The SPIZWURZ bundle test was carried out at the LICAS facility at Karlsruhe Institute of Technology. The test involved a bundle of 21 unirradiated hydrogenated cladding tubes (Zircaloy-4, opt. ZIRLOTM and Duplex DX-D4) which were subjected to an internal gas pressure and slow cooling for 250 days simulating the dry storage conditions of spent nuclear fuel. The following hydride morphology metrics were measured after the test: radial hydride fraction; hydride continuity coefficient; hydride length (mean and maximum, and their statistical distributions and densities); and number density. Hydrides in Zircaloy-4 have a higher tendency to reorientation, higher connectivity and are 2-3 times shorter (in both maximal and mean length) than those in opt. ZIRLOTM. The evolution of hydride morphology in DX-D4 is mainly characterized by diffusional redistribution of hydrides from the substrate to the liner and the liner/substrate interface. Several samples exhibited local anomalies in hydride morphology, which can cause significant local embrittlement. The main experimental results are in good agreement with earlier published data. The supplementary materials include records of all the thermocouples and pressure sensors, which can be used to simulate the SPIZWURZ bundle test in detail.

In Light Water Reactors, fuel rod cladding tubes made of zirconium alloys undergo an oxidation process where a certain fraction of hydrogen may penetrate into the cladding. Hydrogen precipitates as hydride in the metallic matrix during cooling in the post-operation phase. An applied tensile force in the elastic range increases the hydrogen solubility and diffusivity and thus may initialise a reorientation process of hydrides that could be destructive for the metal. During long-term dry storage conditions, the stress influence in the elastic range on the zirconium claddings seems to be relevant, but not quantified yet. This paper presents in-situ neutron radiography experiments of hydrogenated zirconium tensile samples under the influence of different elastic tensile stresses. The experiments were conducted during the commissioning of the In-situ Neutron radiography CHAmber for tests under MEchanical Load facility, a tensile testing machine with an inductive heating system. The stress influence on hydrogen movements in the elastic range for short time frames (hours) was shown to be negligible. Only under the influence of plastic tensile stresses, hydrogen movements from regions with lower to higher stress were observed. Nevertheless, the first results show that the new facility can be successfully used for investigations of the influence of elastic stresses on the hydrogen diffusion and solubility in zirconium. However, the duration of the experiments and/or the resolution during the neutron imaging have to be optimised.

When hydrogen penetrates into cladding tubes during operation of nuclear power plants, it may affect their mechanical stability. During the dry storage of spent nuclear fuel, the initially dissolved hydrogen precipitates as hydrides in the metallic matrix. This precipitation process induces a certain stress in the zirconium structure of the cladding tube. In the framework of the SPIZWURZ project, cladding tubes are charged with hydrogen and exposed to a certain pressure - either internal or external, in order to investigate the stress influence on the hydrogen diffusion and solubility under dry storage conditions. The relevant long-term bundle test that simulated dry storage conditions with a constant cooling period of 250 days has recently ended. As described in the paper, the preliminary calculations for the temperature development and the hydride state shall be compared to the real state of the bundle after the test. Therefore, various analytical methods will help to determine relevant parameters that will be used for the fuel rod performance code TESPA-ROD from the Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS), which shall be revalidated in this framework.

Accident tolerant fuel (ATF) cladding is a new type of nuclear fuel cladding designed to improve the safety and performance of nuclear reactors. In this paper, the kinetics and degradation mechanisms during high-temperature oxidation in steam of the three most promising ATF cladding materials, i.e., chromium-coated zirconium alloys, FeCrAl alloys, and silicon carbide-based composites, are described. Each system has its own degradation mechanisms leading to different maximum survival temperatures. After providing general information and data to understand the oxidation and degradation processes, illustrative examples obtained at the Karlsruhe Institute of Technology are given for each type of cladding. The maximum temperatures at which the barrier effect of the cladding can be maintained for a reasonable period of time during nuclear accident scenarios are 1200–1300 °C for Cr-coated Zr alloys, 1400 °C for FeCrAl alloys, and 1700 °C for SiC-based composite claddings.

