SESSION II: CORROSION MITIGATION MEASURES: COATING, NEW
4. SUMMARY AND CONCLUSIONS
PLD‑grown aluminium oxide coatings hold great promise to enable the development of structural materials for LFRs. Basic characterization of the material showed high compactness and optimal adhesion to the substrate, combined with exceptional metal‑like mechanical properties and high hardness. The coatings underwent several tests in order to be qualified for their application in the nuclear field. Different heavy ions irradiation campaigns were undertaken to simulate the effect of neutrons, showing a high tolerance towards radiation‑induced damages of the films. In addition, the ability of the coatings to act as an anti‑permeation barrier was evaluated. The calculated PRF value was in the order of 104 at 550
°C, which was also confirmed irradiating the samples with high energy electrons during the permeation tests, and after several days of thermal cycling. These values will guarantee the suppression of the permeation of tritium through the structural steel. Finally, the chemical stability of the coatings against liquid lead were evaluated by designing corrosion tests with
different concentration of oxygen. No delamination and no reduction of the thickness of the coating were observed, meaning an optimal corrosion resistance of the material. Thanks to these peculiar and promising properties, PLD‑grown coatings could ensure the mitigation of the material related criticalities for the LFRs, proving to be a viable solution to decouple the protective and the structural means of materials. In principle, without requiring any strict control on the oxygen content in the cooling medium, coatings could translate in the opportunity of increasing the operation temperature of the reactor, hence improving the efficiency of the system.
ACKNOWLEDGEMENTS
We are grateful to Konstantza Lambrinou (SCK·CEN, Belgium) and Teresa Hernandez (CIEMAT, Spain) for the scientific discussions and the technical support. The authors acknowledge financial aid by ENEA through the AdP MiSE-ENEA projects (PAR 2014 – 2018), and the European Commission for having funded the H2020 projects “GEMMA” (grant agreement N° 755269), “TRANSAT” (N° 754586) and “IL TROVATORE” (N° 740415). This work has been carried out within the framework of the EURO-fusion Consortium and has received funding from the Euratom research and training program 2014 – 2018, under grant agreement N° 633053. The French network EMIR is acknowledged for supporting financially the ion irradiation experiments at the JANNUS (Joint Accelerators for Nanoscience and Nuclear Simulation) platforms of CEA-Saclay. This work contributes also to Sub-Program 3 (SP3) of the Joint Program on Nuclear Materials (JPNM) of the European Energy Research Alliance (EERA).
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EVALUATION OF THE HIGH-ENTROPY CR-FE-MN-NI ALLOYS COMPATIBILITY WITH A LIQUID LEAD COOLANT
V.M. VOYEVODIN
National Science Center “Kharkiv Institute of Physics and Technology”
NAS of Ukraine, Kharkiv, Ukraine E-mail: voyev@kipt.kharkov.ua V.M. FEDIRKO
Karpenko Physico-Mechanical Institute of the NAS of Ukraine Lviv, Ukraine
O.M. VELIKODNYI
National Science Center “Kharkiv Institute of Physics and Technology”
NAS of Ukraine, Kharkiv, Ukraine I.S. KUKHAR
Karpenko Physico-Mechanical Institute of the NAS of Ukraine Lviv, Ukraine
Kh.R. MELNYK
Karpenko Physico-Mechanical Institute of the NAS of Ukraine Lviv, Ukraine
I.V. KOLODIY
National Science Center “Kharkiv Institute of Physics and Technology”
NAS of Ukraine, Kharkiv, Ukraine M.A. TIKHONOVSKY
National Science Center “Kharkiv Institute of Physics and Technology”
NAS of Ukraine, Kharkiv, Ukraine Abstract
The corrosion compatibility of the three newly developed high-entropy alloys of the Cr-Fe-Mn-Ni system with liquid lead coolant has been evaluated at temperatures of 480 °C and 580 °C up to 1000 h exposure. It was established that for all alloys at temperatures up to 480 °C, the main mechanism of the surface layers damage is intergranular corrosion, which consists in etching of the grain boundaries, lead penetration into the matrix and dissolving such elements as Ni, Cr and Mn. When the temperature rises to 580 °C, a combined mechanism of corrosion damage is realized. This mechanism consists in the combination of intergranular corrosion and diffusion of oxygen from the melt into the matrix, which results in internal oxidation of Cr and Mn. The composition of the alloy, which has a minimal corrosion tendency and can be recommended for a usage in contact with a liquid lead up to 500 °C, was established.
1. INTRODUCTION
Currently, oxide dispersion-strengthened ferrite-martensitic and austenitic steels (so called ODS steels) are considered as promising materials for the new generation nuclear reactors [1-6]. These steels have high mechanical properties and can be capable up to damaging doses of 180 dpa (displacement per atom). But the technology of ODS steel prod uction is complicated and expensive. Therefore, the actual problem is the development of a new class of materials with much higher radiation resistance and improved values of mechanical and corrosion characteristics, which can be produced by conventional processing route. Existing theoretical studies and preliminary experimental results [7-10] show that so-called high-entropy alloys (HEAs) are promising candidates for a new generation of radiation-resistant materials due to its peculiarities of the crystalline structure.
Usually HEAs contain four or five elements with equal or close concentrations [11 -13]. The entropy of mixing increases significantly with a large number of elements in the alloys and their high concentration, which contributes to the formation of the disordered substitutional solid solutions [11]. HEAs are characterized by a strong lattice distortion caused by the difference in atomic radii of the dissolved elements, which affects the solid solution hardening, and also slow diffusion [13]. Therefore, these factors determine attractive mechanical properties, which are demonstrated by some HEA compositions [13-17].
Liquid heavy metals (Pb, Pb-Bi) are considered as a coolant in the next generation nuclear power plants (NPP) [18-20]. At the same time, in the “Accelerator Driven Systems” (ADS) hybrid systems liquid lead is an environment for nuclear reactions and adiabatic deceleration of neutrons [21]. In this regard, the specific requirement for structural materials of modern nuclear power plants is their compatibility with liquid metals [22-23]. As a rule, upon the contact of a compositionally complex alloy with a liquid metal a selective dissolution of its components is observed. The thermodynamic activity of the alloying component is determined by its propensity to selective dissolution in the liquid metal [24]. Due to the selective dissolution of some of the alloying components a layer, depleted on these elements, is forming on the surface of a solid metal that facilitates phase transformation. For example, austenite in chromium-nickel steels can transform into ferrite upon the contact with Pb, Pb-Bi melts [24].
In this regard, the purpose of the work is to estimate the compatibility of the developed new class of HEAs austenitic matrix with coolant, based on lead melts.