Hydrogen diffusion and precipitation in duplex zirconium nuclear fuel cladding quantified by high-resolution neutron imaging (original) (raw)
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In-situ neutron radiography investigations of hydrogen diffusion and absorption in zirconium alloys
Nuclear Instruments & Methods in Physics Research Section a-Accelerators Spectrometers Detectors and Associated Equipment, 2011
In the framework of nuclear safety research the high temperature reaction behavior of the classical pressurized water reactor cladding material Zircaloy-4 in nitrogen containing wet atmospheres was investigated by means of in-situ neutron radiography. The paper describes experimental details and gives an overview of the results. The dependence of the reaction rate on the sum of oxygen þ steam flux is discussed and compared with a model developed at KIT. The results confirm the, at first glance surprising, prediction of the model that the time of the transition from parabolic to linear kinetics increases and the post-transition reaction decreases with increasing flow rate of oxygen þ steam.
Investigation of the 3D hydrogen distribution in zirconium alloys by means of neutron tomography
International Journal of Materials Research, 2020
The fuel rod claddings in nuclear light water reactors are made of zirconium alloys. Corrosion of these alloys during operation and in particular high temperature oxidation during nuclear accidents results in the production of free hydrogen. The cladding can absorb this hydrogen. It affects the mechanical properties of the cladding material. Hydrogen embrittlement of these materials provides the risk of brittle fracture of the cladding by thermo-shock during emergency cooling. At KIT the behaviour of cladding materials under different hypothetical nuclear accident scenarios was investigated. One focus was on hydrogen absorption and distribution/re-distribution in the alloys. The hydrogen distribution was determined mainly by neutron tomography. Examples for the determination of the 3D hydrogen distribution in cladding tubes after loss of coolant accident simulation tests are given and discussed.
Modeling hydrogen localization in Zircaloy cladding subjected to temperature gradients
Journal of Nuclear Materials, 2024
In light water reactors, Zr-based alloys used for nuclear fuel cladding undergo oxidation and absorb hydrogen, which can precipitate as brittle zirconium hydrides. During reactor normal operation conditions, hydrogen concentration varies locally due to various factors, where local high concentration can significantly degrade cladding mechanical properties and, hence, its service life. This study aims to quantify the redistribution of hydrogen within the fuel rods caused by temperature gradients, considering geometric irregularities such as inter-pellet regions, oxide spallation, and missing pellet surfaces. Through simulations, we have discovered that these temperature variations can lead to significant local hydrogen enrichment, even under normal operational conditions. Consequently, the Zircaloy cladding may suffer from critical weakening in its mechanical performance due to excessive hydride precipitation at specific locations. These findings underscore the importance of accounting for local hydrogen concentrations when evaluating the overall reliability of the cladding
Determination of very low concentrations of hydrogen in zirconium alloys by neutron imaging
Journal of Nuclear Materials, 2018
Zr-based alloys are used in nuclear power plants because of a unique combination of very low neutron absorption and excellent mechanical properties and corrosion resistance at operating conditions. However, Hydrogen (H) or Deuterium ingress due to waterside corrosion during operation can embrittle these materials. In particular, Zr alloys are affected by Delayed Hydride Cracking (DHC), a stress-corrosion cracking mechanism operating at very low H content (~100-300 wt ppm), which involves the diffusion of H to the crack tip. H content in Zr alloys is commonly determined by destructive techniques such as inert gas fusion and vacuum extraction. In this work, we have used neutron imaging to non-destructively quantify the spatial distribution of H in Zr alloys specimens with a resolution of ~5 wt ppm, an accuracy of ~10 wt ppm and a spatial resolution of ~25 µm x 5 mm x 10 mm. Non-destructive experiments performed on a comprehensive set of calibrated specimens of Zircaloy-2 and Zr2.5%Nb at four neutron facilities worldwide show the typical precision and repeatability of the technique. We have observed that the
H2 PERMEATION BEHAVIOR OF Cr2AlC AND Ti2AlC MAX PHASE COATED ZIRCALOY-4 BY NEUTRON RADIOGRAPHY
Acta Polytechnica, 2018
Hydrogen uptake by nuclear fuel claddings during normal operation as well as loss of coolant during design basis and severe accidents beyond design basis has a high safety relevance because hydrogen degrade the mechanical properties of the zirconium alloys applied as cladding material. Currently, claddings with enhanced accident tolerance are under development. One group of such accident tolerant fuel (ATF) claddings are zirconium alloys with surface coatings reducing corrosion and high-temperature oxidation rate, as well as the chemical heat and hydrogen release during hypothetical accidents. The hydrogen permeation through the coating is an important parameter ensuring material safety. In this work, the hydrogen permeation of Ti2AlC and Cr2AlC MAX phase coatings on Zircaloy-4 is investigated by means of neutron radiography. Both coatings are robust hydrogen diffusion barriers that effectively suppress hydrogen permeation into the matrix.
Hydrogen in zirconium alloys: A review
Journal of Nuclear Materials, 2019
Hydrogen absorbed into zirconium alloy nuclear fuel cladding as a result of the waterside corrosion reaction can affect the properties of nuclear fuel, principally through the precipitation of brittle hydride particles. Multiple phenomena are involved in this overall process: after hydrogen pickup degradation of mechanical properties is controlled by hydrogen transport, hydride precipitation and dissolution kinetics and the formation of specific mesoscale hydride microstructures. The precipitation of hydrides especially affects cladding ductility and fracture toughness, but can also affect other phenomena, including via stress-induced hydride reorientation. These processes can affect cladding performance both during normal operation and during extended dry storage, as hydride morphology can be modified during the preparatory vacuum drying processes. We review the processes of hydrogen transport, hydride precipitation and dissolution and formation of mesoscale hydride microstructures, and highlight where more research is needed, both from an experimental and from a modeling point of view.
2013
Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel.
Corrosion and Hydrogen Uptake in Zirconium Claddings Irradiated in Light Water Reactors
Zirconium in the Nuclear Industry: 17th Volume, 2014
The objective of this paper is to summarize the results of the latest observations performed at Paul Scherrer Institut on irradiated fuel claddings, to characterize their corrosion and hydrogen-uptake behavior. Two categories of studies have been performed. (1) A series of destructive tests were achieved on the fuel rods