Neutrons interact with the magnetic moment of the atomic shell of an atom, as is common for X-rays, but mainly they interact directly with the nucleus. Therefore, the atomic number and the related number of electrons does not play a role in the strength of an interaction. Instead, hydrogen that is nearly invisible for X-rays has a higher attenuation for neutrons than most of the metals, e.g., zirconium, and thus would be visible through dark contrast in neutron images. Consequently, neutron imaging is a precise, non-destructive method to quantify the amount of hydrogen in materials with low attenuation. Because nuclear fuel cladding tubes of light water reactors are made of zirconium (98%), the hydrogen amount and distribution in metallic claddings can be investigated. Even hydrogen concentrations smaller than 10 wt.ppm can be determined locally with a spatial resolution of less than 10 μm (with a high-resolution neutron microscope). All in all, neutron imaging is a very fast and precise method for several applications. This article explains the basics of neutron imaging and provides samples of investigation possibilities, e.g., for hydrogen in zirconium alloy cladding tubes or in situ investigations of hydrogen diffusion in metals.

In order to investigate the occurring processes during long-term dry storage of spent fuel assemblies, a joined project called SPIZWURZ, between the Karlsruhe Institute of Technology and the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), was started. Aim of the SPIZWURZ project is the determination and quantification of the influence of texture and elastic strain on diffusion and solubility of hydrogen in three different zirconium alloys used in western Europe during a long-term cooling transient (1 K/d) starting at 400 °C. The strain in the cladding of an irradiated spent fuel rod shall be measured. Models predicting the formation of radial oriented hydrides will be validated, improved, and implemented in the GRS fuel rod performance code TESPA-ROD. This paper describes the SPIZWURZ project and already obtained first results.

Abstract. During operation in light-water reactors, Zircaloy cladding tubes take up hydrogen that precipitates under specific temperature–pressure conditions after operation in the form of zirconium hydrides. These zirconium hydrides have a detrimental effect on the mechanical stability of the cladding tubes. The conditions for their formation and their orientation need to be forecasted in order to determine that the spent nuclear fuel remains shielded during interim dry storage. Internal and external stresses that affect the cladding tubes throughout their slow cooling process during interim dry storage may worsen and accelerate the cladding tubes' embrittlement with regard to possible reorientation processes of zirconium hydrides. Generally, hydrogen in solid solution follows gradients in temperature, concentration, and stress. Consequently, hydrogen moves from higher to lower temperatures and from lower to higher stresses due to the thermodynamically more favourable conditions. In order to investigate elastic stress conditions similar to those during interim dry storage of hydrogenated cladding tubes, a modified mobile tensile testing machine was designed to elongate zirconium samples under defined temperatures of up to 500 ∘C. During the experiments, hydrogen movements within the samples are monitored by in situ neutron radiography. Because of the very low neutron cross section of zirconium, the metal is nearly invisible for neutrons, and the contrarily behaving hydrogen that scatters neutrons strongly appears as dark contrast in neutron images. Due to the sample's different cross section profile, the tensile stress created by the tensile testing machine is amplified and thus should lead to visible hydrogen movements. This paper describes neutron radiography experiments under different temperature–time conditions with hydrogenated tensile zirconium samples that simulate stress conditions of cladding tubes during interim dry storage.

