEM) and Energy Dispersive Spectroscopy (EDS). The change in amplitude severely affects the wear than other factors. The wear mechanism transits from adhesion to abrasion with increasing test amplitude. The dominant wear mechanisms are micro-cutting and micro-plowing followed by delamination in some areas. However, the coefficient of friction has indifferent behaviors.

Keywords: zircaloy-4, tribology, calandria tube, LISS nozzle, PHWR

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Authors: Jefri Draup, Antoine Ambard, Chi-Toan Nguyen

Abstract:

Zirconium alloys exhibit solid-state phase transformations under thermal loading. These can lead to a significant evolution of the microstructure and associated mechanical properties of materials used in nuclear fuel cladding structures. Therefore, the ability to capture effects of phase transformation on the material constitutive behavior is of interest during conditions of severe transient thermal loading. Whilst typical Avrami, or Johnson-Mehl-Avrami-Kolmogorov (JMAK), type models for phase transformations have been shown to have a good correlation with the behavior of Zircaloy-4 under constant heating rates, the effects of variable and fast heating rates are not fully explored. The present study utilises the results of in-situ high energy synchrotron X-ray diffraction (SXRD) measurements in order to validate the phase transformation models for Zircaloy-4 under fast variable heating rates. These models are used to assess the performance of fuel cladding structures under loss of coolant accident (LOCA) scenarios. The results indicate that simple Avrami type models can provide a reasonable indication of the phase distribution in experimental test specimens under variable fast thermal loading. However, the accuracy of these models deteriorates under the faster heating regimes, i.e., 100Cs⁻¹. The studies highlight areas for improvement of simple Avrami type models, such as the inclusion of temperature rate dependence of the JMAK n-exponent.

Keywords: accident, fuel, modelling, zirconium

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Authors: R. S. Rajpurohit, N. C. Santhi Srinivas, Vakil Singh

Abstract:

Asymmetric stress cycling leads to accumulation of plastic strain which is called as ratcheting strain. The problem is generally associated with nuclear fuel cladding materials used in nuclear power plants and pressurized pipelines. In the present investigation, asymmetric stress controlled fatigue tests were conducted with three different parameters namely, mean stress, stress amplitude and stress rate (keeping two parameters constant and varying third parameter) to see the plastic strain accumulation and its effect on fatigue life and deformation behavior of Zircaloy-2 at 400°C. The tests were conducted with variable mean stress (45-70 MPa), stress amplitude (95-120 MPa) and stress rate (30-750 MPa/s) and tested specimens were characterized using transmission and scanning electron microscopy. The experimental results show that with the increase in mean stress and stress amplitude, the ratcheting strain accumulation increases with reduction in fatigue life. However, increase in stress rate leads to improvement in fatigue life of the material due to small ratcheting strain accumulation. Fractographs showed a decrease in area fraction of fatigue failed region.

Keywords: asymmetric cyclic loading, ratcheting fatigue, mean stress, stress amplitude, stress rate, plastic strain

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Authors: Alexia Wu, Joel Ribis, Jean-Christophe Brachet, Emmanuel Clouet, Benoit Arnal, Elodie Rouesne, Stéphane Urvoy, Justine Roubaud, Yves Serruys, Frederic Lepretre

Abstract:

To enhance the safety of Light Water Reactor, accident tolerant fuel (ATF) claddings materials are under development. In the framework of CEA-AREVA-EDF collaborative program on ATF cladding materials, CEA has engaged specific studies on chromium coated zirconium alloys. Especially for Loss-of-Coolant-Accident situations, chromium coated claddings have shown some additional 'coping' time before achieving full embrittlement of the oxidized cladding, when compared to uncoated references – both tested in steam environment up to 1300°C. Nevertheless, the behavior of chromium coatings and the stability of the Zr-Cr interface under neutron irradiation remain unknown. Two main points are addressed: 1. Bulk Cr behavior under irradiation: Due to its BCC crystallographic structure, Cr is prone to Ductile-to-Brittle-Transition at quite high temperature. Irradiation could be responsible for a significant additional DBTT shift towards higher temperatures. 2. Zircaloy/Cr interface behavior under irradiation: Preliminary TEM examinations of un-irradiated samples revealed a singular Zircaloy-4/Cr interface with nanometric intermetallic phase layers. Such particular interfaces highlight questions of how they would behave under irradiation - intermetallic zirconium phases are known to be more or less stable under irradiations. Another concern is a potential enhancement of chromium diffusion into the zirconium-alpha based substrate. The purpose of this study is then to determine the behavior of such coatings after ion irradiations, as a surrogate to neutron irradiation. Ion irradiations were performed at the Jannus-Saclay facility (France). 20 MeV Kr8+ ions at 400°C with a flux of 2.8x1011 ions.cm-2.s-1 were used to irradiate chromium coatings of 1-2 µm thick on Zircaloy-4 sheets substrate. At the interface, the calculated damage is close to 10 dpa (SRIM, Quick Calculation Damage mode). Thin foil samples were prepared with FIB for both as-received and irradiated coated samples. Transmission Electron Microscopy (TEM) and in-situ tensile tests in a Scanning Electron Microscope are being used to characterize the un-irradiated and irradiated materials. High Resolution TEM highlights a great complexity of the interface before irradiation since it is formed of an alternation of intermetallic phases – C14 and C15. The interfaces formed by these intermetallic phases with chromium and zirconium show semi-coherency. Chemical analysis performed before irradiation shows some iron enrichment at the interface. The chromium coating bulk microstructures and properties are also studied before and after irradiation. On-going in-situ tensile tests focus on the capacity of chromium coatings to sustain some plastic deformation when tested up to 350°C. The stability of the Cr/Zr interface is shown after ion irradiation up to 10 dpa. This observation constitutes the first result after irradiation on these new coated claddings materials.

