Zirconium Alloys for LWR Fuel Cladding and Core Internals

2019 ◽  
pp. 247-291
Author(s):  
Suresh Yagnik ◽  
Anand Garde
Keyword(s):  
Zirconium Alloys ◽  
Fuel Cladding

scholarly journals Solid state phase transformation kinetics in Zr-base alloys

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
A. R. Massih ◽  
Lars O. Jernkvist
Keyword(s):  
Phase Transformation ◽  
Solid State ◽  
Zirconium Alloys ◽  
Excess Oxygen ◽  
Transformation Kinetics ◽  
Fuel Cladding ◽  
Reactor Accident ◽  
Phase Transformation Kinetics ◽  
Solid State Phase Transformation ◽  
Water Reactors

AbstractWe present a kinetic model for solid state phase transformation ($$\alpha \rightleftharpoons \beta$$ α ⇌ β ) of common zirconium alloys used as fuel cladding material in light water reactors. The model computes the relative amounts of $$\beta$$ β or $$\alpha$$ α phase fraction as a function of time or temperature in the alloys. The model accounts for the influence of excess oxygen (due to oxidation) and hydrogen concentration (due to hydrogen pickup) on phase transformation kinetics. Two variants of the model denoted by A and B are presented. Model A is suitable for simulation of laboratory experiments in which the heating/cooling rate is constant and is prescribed. Model B is more generic. We compare the results of our model computations, for both A and B variants, with accessible experimental data reported in the literature covering heating/cooling rates of up to 100 K/s. The results of our comparison are satisfactory, especially for model A. Our model B is intended for implementation in fuel rod behavior computer programs, applicable to a reactor accident situation, in which the Zr-based fuel cladding may go through $$\alpha \rightleftharpoons \beta$$ α ⇌ β phase transformation.

Corrosion of Zirconium Alloys Used for Nuclear Fuel Cladding

2015 ◽  
Vol 45 (1) ◽  
pp. 311-343 ◽  
Author(s):  
Arthur T. Motta ◽  
Adrien Couet ◽  
Robert J. Comstock
Keyword(s):  
Nuclear Fuel ◽  
Zirconium Alloys ◽  
Fuel Cladding ◽  
Nuclear Fuel Cladding

Performance degradation of candidate accident-tolerant cladding under corrosive environment

2017 ◽  
Vol 35 (3) ◽  
pp. 129-140 ◽  
Author(s):  
Fumihisa Nagase ◽  
Kan Sakamoto ◽  
Shinichiro Yamashita
Keyword(s):  
Nuclear Power ◽  
Fission Products ◽  
Zirconium Alloys ◽  
Performance Degradation ◽  
Lessons Learned ◽  
Light Water Reactor ◽  
Fuel Cladding ◽  
Power Station ◽  
Light Water ◽  
Fukushima Daiichi

AbstractLight-water reactor (LWR) fuel cladding shall retain the performance as the barrier for nuclear fuel materials and fission products in high-pressure and high-temperature coolant under irradiation conditions for long periods. The cladding also has to withstand temperature increase and severe loading under accidental conditions. As lessons learned from the accident at the Fukushima Daiichi nuclear power station, advanced cladding materials are being developed to enhance accident tolerance compared to conventional zirconium alloys. The present paper reviews the progress of the development and summarizes the subjects to be solved for enhanced accident-tolerant fuel cladding, focusing on performance degradation under various corrosive environmental conditions that should be considered in designing the LWR fuel.

scholarly journals Continuum model for hydrogen pickup in zirconium alloys of LWR fuel cladding

2017 ◽  
Vol 121 (13) ◽  
pp. 135101 ◽  
Author(s):  
Xing Wang ◽  
Ming-Jie Zheng ◽  
Izabela Szlufarska ◽  
Dane Morgan
Keyword(s):  
Continuum Model ◽  
Zirconium Alloys ◽  
Fuel Cladding ◽  
Hydrogen Pickup

scholarly journals Phase transformation of the corrosion layers on nuclear fuel cladding. Raman spectroscopy study.

Paliva ◽  
2021 ◽  
pp. 113-117
Author(s):  
Kryštof Frank ◽  
Ladislav Lapčák ◽  
Jan Macák
Keyword(s):  
Raman Spectroscopy ◽  
Zirconium Oxide ◽  
Nuclear Reactors ◽  
Relative Proportion ◽  
Protective Layer ◽  
Zirconium Alloys ◽  
Corrosion Layer ◽  
Spectroscopy Study ◽  
Fuel Cladding ◽  
Nuclear Fuel Cladding

The goal of this work was the phase analysis of corrosion layers on zirconium alloys. In the environment of nuclear reactors, zirconium alloys are covered with a protective layer of zirconium oxide, which occurs in two crystalline modifications - monoclinic and tetragonal. The distribution of these phases in the corrosion layer can affect the overall corrosion rate. Raman spectroscopy was used to determine the composition of the corrosion layers. The use of this method is advantageous because the monoclinic and tetragonal phases can be easily distinguished in the spectra of the corrosion layers. In total, samples of two alloys were measured. The samples were pre-exposed at 360 °C in Li+ containing water (70 mg/l Li as LiOH) . Exposure times were between 21 d and 231 d, so the series contained both pre- and post- transition samples. The relative proportion of the tetragonal phase decreases significantly after the transient. It has also been found that the corrosion layers are highly heterogeneous in terms of the distribution of crystalline modifications.