The Cr-coated zirconium alloy claddings are near-term approach for accident tolerant fuel (ATF) claddings. The performance of magnetron-sputtered Cr-coated zirconium alloys is investigated under conditions simulating severe accidents. The transient oxidation experiments were performed using a thermogravimetric analyzer (TGA) with five different heating rates (2-50 K/min) up to 1380 degrees C, which is above the Zr-Cr eutectic temperature. The weight gain of the Cr-coated specimens was found to be significantly lower than that of the uncoated reference up to approximately 1330 degrees C. However, after the occurrence of the Zr-Cr eutectic reaction and failure of the coating at 1330 similar to 1380 degrees C, the oxidation rate rapidly increased and the weight gain rate of the Cr-coated specimens was 1.5 to 10 times higher than that of the uncoated specimens. The difference was greater at higher heating rates. These results suggest that the Cr-coating has a limited effect in preventing oxidation and degradation of cladding in case of eutectic reaction. The oxidation was faster at higher heating rate, and it may even cause runaway oxidation of cladding due to the temperature rise caused by oxidation heat. These findings are important for understanding the capabilities and degradation behaviors of Cr-coated cladding under severe accidental scenarios.

The susceptibility of delayed hydride cracking (DHC) in the Zr-2.5%Nb alloy was evaluated in six microstructures produced from an extruded tube of Zr-2.5%Nb, which underwent different thermomechanical treatments, divided into two separate groups: Low temperature samples (LT) were heat-treated below the monotectoid temperature in the alpha-Zr + beta-Nb field, and included pressure tube sample of CANDU-type material obtained through two different cold deformation methods, rolling and drawing, and stress-relieved at 400 degrees C for 24 h, and heat-treated samples at 600 degrees C/4 h. High-temperature samples (HT) were heat-treated in the beta-Zr field at 900 degrees C/ 3 h, and two different cooling sequences up to room temperature.The increase in the ultimate tensile strength (UTS) and hardness due to metallurgical processing in LT materials made them more susceptible to DHC, reducing the stress intensity, KIH, from 11.8 to 8.5 MPa m0.5, together with an increase in crack propagation velocity from 1.6 10-8 to 4.5 10-8 m/s.In situ hydrogen diffusion experiments were performed at ANTARES, the cold neutron imaging facility at the FRM-2 reactor, on LT materials. These experiments demonstrated that the recrystallization treatment-induced discontinuity of the beta-Zr phase in the parent material has a significant impact on hydrogen diffusion. It results in a 35% reduction in the diffusion coefficient compared to the parent material and a decrease in the terminal solid solubility (TSS) was observed. This resulted in a slight increase in KIH (7%), an increase in the hydride incubation time, and a decrease of about 20% in the crack propagation velocity. Under identical testing conditions, HT specimens were not susceptible to DHC phenomenon.

In order to in-situ quantify the hydrogen diffusion in metals or more precisely in zircaloy cladding tubes with the influence of an applied stress field, a new device was constructed in cooperation with the company ZwickRoell - the transportable INCHAMEL facility. It is a modification of ZwickRoell’s Kappa Mini 1 kN tensile testing machine. The facility is equipped with all features of a tensile testing machine, additionally a defined temperature can be applied. The machine’s design was dedicated in detail to fulfil the requirements for the usage in facilities with neutron radiation. In this manner, neutron radiography investigations can be performed in-situ with the neutron beam passing through the sample without any disturbances by installations like beam windows, thermo-couples, furnace tubes, heater wires, clamps for strain measurements, etc.

The QUENCH-19 experiment was a first-of-its-kind full-bundle test simulating accident conditions fol-lowed by water quench on accident-tolerant fuel (ATF) cladding. A type of FeCrAl(Y) alloy, B136Y3, was developed at Oak Ridge National Laboratory and tested at the Karlsruhe Institute of Technology using Kanthal APM corner rods, a shroud, and Kanthal AF spacer grids. Testing conditions were similar to those in QUENCH-15-which tested ZIRLO cladding behavior-so that B136Y3 and ZIRLO cladding could be com-pared. QUENCH-19 consisted of an initial pre-oxidation heating followed by a transient. Then, a maximum power hold, which was not present in QUENCH-15, was executed to extend the heating period for the FeCrAl(Y) rods. Finally, a rapid water quench was executed that was similar to emergency core coolant system (ECCS) actuation. Compared with the ZIRLO rods in QUENCH-15, the bundle in QUENCH-19 re-leased significantly less H 2 (9.2 g vs. 47.6 g) and achieved a much lower maximum temperature (1455 degrees C vs. 1880 degrees C). Furthermore, no breakaway oxidation was observed in QUENCH-19. Metallographic mounts revealed that despite the symmetry of the setup, at elevations near the maximum temperature, cladding and thermocouples were heavily damaged, substantial melting and oxidation occurred, and the cladding underwent chemical interaction with the thermocouple sheaths. Additionally, the ZrO2 spacers detrimen-tally interacted with the cladding, leading to mixed oxide debris and the full destruction of some rods. Additional failure was found in certain cooler rods that may have risen due to the high thermal expan-sion coefficient of FeCrAl alloys. This paper presents an analysis of this work, which suggests that Fe-CrAl cladding can chemically survive anticipated loss-of-coolant accident events followed by rapid ECCS quench if the correct geometry and core design are present.(c) 2023 Elsevier B.V. All rights reserved.

