High Temperature Oxidation of Zircaloy-4 Under Conditions Simulating a Loss of Cooling Accident in Spent Fuel Pools Examined with Raman Imaging and 18O Tracer Techniques

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.

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

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4043462 ◽

2019 ◽

Vol 5 (3) ◽

Author(s):

Petra Gávelová ◽

Patricie Halodová ◽

Hygreeva Kiran Namburi ◽

Iveta Adéla Prokůpková ◽

Marek Mikloš ◽

...

Keyword(s):

High Temperature ◽

Nuclear Power ◽

High Temperature Oxidation ◽

Nuclear Research ◽

Material Behavior ◽

Creep Testing ◽

Spent Fuel ◽

Reactor Operation ◽

Temperature Oxidation ◽

Characterization Methods

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.

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