From Micro to Nano: Material Characterization Methods for Testing of Nuclear Core and Structural Materials

Author:

Gávelová Petra1,Halodová Patricie1,Namburi Hygreeva Kiran1,Prokůpková Iveta Adéla1,Mikloš Marek1,Krejčí Jakub2

Affiliation:

1. Research Centre Řež, Hlavní 130, Husinec-Řež 250 68, Czech Republic e-mail:

2. UJP Praha, Nad Kamínkou 1345, Praha-Zbraslav 156 10, Czech Republic e-mail:

Abstract

Nuclear grade fuel claddings act as a barrier against release of fuel particles into the coolant water during nuclear power plant operation, handling, and wet or dry storage of the spent fuel rods. The integrity of claddings is always a critical issue during the reactor operation, storage, or loss of coolant accidents (LOCA). After Fukushima disaster, cladding materials are widely studied, as the neutron transparent zirconium (Zr)-based alloys or accident tolerant fuel claddings, which should reduce the high-temperature oxidation rate of Zr-based alloys and enhance the accident tolerance. One part of the long-term investigation of Zr-based claddings in the Research Center Řež contributes to the characterization of material behavior after creep testing and high-temperature oxidation, i.e., conditions that may occur during reactor operation or storage of the spent fuel. Then, the microstructure response can be investigated by means of standard as well as advanced microscopy methods. In this study, we introduce the methodology of the cladding microstructure complex characterization demonstrated on the reference (nonirradiated) Zr-1%Nb alloy after creep testing. In the Zr-1%Nb microstructure, the hydrogen is present under some conditions as the hydride phases ZrH or ZrH2 that cause the mechanical properties degradation. As well, the hydride reorientation can pose a risk. Therefore, currently used methods of hydrides evaluation are presented. Furthermore, we introduce the microstructure analysis of the zircalloy-4 cladding after high-temperature oxidation at the temperature 1300 °C simulating LOCA. For this study, the new automatic chemical analysis procedure was developed. The main issue is the determination of the oxygen concentration and alloying elements behavior after critical increase of oxygen in the Zr-alloy tube at LOCA, also concerning the hydrogen content in the microstructure. Based on the element concentration gradients, the phase transitions from the oxide ZrO2 to the prior β-Zr can be determined and contributes to the phase diagrams experimental verification. In our contribution, the significant role of the electron microscopy methods “from micro to nano-scale” in the nuclear research is emphasized in these two research topics.

Publisher

ASME International

Subject

Nuclear Energy and Engineering,Radiation

Reference6 articles.

1. Microstructural Evaluation of High Temperature Creep Behavior in Hydrogenated E110 Cladding,2017

2. Study of Creep and Hydride Re-Orientation Behavior in E110 Fuel Cladding at Dry Storage Conditions,2017

3. Microstructure and Local Mechanical Characteristics of Zr1Nb Alloy After Hardening;Chemicke Listy,2011

4. Negyesi, M., Krejčí, J., Linhart, S., Novotny, L., Přibyl, A., Burda, J., Klouček, V., Lorinčík, J., Sopoušek, J., Adámek, J., Siegl, J., and Vrtílková, V., 2014, “Contribution to the Study of the Pseudobinary Zr1Nb–O Phase Diagram and Its Application to Numerical Modeling of the High—Temperature Steam Oxidation of Zr1Nb Fuel Cladding,” 17th International Symposium on Zirconium in the Nuclear Industry, ASTM International, West Conshohocken, PA, Paper No. STP 1543.

5. Namburi, H., Chocholoušek, M., Ottazi, L., Krejčí, J., and Bublíková, P., 2018, “Study of Hydrogen Embrittlement and Hydride Re-Orientation Behaviour in Zirconium Based E110 Fuel Cladding by Ring Compression Testing,” COMAT2018 Recent Trends in Nuclear Materials, Fifth International Conference On Recent Trends In Structural Materials Proceedings, Plzeň, Czech Republic, Nov. 14–16.

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