The annealing behavior of the pre-oxidized Cr-coated Zry-4 at 1200 °C in argon is systematically investigated. A Cr2O3 scale with a thickness of ~9 μm formed on the sample surface after pre-oxidation of the Cr-coated Zry-4 in steam at 1200 °C for 30 min. During annealing in inert atmosphere, the thickness of the Cr2O3 scale decreases with the increase of the annealing time because of the reaction between the Cr2O3 scale and the outward diffused Zr at the Cr2O3/Cr interface, the decomposition of Cr2O3, and the inward diffusion of O into the Zircaloy substrate along the ZrO2 precipitates at the Cr grain boundaries. The Cr2O3 scale has completely transformed into Cr after annealing for one hour. Pores and Cr grains are observed inside the Cr2O3 scale after annealing due to the decomposition of Cr2O3. ZrO2 precipitates at Cr grain boundaries gradually diffuse outward during the annealing and finally reach the outer surface of the Cr coating. The results in this paper provide new insights into the thickness decrease mechanism and the decomposition behavior of the Cr2O3 scale.

Zirconium (Zr) alloys are widely used in nuclear power plants as fuel cladding and are susceptible to hydrogen (H) degradation. For long operational service, Zr-based components can suffer a mechanism known as Delayed Hydride Cracking (DHC) associated to an increase of the crack propagation velocity by the re-orientation and precipitation of Zr hydride. In this process, the H mobility has a great influence. In the present work, the isothermal diffusion of H in Zr-2.5%Nb specimens obtained from a CANDU pressure tube were studied at consecutive temperatures of 300°C, 350°C, 375°C and 400°C. H content and mobility were quantified by in-situ neutron imaging experiments performed on ANTARES, the cold neutron imaging facility of FRM II. The time evolution of the H concentration across the specimen allowed the determination of diffusion coefficients, and an assessment of the limitations of existing models commonly used to describe H diffusion.

Delayed Hydride Cracking (DHC) is a failure mechanism that occurs in Zr alloys under certain conditions of hydrogen concentration, temperature and stress gradient. In service, hydrogen produced by corrosion reaction can be incorporated in Zr alloys and if the solid solubility is exceeded, hydrogen precipitates as zirconium hydride. The presence of a stress concentrator, such as a crack, generates the hydrogen diffusion and precipitation to the high stress zone beginning the DHC process. In this work, in-situ DHC tests in air at 250°C were performed at ANTARES, the neutron imaging facility of the FRM-II reactor. Samples of Zr2.5%Nb produced from extruded tubes and pressure tubes were studied using a stress rig specially modified to perform DHC tests in the neutron beam. H redistribution during mechanical testing was followed in-situ by registering the changes in neutron transmission. The results were compared with the images obtained by light optical microscopy after the tests. The results highlight the capabilities of neutron imaging to analyze time-dependent H distribution during DHC crack growth.