Keywords: accident tolerant fuel, HRTEM, interface, ion-irradiation

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Authors: Ho-A Kim, Jae-Soo Noh

Abstract:

During reactor operation, hydride blister can occur in spent nuclear fuel (SNF) claddings, and it could worsen the integrity of the claddings locally. Hydride blister can be critical when a pinch-type load is applied in the process of SNF handling and transportation. Micro-cantilever tests were performed to evaluate the risk of local hydride blister by comparing the fracture toughness of local hydride blister and pre-hydrided Zr alloy matrix of SNF cladding on a microscale. Hydride blister was generated by a gaseous charging procedure to simulate an SNF cladding. Micro-cantilevers and pre-cracks were ion-milled with the Ga+ ion beam of FEI Helios 600 at 30kV acceleration voltage. Micro-cantilever tests were conducted using PI 85 pico-indenter (HYSTRON) with for sided conductive diamond flat tip (1 μm x 1 μm) at a speed of 5 nm/sec. The results show that the hydride blister specimen could be fractured in the elastic deformation region, and the fracture toughness of the hydride blister specimen could drop up to 60% of that of the pre-hydrided Zr alloy matrix. Therefore, local hydride blister can degrade the integrity of SNF cladding, and the effect of hydride blister should be taken into account when evaluating failure criteria of claddings during handling, storage, and transportation of SNF.

Keywords: fracture toughness, hydride blister, micro-cantilever test, spent nuclear fuel cladding.

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Authors: Bharat Kumar, Deepak Kumar, Vijay Chaudhry

Abstract:

Zirconium alloys are used as core components of nuclear reactors due to their high wear resistance, good corrosion properties, and good mechanical stability at high temperatures. The present work simulates the contact between the calandria tube and the liquid injection shutdown system (LISS) nozzle. The Calandria tube is the outer covering of the pressure tube. Water flows inside the pressure tube through fuel claddings which produces vibration in the pressure tube along with vibration in the calandria tube. Fretting wear takes place at the point of contact between the calandria tube and the LISS nozzle. Fretting tests were performed under different conditions, such as; varying fretting duration (i.e., 1 to 4 hours), varying frequency (i.e., 5 to 6.5 Hz), and varying amplitude (100 to 400 µm). The formation of the oxide layer was observed during the fretting wear test; as a result, the worn product. The worn surfaces were analyzed with scanning electron microscopy (SEM) to analyze the wear mechanism involved in the fretting test, and Energy dispersive x-ray spectroscopy (EDS) and Raman spectroscopy were used to confirm the presence of an oxide layer on the worn surface. The oxide layer becomes more uniform with fretting duration in case of water submerged condition as compared to dry contact condition. The oxide layer is deeply removed at high amplitude due to the change of wear mechanism from adhesion to abrasion, as confirmed by the presence of micro ploughing and micro cutting. Low amplitude fretting favors the formation of the tribo-oxide layer.

Keywords: tribo-oxide layer, wear, mechanically mixed layer, zircaloy

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Authors: Raphaelle Guillou, Matthieu Le Saux, Jean-Christophe Brachet, Thomas Guilbert, Elodie Rouesne, Denis Menut, Caroline Toffolon-Masclet, Dominique Thiaudiere

Abstract:

In some hypothetical accidental situations, such as during a Loss Of Coolant Accident (LOCA) in pressurized water reactors, fuel cladding tubes made of zirconium alloys can be exposed for a few minutes to steam at High Temperature (HT up to 1200°C) before being cooled and then quenched in water. Under LOCA-like conditions, the cladding undergoes a number of metallurgical changes (phase transformations, oxygen diffusion and growth of an oxide layer...) and is consequently submitted to internal stresses whose state evolves during the transient. These stresses can have an effect on the oxide structure and the oxidation kinetics of the material. They evolve during cooling, owing to differences between the thermal expansion coefficients of the various phases and phase transformations of the metal and the oxide. These stresses may result in the failure of the cladding during quenching, once the material is embrittled by oxidation. In order to progress in the evaluation of these internal stresses, X-ray diffraction experiments were performed in-situ under synchrotron radiation during HT oxidation and subsequent cooling on Zircaloy-4 sheet samples. First, structural evolutions, such as phase transformations, have been studied as a function of temperature for both the oxide layer and the metallic substrate. Then, internal stresses generated within the material oxidized at temperatures between 700 and 900°C have been evaluated thanks to the 2θ diffraction peak position shift measured during the in-situ experiments. Electron backscatter diffraction (EBSD) analysis was performed on the samples after cooling in order to characterize their crystallographic texture. Furthermore, macroscopic strains induced by oxidation in the conditions investigated during the in-situ X-ray diffraction experiments were measured in-situ in a dilatometer.

Keywords: APRP, stains measurements, synchrotron diffraction, zirconium allows

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