A Review of Development of Zirconium Alloys as a Fuel Cladding Material and its Oxidation Behavior at High-Temperature Steam

2020 ◽  
Vol 46 ◽  
pp. 7-14 ◽  
Author(s):  
Mamoun I.A. Sagiroun ◽  
Xin Rong Cao ◽  
Wasim M.K. Helal ◽  
John N. Njoroge
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Oxidation Behavior ◽  
Chemical Interaction ◽  
Zirconium Alloys ◽  
Fuel Cladding ◽  
Breakaway Oxidation ◽  
High Temperature Steam ◽  
Cladding Material ◽  
Zr Alloys

Currently, Zr-alloys are widely used in nuclear power reactors for fuel cladding and structural components. Many types of zr-based alloys were developed to overcome the challenges encountered in the progress of nuclear reactors (high-burnup and high-duty). Oxygen diffused into the cladding, hydrogen absorbed in the cladding (breakaway oxidation and ruptured balloons) and rapid oxidation rate are results of chemical interaction of cladding material with steam at high temperature. Zirconium alloys seem to be the most suitable for use in fuel cladding, if they can overcome the rapid oxidation at temperature higher than 1200 °C. Previous studies on the oxidation behavior for some Zr-alloys nuclear fuel cladding tubes in steam and steam–air atmospheres at high temperatures are reviewed. The oxidation behavior of zirconium-alloys is strongly affected by the chemical composition of alloys and its surface conditions.

3D Visualisation of Crack Distributions in Oxidised Zirconium Alloys by FIB-Slicing

2012 ◽  
Vol 1421 ◽  
Author(s):  
Chris R. M. Grovenor ◽  
Na Ni ◽  
Sean S. Yardley ◽  
Gareth Hughes ◽  
Sergio Lozano-Perez ◽  
...  
Keyword(s):  
Oxidation Process ◽  
Nuclear Reactors ◽  
Zirconium Alloys ◽  
Second Phase ◽  
Fuel Cladding ◽  
Metal Interface ◽  
Aqueous Corrosion ◽  
Kinetic Transition ◽  
Crack Morphology ◽  
Characteristic Variation

ABSTRACTZirconium alloys have been used as fuel cladding and structural fuel assembly components in nuclear reactors since the 1950s, and show a characteristic variation in oxidation rate and layered crack morphology during aqueous corrosion. It is common to associate the first phenomenon with the appearance of the second. We have used 3D serial sectioning to study the morphology and distribution of cracks in corroded ZIRLO samples at different stages of oxidation, and have shown that cracks nucleate and grow at all stages of the oxidation process not just at the kinetic transition. We have used this data to analyse the nucleation of cracks with reference to the shape of the oxide/metal interface and the distribution of second phase precipitates.

Control of hydrogen absorption by nickel films obtained upon magnetic spraying of zirconium alloy using the thermoEMF method

2020 ◽  
Vol 86 (8) ◽  
pp. 32-37
Author(s):  
V. V. Larionov ◽  
Xu Shupeng ◽  
V. N. Kudiyarov
Keyword(s):  
Magnetron Sputtering ◽  
Hydrogen Concentration ◽  
Hydrogen Absorption ◽  
Zirconium Alloy ◽  
Hydrogen Adsorption ◽  
Zirconium Alloys ◽  
Nickel Film ◽  
Protective Film ◽  
Nickel Films ◽  
Hydrogen Penetration

Nickel films formed on the surface of zirconium alloys are often used to protect materials against hydrogen penetration. Hydrogen adsorption on nickel is faster since the latter actively interacts with hydrogen, oxidizes and forms a protective film. The goal of the study is to develop a method providing control of hydrogen absorption by nickel films during vacuum-magnetron sputtering and hydrogenation via measuring thermoEMF. Zirconium alloy E110 was saturated from the gas phase with hydrogen at a temperature of 350°C and a pressure of 2 atm. A specialized Rainbow Spectrum unit was used for coating. It is shown that a nickel film present on the surface significantly affects the hydrogen penetration into the alloy. A coating with a thickness of more than 2 μm deposited by magnetron sputtering on the surface of a zirconium alloy with 1% Nb, almost completely protects the alloy against hydrogen penetration. The magnitude of thermoemf depends on the hydrogen concentration in the zirconium alloy and film thickness. An analysis of the hysteresis width of the thermoEMF temperature loop and a method for determining the effective activation energy of the conductivity of a hydrogenated material coated with a nickel film are presented. The results of the study can be used in assessing the hydrogen concentration and, hence, corrosion protection of the material.

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