Chromium-coated zirconium alloys are one of the promising candidates for accident-tolerant fuel cladding (ATF) tubes for light water reactors (LWRs). In this study, the high temperature oxidation and degradation of two types of Cr coatings (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on prototype rod segment samples filled with ZrO2 pellets and tightly sealed with welded end caps. Isothermal tests at 1100 and 1200 degrees C were terminated by quenching with water; transient tests were performed up to 1500-1600 degrees C until complete coating failure. The positive effect of both types of Cr coatings was observed in all tests. Pre-damaged specimens showed only locally increased oxidation of the scratched zone, but no negative effects on the adjacent Cr coating. The behavior of the two types of coatings is compared and the degradation mechanisms are discussed based on hydrogen release data, extensive metallographic post-test examinations, and the current state of global research. (C) 2021 Elsevier B.V. All rights reserved.

The oxidation mechanism and kinetics of two nuclear-grade FeCrAl alloys were investigated in steam up to 1500 degrees C by transient and isothermal oxidation tests. The slow alpha-alumina formation kinetics well matched only for the temperature range from 1000 degrees C to 1300 degrees C. Below 1000 degrees C, formation of transient alumina caused faster kinetics. In addition, an excessive Fe-rich oxide formation was observed on the inner surface due to rough surface at 600 degrees C. Above 1300 degrees C, convoluted alpha-alumina was easily spalled and caused faster kinetics. Moreover, the oxide spallation caused Cr and Al depletion and catastrophic oxidation above 1400 degrees C by the formation of Fe-rich oxide. The catastrophic oxidation caused a liquid phase in Fe-rich oxides, which significantly changed the tube segment geometry. (C) 2022 Elsevier B.V. All rights reserved.

Single-rod oxidation and quench experiments at very high temperatures in steam atmosphere were conducted with advanced, nuclear grade SiCf/SiC CMC cladding tube segments. A transient experiment was performed until severe local degradation of the sample at maximum temperature of approximately 1845 °C. The degradation was caused by complete consumption of the external CVD-SiC sealcoat, resulting in steam access to the fiber–matrix composite with less corrosion resistance. Approaching these very high temperatures was accompanied by accelerated gas release mainly of H2 and CO2, the formation of surface bubbles and white smoke. Three one-hour isothermal tests at 1700 °C in steam with final water flooding and one three-hour experiment with fast cool-down in Ar atmosphere were run under nominally identical conditions. All isothermally tested samples survived the tests without any macroscopic degradation. The mechanical performance of these quenched clad segments was not significantly affected, while maintaining a high capability to tolerate damages. Despite these harsh exposure conditions, load transfer between SiC fibers and matrix remained efficient, allowing the composites to accommodate deformation.

Journal Article Quantification of Hydrogen in Metals Applying Neutron Imaging Techniques Nikolay Kardjilov, Nikolay Kardjilov Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, 14109, Berlin, Germany Corresponding author: kardjilov@helmholtz-berlin.de Search for other works by this author on: Oxford Academic Google Scholar André Hilger, André Hilger Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, 14109, Berlin, Germany Search for other works by this author on: Oxford Academic Google Scholar Henning Markötter, Henning Markötter Bundesanstalt für Materialforschung und -Prüfung, Unter Den Eichen 87, 12205, Berlin, Germany Search for other works by this author on: Oxford Academic Google Scholar Axel Griesche, Axel Griesche Bundesanstalt für Materialforschung und -Prüfung, Unter Den Eichen 87, 12205, Berlin, Germany Search for other works by this author on: Oxford Academic Google Scholar Robin Woracek, Robin Woracek European Spallation Source ESS ERIC, SE-221 00, Lund, Sweden Search for other works by this author on: Oxford Academic Google Scholar Felix Heubner, Felix Heubner Fraunhofer Institute for Manufacturing Technology and Advanced Materials IFAM, Branch Lab Dresden, Winterbergstraße 28, 01277, Dresden, Germany; Search for other works by this author on: Oxford Academic Google Scholar Lars Röntzsch, Lars Röntzsch Fraunhofer Institute for Manufacturing Technology and Advanced Materials IFAM, Branch Lab Dresden, Winterbergstraße 28, 01277, Dresden, Germany; Search for other works by this author on: Oxford Academic Google Scholar Mirco Grosse, Mirco Grosse Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen 76344, Germany Search for other works by this author on: Oxford Academic Google Scholar Ingo Manke, Ingo Manke Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, 14109, Berlin, Germany Search for other works by this author on: Oxford Academic Google Scholar John Banhart John Banhart Helmholtz-Zentrum Berlin für Materialien und Energie GmbH, Hahn-Meitner-Platz 1, 14109, Berlin, GermanyTechnische Universität Berlin, Straße des 17. Juni 135, 10623, Berlin, Germany Search for other works by this author on: Oxford Academic Google Scholar Microscopy and Microanalysis, Volume 28, Issue S1, 1 August 2022, Page 1666, https://doi.org/10.1017/S1431927622006638 Published: 01 August 2022

