scholarly journals In Situ Neutron Radiography Investigations of Hydrogen Related Processes in Zirconium Alloys

2021 ◽  
Vol 11 (13) ◽  
pp. 5775
Author(s):  
Mirco Grosse ◽  
Burkhardt Schillinger ◽  
Anders Kaestner
Keyword(s):  
Zirconium Alloy ◽  
High Temperature Oxidation ◽  
Hydrogen Diffusion ◽  
Neutron Radiography ◽  
Zirconium Alloys ◽  
Hydrogen Uptake ◽  
Temperature Oxidation ◽  
Diffusive Processes ◽  
Kinetics Of

In situ neutron radiography experiments can provide information about diffusive processes and the kinetics of chemical reactions. The paper discusses requirements for such investigations. As examples of the zirconium alloy Zircaloy-4, the hydrogen diffusion, the hydrogen uptake during high-temperature oxidation in steam, and the reaction in nitrogen/steam and air/steam atmospheres, results of in situ neutron radiography investigations are reviewed, and their benefit is discussed.

scholarly journals H2 PERMEATION BEHAVIOR OF Cr2AlC AND Ti2AlC MAX PHASE COATED ZIRCALOY-4 BY NEUTRON RADIOGRAPHY

2018 ◽  
Vol 58 (1) ◽  
pp. 69 ◽  
Author(s):  
Chongchong Tang ◽  
Mirco Karl Grosse ◽  
Pavel Trtik ◽  
Martin Steinbrück ◽  
Michael Stüber ◽  
...  
Keyword(s):  
Hydrogen Diffusion ◽  
Hydrogen Permeation ◽  
Neutron Radiography ◽  
Zirconium Alloys ◽  
Hydrogen Uptake ◽  
Hydrogen Release ◽  
Max Phase ◽  
Normal Operation ◽  
Temperature Oxidation ◽  
Design Basis

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.

Ex-situ and in-situ Neutron Radiography Investigations of the Hydrogen Uptake of Nuclear Fuel Cladding Materials During Steam Oxidation at 1000°C and Above

2010 ◽  
Vol 1262 ◽  
Author(s):  
Mirco Grosse ◽  
Marius van den Berg ◽  
Eberhard Helmar Lehmann ◽  
Burkhard Schillinger
Keyword(s):  
Hydrogen Diffusion ◽  
Neutron Radiography ◽  
Zirconium Alloys ◽  
Hydrogen Uptake ◽  
Steam Oxidation ◽  
Ex Situ ◽  
In Situ Experiments ◽  
Nuclear Fuel Cladding ◽  
Non Destructive

AbstractNeutron radiography is a powerful tool for the investigation of the hydrogen uptake of zirconium alloys. It is fast, fully quantitative, non-destructive and provides a spatial resolution of 30 μm. The non-destructive character of neutron radiography provides the possibility of in-situ investigations. The paper describes the calibration of the method and delivers results of ex-situ measurements of the hydrogen concentration distribution after steam oxidation, as well as in-situ experiments of hydrogen diffusion in β-Zr and in-situ investigations of the hydrogen uptake during steam oxidation.

scholarly journals HIGH-TEMPERATURE OXIDATION OF CHROME-NICKEL ALLOY

2020 ◽  
Vol 28 ◽  
pp. 8-14
Author(s):  
Adéla Chalupová ◽  
Martin Steinbrück ◽  
Mirco Grosse ◽  
Jakub Krejčí ◽  
Martin Ševeček
Keyword(s):  
High Temperature ◽  
Melting Point ◽  
High Temperature Oxidation ◽  
Zirconium Alloys ◽  
Temperature Oxidation ◽  
Ni Alloy ◽  
Institute Of Technology ◽  
Kinetics Of ◽  
Karlsruhe Institute ◽  
Nuclear Applications

The investigations in this paper deal with the Cr-Ni alloy. The material has been recently proposed as a potential ATF concept, primarily due to its behaviour under high-temperature oxidation. A set of experiments to determine the melting point and describe the oxidation kinetics of the Cr-Ni alloy were performed in Karlsruhe Institute of Technology. Presented results reveal its superb oxidation resistance comparing to zirconium alloys. Therefore, the alloy has a great potential for nuclear applications.

scholarly journals In-situ time-resolved study of structural evolutions in a zirconium alloy during high temperature oxidation and cooling

