Experimental Verification of Phase Diagram Calculations of Zr-Based Alloys after High-Temperature Oxidation

Abstract:

Zirconium-based alloys are commonly used as a material for nuclear fuel claddings in the light water reactors. The cladding material must function to fix a huge number of pellets, while conducting heat into the coolant that flows turbulently around the fuel rods. Cladding tubes can contain gaseous fission products that escape the fuel. Thus, by functioning as a sealed unit, it prevents a contamination of the coolant water with high-radioactive fission products. The integrity of claddings is always a critical issue during reactor operation and wet or dry storage and transport of the spent fuel rods. Moreover, the role gains importance at Loss of Coolant Accidents (LOCA). After Fukushima accident, cladding materials are widely studied with the purpose to reduce the high-temperature oxidation rate and enhance accident tolerance. In our contribution, we introduce the studies on Zr-1Nb (E110) cladding tubes after high-temperature steam oxidation at 1350 °C. During the testing of claddings, microscopy analytical methods play an important role in experimental verification of pseudo-binary phase diagram Zr1Nb-O, i. e. particularly in oxygen content determination at phase transitions. Wave Dispersive Spectroscopy (WDS) with complementary nano-indentation method were used to characterize the Zr1Nb microstructure formed after LOCA. It includes the regions from an oxide and oxygen-stabilized α-Zr(O) to the acicular prior β-Zr phase. The decrease of hardness and Young's modulus corresponds with oxygen content measured in line-profiles by WDS. The oxygen level at transition points was partly determined from Fe, Nb β-stabilizers and significant change in mechanical properties in fine-grained prior β-Zr. The slight fluctuation of oxygen values in adjacent grains can be caused by preferential oxidation through the favorably oriented α-Zr(O) grains studied by WDS+EBSD. As well, the non-uniform oxygen-rich α-Zr(O) phase adjacent to the oxide was characterized by EBSD & WDS. Increasing hydrogen content in specimens, 10, 700 and 1000 ppm H, caused increasing solubility of oxygen in prior β-Zr phase upon high-temperature and the cladding material hardening.

• Periodical: Defect and Diffusion Forum (Volume 405)
• Pages: 351-356
• DOI: https://doi.org/10.4028/www.scientific.net/DDF.405.351
• Online since: November 2020
• Authors: Petra Gávelová*, Patricie Halodová, Ondřej Libera, Iveta Adéla Prokůpková, Věra Vrtílková, Jakub Krejčí
• Keywords: High-Temperature Oxidation, LOCA, Nanoindentation, WDS, Zirconium-Based Alloys

References

[1] H.K. Namburi, L. Ottazzi, M. Chocholousek, J. Krejci, Study of hydrogen embrittlement and determination of E110 fuel cladding mechanical properties by ring compression testing, 2018. Internat. Conference on Materials and Metallurgy 2018. ISBN 978-80-87294-84-0.

DOI: 10.1088/1757-899x/461/1/012059

[2] H.K. Namburi, J. Šoltés, J. Krejčí, M. Mikloš, Neutron radiography studies for detection of hydrogen distribution in nuclear fuel claddings at Research Centre Řež, 2017. ICARST Proceedings Website: http://www-pub.iaea.org/iaeameetings/50814/International-Conference-on-Applications-of-RadiationScience-and-Technology-ICARST-(2017).

[3] J.-Ch. Brachet, et al., High-temperature (LOCA) steam oxidation and quenching/post-quenching behavior of Zr alloy. Lecture in Metallurgy and properties of Zr alloys for nuclear applications. Saclay. (2017).

[4] Available from www.atominfo.cz/2013/12/havarie-spojene-se-ztratou-chladiva/.

[5] M. Negyesi, J. Krejčí, S. Linhart, L. Novotný, A. Přibyl, J. Burda, V. Klouček, J. Lorinčík, J. Sopoušek, J. Adámek, J. Siegl, V. Vrtílková, Contribution to the Study of the Pseudobinary Zr1Nb–O Phase Diagram and Its Application to Numerical Modeling of the HighTemperature Steam Oxidation of Zr1Nb Fuel Cladding. 17th International Symposium: Zirconium in Nuclear Industry. STP 1543, 2014 / available online at www.astm.org /.

DOI: 10.1520/stp154320120162

[6] R.W.L. Fong et al., A Differential Scanning Calorimetry (DSC) Study of Phase Changes in an As-Received Zr-2.5Nb Pressure Tube Material during Continuous Heating and Cooling. Materials Science Forum, 706-709 (2012) 853-858.

DOI: 10.4028/www.scientific.net/msf.706-709.853

[7] J. Čížek, Phase transformations in Zr-Nb alloys – thermal analysis. Regular technical meeting ČEZ about the Zr-alloys. Vila Diamo. Barrandov. (2019).

[8] Nuclear energy agency organization for economic co-operation and development. Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions. State-of-the-art Report. In Nuclear Safety. ISBN 978-92-64-99091-3.

[9] D.V. Markov, et al., State of Fuel Rods Spent in the VVER-1000 Reactor up to a Fuel Burnup of 75 MW.day/kgU, Proceeding of 9th International Conference on WWER Fuel Performance, Modelling and Experimental Support, Helena Resort, Bulgaria, September 17-24, (2011).

[10] J. Sopoušek: Phase diagram of Zr-Nb-Fe-O-H based alloys. Masaryk University. Research report. 2015. Brno.

[11] J. Krejčí, J. Kabátová, J. Kočí, Z. Weishauptová, V. Vrtílková, Hydrogen Charging of Fuel Cladding Methodology, Acta Polytechnica CTU Proceedings 4 (2016) 38-42, available online at http://ojs.cvut.cz/ojs/index.php/app,.

DOI: 10.14311/ap.2016.4.0038

[12] C. Oliver and M. Pharr, An improved technique for determining hardness and elastic modulus using load and displacement sensing indentation experiments, Journal of Materials Research 7, 11 (1992) 1564-1583.

DOI: 10.1557/jmr.1992.1564

[13] J. Krejčí, M. Ševeček, J. Kabátová, F. Manoch, J. Kočí, L. Cvrček, J. Málek, S. Krum, P. Šutta, P. Bublíková, P. Halodová, H.K., Namburi, Experimental Behavior of chromium-based coatings. TopFuel (2018).

DOI: 10.1016/j.net.2019.08.015