The Fukushima-Daiichi accident revealed that the zirconium fuel claddings have the significant safety risk of hydrogen detonation due to the strong oxidation and hydrogen release during the design basis accidents (DBA) and beyond design basis accidents (BDBA). Therefore, research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios have been boosted in the last decade. Deposition of protective coatings on Zircaloy cladding tubes has been considered as a near-term solution of enhanced ATF cladding. Among the candidate coating materials, there is no doubt that the research progress of Cr coating is the fastest around the world because of the advantages of such type of coating: excellent good chemical stability (including oxidation resistance and hydrothermal corrosion resistance), low thermal neutrons absorption cross-section, and excellent adherent. In this paper, the oxidation, diffusion, and mechanical properties of Cr-coated Zr alloys in normal operation conditions and accident conditions of nuclear reactors are reviewed. The factors that cause the failure of the coating are analyzed, and some questions that need to be clarified and further studied are proposed.

Zirconium alloys in nuclear power plants operate in high-pressure water at temperatures between 250 and 350 degrees C. Hydrogen (or deuterium) ingress due to waterside corrosion and if the solubility is exceeded H precipitates as a brittle hydride phase. Degradation mechanisms involve the accumulation of these brittle hydrides at cold spots or crack tips, as a result of H redistribution in response to thermal and stress gradients, respectively. Knowledge of H diffusion coefficients at operating temperatures is central to evaluating the rate of hydride accumulation and crack growth velocity. We determine the diffusion coefficients of H in Zircaloy-2 and Zr-2.5%Nb rolled plates at 250 degrees C, 300 degrees C and 350 degrees C along the rolling and normal directions by neutron imaging experiments with sensitivity of 5 wt ppm H for a spatial resolution 0.04 mm x 2 mm. These values were evaluated from H concentration profiles measured at room temperature on specimens of dimensions 10 x 10 x 4 mm(3) containing a hydride layer on one face, after annealing treatments between 60 and 600 min. This allowed the identification of a transition zone of 200-300 mu m between the hydride layer and the Zr alloy material, composed by large, sparsely distributed hydrides. In Zircaloy-2 plates, no substantial differences were observed in H diffusion along different directions or metallurgical conditions, and diffusion coefficients (0.6 +/- 0.1 10(-10) m(2)/s at 300 degrees C). By contrast, in hot rolled Zr-2.5%Nb plates the diffusion along the rolling direction (5.5 +/- 0.5 x 10(-10) m(2)/s at 300 degrees C) was much faster than along the normal direction (2.5 +/- 0.7 10(-10) m(2)/s at 300 degrees C), very likely due to H diffusing along the continuous network of beta filaments. After a thermal treatment of 3 h at 860 degrees C the plate microstructure changed generating radically changed H diffusion coefficients, resulting in H diffusion being much faster along the normal direction (4.0 +/- 0.5 10(-10) m(2)/s at 300 degrees C) than along the rolling direction (1.4 +/- 0.5 10(-10) m(2)/s at 300 degrees C). (c) 2022 Elsevier B.V. All rights reserved.