2019 ◽  
Vol 158 ◽  
pp. 109971 ◽  
Author(s):  
R. Guillou ◽  
M. Le Saux ◽  
E. Rouesne ◽  
D. Hamon ◽  
C. Toffolon-Masclet ◽  
...  
Keyword(s):  
High Temperature ◽  
Zirconium Alloy ◽  
High Temperature Oxidation ◽  
Time Resolved ◽  
Temperature Oxidation

On the problems of creating shells of fuel rods from zirconium alloys for tolerant fuel

2021 ◽  
Vol 3 ◽  
pp. 69-78
Author(s):  
A. A. Yakushkin ◽  
Keyword(s):  
High Temperature ◽  
Production Technology ◽  
Zirconium Alloy ◽  
High Temperature Oxidation ◽  
Fuel Cladding ◽  
Temperature Oxidation ◽  
Corrosion Resistant ◽  
Fuel Rod ◽  
Kinetics Of ◽  
Corrosion Resistant Coating

Three directions of the establishment of accident tolerant fuel cladding for light water reactors are actively exploring at present: 1) replacement zirconium alloy E110 for more corrosion-resistant material in accident operation conditions; 2) surface dispersion hardening or doping of the zirconium cladding of fuel element; 3) deposition a corrosion-resistant coating to the fuel cladding. The first direction requires significant and irreversible changes in fuel rod production technology and has long-term prospects. Conversely, the second direction suggest minimal changes in the fuel rod production technology, however, it has no significant effect on the high temperature oxidation kinetics of fuel claddings in steam. Using of a corrosion resistant coating results in a significant change in the high temperature oxidation kinetics of the zirconium alloy, (no transition to linear oxidation) that is related to maintaining the continuity of the oxide layer formed during oxidation. The issue provides a brief overview of the current state of research in the field of fuel, tolerant to the effects of coolant in emergency situations.

Effect of Copper Addition on the High Temperature Oxidation of Zirconium Alloy ZrNbMoGe for Advanced Reactor Fuel Cladding Material

2014 ◽  
Vol 896 ◽  
pp. 617-620 ◽  
Author(s):  
Bernardus Bandriyana ◽  
Djoko Hadi Prajitno ◽  
Arbi Dimyati
Keyword(s):  
High Temperature ◽  
Zirconium Alloy ◽  
High Temperature Oxidation ◽  
Zirconium Alloys ◽  
Magnetic Suspension ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Copper Addition ◽  
Effect Of Copper ◽  
Cladding Material

The zirconium alloys ZrNbMoGe have been developed with the aim to improve its high temperature oxidation for employment as a cladding material in Pressurized Water Reactor (PWR) and to extend the over all fuel burn-up. In this paper the effect of Cu addition on the high temperature oxidation behavior of ZrNbMoGe alloy was investigated. The zirconium alloy was produced by melting the zirconium-niobium-molybdenum-germanium and copper-sponge in an arc furnace in an argon environment by the temperature higher than 1850C. The weight percentages of the elements were 2.50 wt.% Nb, 0.5 wt.% Mo, 0.1 wt.% Ge, 0.5 wt.% Cu and Zr in balanced. The oxidation test was carried out in the Magnetic Suspension Balance (MSB) workstation. Two specimens of ZrNbMoGe alloys without and with Cu addition were oxidized in atmosphere at temperature of 500 °C and 700 °C for 8 hours. The results show the oxidation kinetics followed the parabolic rate law. The difference of oxidation behaviors of the two specimens were considered to be caused by the formation of different kind of oxide layers due to the Cu addition.

High Temperature Oxidation of Polycrystalline Diamond Coated Zirconium Alloy

2016 ◽  
Author(s):  
Jan Škarohlíd ◽  
Radek Škoda ◽  
Irena Kratochvílová
Keyword(s):  
Raman Spectroscopy ◽  
High Temperature ◽  
Zirconium Alloy ◽  
High Temperature Oxidation ◽  
Polycrystalline Diamond ◽  
Zirconium Alloys ◽  
Hydrogen Release ◽  
Primary Circuit ◽  
Temperature Oxidation ◽  
Diamond Layer

Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys and decrease corrosion rate of zirconium alloy during standard operation. Zirconium alloys are widely used as cladding and construction material in almost all types of nuclear reactors, where usually creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper zirconium alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations (Raman spectroscopy) were done for zirconium alloy covered with polycrystalline diamond layer before and after high temperature steam exposure. Weight increase and hydrogen release ware measured during steam exposure.

Sign in / Sign up

Export Citation Format

Share Document