0
Special Section on Research Center Řež: Nuclear-Engineering Activities in 2018

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

[+] Author and Article Information
Petra Gávelová

Research Centre Řež,
Hlavní 130,
Husinec-Řež 250 68, Czech Republic
e-mail: Petra.Gavelova@cvrez.cz

Patricie Halodová

Research Centre Řež,
Hlavní 130,
Husinec-Řež 250 68, Czech Republic
e-mail: Patricie.Halodova@cvrez.cz

Hygreeva Kiran Namburi

Research Centre Řež,
Hlavní 130,
Husinec-Řež 250 68, Czech Republic
e-mail: Hygreeva.Namburi@cvrez.cz

Iveta Adéla Prokůpková

Research Centre Řež,
Hlavní 130,
Husinec-Řež 250 68, Czech Republic
e-mail: Iveta.Prokupkova@cvrez.cz

Marek Mikloš

Research Centre Řež,
Hlavní 130,
Husinec-Řež 250 68, Czech Republic
e-mail: Marek.Miklos@cvrez.cz

Jakub Krejčí

UJP Praha,
Nad Kamínkou 1345,
Praha-Zbraslav 156 10, Czech Republic
e-mail: krejci@ujp.cz

1Corresponding author.

Manuscript received February 28, 2018; final manuscript received April 3, 2019; published online May 3, 2019. Assoc. Editor: Martin Schulc.

ASME J of Nuclear Rad Sci 5(3), 030917 (May 03, 2019) (6 pages) Paper No: NERS-18-1017; doi: 10.1115/1.4043462 History: Received February 28, 2018; Revised April 03, 2019

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.

FIGURES IN THIS ARTICLE
<>
Copyright © 2019 by ASME
Your Session has timed out. Please sign back in to continue.

References

Namburi, H. , Halodová, P. , Bublíková, P. , Janura, R. , and Krejčí, J. , 2017, “ Microstructural Evaluation of High Temperature Creep Behavior in Hydrogenated E110 Cladding,” NENE 2016: Internation Conference of Nuclear Energy for New Europe: Microstructural Evaluation of Creep Behavior in Hydrided E110 Cladding, Portorož, Slovenia, Sept. 5–8.
Namburi, H. K. , Halodová, P. , Bublíková, P. , and Jakub, K. , 2017, “ Study of Creep and Hydride Re-Orientation Behavior in E110 Fuel Cladding at Dry Storage Conditions,” In ENERGY.
Bláhová, O. , Medlín, R. , and Říha, J. , 2011, “ Microstructure and Local Mechanical Characteristics of Zr1Nb Alloy After Hardening,” Chemicke Listy, 105(14), pp. 202–205.
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.
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.
Namburi, H. , Chuan, H. H. , and Chuang, F. Y. , 2018, “ Assessment of Material Properties by Lateral Compression Testing in Nuclear Grade Fuel Claddings,” Eighth Conference and Workshop REMOO2018, Venice, Italy, May 29–31.

Figures

Grahic Jump Location
Fig. 1

Hydride distribution analysis in the circumferential direction using light optical microscopy (left image) and 3D SPM imaging (right image)

Grahic Jump Location
Fig. 2

SEM imaging of hydrides distributed in the α-Zr matrix. (a) Backscatter electrons imaging (hydrides—dark phase), (b) phase map (hydrides—red darker phase in the α-Zr), and (c) grain orientation map in Euler angles (hydrides—red darker phase in α-Zr).

Grahic Jump Location
Fig. 3

Specimen sectioning for TEM analysis from ballooned regions. 1—Maximum ballooned region, 2—intermediate ballooned region, 3—un-ballooned region, a–f: specimen-sectioning from all regions for TEM foil preparation in tangential direction

Grahic Jump Location
Fig. 4

TEM examination on Zr-1%Nb alloy in BF conditions in intermediate and maximum-ballooned regions. (a) Dislocation network (green marked region) and secondary precipitates Zr(NbFe)2 and β-Nb (red arrows), (b) dislocation network (green marked region), and (c) Zr-matrix grains with high density of dislocations and secondary precipitates in maximum-ballooned region, and (d) hydrides in α-Zr grains (red arrows).

Grahic Jump Location
Fig. 5

EBSD on TEM foils. (a) TEM foil fixed in the specimen holder, imaged in secondary electrons, SEM. (b) EBSD map in Euler angles on TEM foil prepared by electrolytic polishing for α-Zr grain size distribution analysis. (c) Misorientation map with regions of higher dislocation density (green areas). (d) Grain size distribution graph to (b).

Grahic Jump Location
Fig. 6

Element concentration profiles on zircalloy-4 cladding after high-temperature oxidation with marked phase transition points. (a) Microstructure of zircalloy-4 imaged in backscatter electrons, SEM. (b) O, Zr, Cr, Sn, Fe elemental profiles measured by WDS.

Tables

Errata

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In