RESEARCH AND DEVELOPMENT OF SI-CONTAINED STEEL

A new type Si-contained austenitic stainless steel has been developed based on the high-temperature reinforcement theory and a phenomenon that the addition of silicon element can enhance corrosion resistance of steel to LBE. The ratio of Ti/C and the contents of Si, Cr and Ni element were optimized based on composition of 15-15Ti. Figure 12 is typical cross-section morphologies for the Si-contained austenitic stainless steel and 15-15Ti steel exposed to oxygen-controlled LBE (1～3×10^-6 wt%) with 1 m/s flow velocity at 500 °C for 2000 h. For both types of steels, a thin and dense (Fe, Cr)3O4 layer formed in the interface between LBE and steel substrate. However, compared with that of 15-15Ti, (Fe, Cr)3O4 layer of the Si-contained austenitic stainless steel is more even and thinner, which indicates considerable improvement of the corrosion resistance was attained. This phenomenon may be because many tiny SiO2 particles which formed in (Fe, Cr)3O4 layer hinder the mass transfer between LBE and steel matrix.

FIG. 12. Typical cross-section morphologies for Si-contained austenitic stainless steel and 15-15Ti steel

ODS steel is one of the candidate structural materials for lead-based reactors because of its excellent high temperature strength, high temperature creep performance and excellent radiation resistance. A new Si-containing ODS steel for lead-based reactor has been designed based on CLAM, and the corrosion behavior has been studied in static lead with 10^-6 wt%

oxygen concentration at 550 °C. As shown in Fig. 13, after exposure for 3000 h, an oxide film with a three-layer structure was formed on the surface of Si-containing ODS steel. The outermost layer was Fe3O4, the middle layer was composed of (Fe, Cr)3O4 with spinel structure and the innermost layer was a Cr-rich and Si-rich oxide layer. As shown in Fig. 14 and Fig. 15, the thickness of the oxide film on ODS steel after corrosion reduced with the increase of Si content. It is believed that the addition of Si can promote the formation of dense Cr-rich and Si-rich oxide layer and slow down the diffusion of oxygen into the substrate, and thus improve the corrosion resistance of ODS steel.

FIG. 13. Cross-section morphology of ODS steel with different Si content after exposure for 3000h: (a) without Si, (b) with 0.3 wt% Si, (c) with 0.5 wt% Si, (d) with 1 wt% Si [18]

FIG. 14. Evolution of corrosion thickness of ODS steel with different Si content with time [18].

FIG. 15. Results of the EDS area scanning through the cross-section of the ODS containing 1 wt% Si after exposure for 3000 h [18]

6. CONCLUSIONS

A series of compatibility evaluations on structural materials has been carried out in oxygen controlled in LBE at 500 °C and 550 °C. The evaluations obtain the following important results:

(1) The oxygen content in LBE is a key factor in determining the corrosion behaviours of T91 steel and 15-15Ti steel and the dominant factor affecting the types and properties of corrosion interface products. As for T91, the oxide-scale structure changes from a three-layer magnetite/spinel/internal oxidation zone (IOZ) scale under the oxygen concentration of 10^-6 wt%, to formation of a two-layer spinel/IOZ scale under the oxygen concentrations of 10^-7 and 10^-8 wt%. As for 15-15Ti, oxidation occurred, and a two-layer magnetite/spinel scale was formed under the oxygen concentration of 10^-6 wt%.

(2) After long-term corrosion test, the whole oxide layers of T91 steel and CLAM steel from exterior to interior are magnetite/spinel/internal oxidation zone, respectively. However, the oxide scales of 15-15Ti steel and 316L steel consisted mainly of a Fe-Cr spinel layer. The growth kinetics curves of oxide layers for T91, 15-15Ti, CLAM and 316L steels follow a parabolic rule (Δx²=Kpt), and the rate constant for 15-15Ti steel is lowest.

(3) Si-contained stainless steel and ODS-9Cr steel has been developed. Compatibility evaluation revealed that the corrosion resistances of the above steels have attained considerable improvement.

ACKNOWLEDGEMENTS

This work was funded with the National Magnetic Confinement Fusion Science Program of China with Grant Nos. 2018YFE0312200, the National Natural Science Foundation of China (nos. 51401204 and51501185) and the Natural Science Foundation of Anhui Province of China (1708085QE96).

REFERENCES

[1] WU Y. Neutronics of Advanced Nuclear Systems[M]. Springer, 2019.

[2] WU Y., BAI Y., SONG Y., HUANG Q., ZHAO Z., HU L.. Development strategy and conceptual design of China Lead-based Research Reactor [J]. Annals of Nuclear Energy, 2016, 87: 511 -516.

[3] WANG M, LIAN C, LI Y, ET AL. Preliminary conceptual design of a lead–bismuth cooled small reactor (CLEAR-SR) [J]. International Journal of Hydrogen Energy, 2015, 40(44): 15132-15136.

[4] ALEMBERTI, A., CARLSSON, J., MALAMBU, E., ORDEN, A., STRUWE, D., AGOSTINI, P., MONTI, S., European lead fast reactor-ELSY[J]. Nucl. Eng. Des. 241 (2011) 3470-3480.

[5] WEISENBURGER, A., SCHROER, C., JIANU, A., HEINZEL, A., KONYS, J., STEINER, H., MULLER, G., FAZIO, C., GESSI, A., BABAYAN, S., KOBZOVA, A., MARTINELLI, L., GINESTAR, K., BALBAUD-CELERIER, F., MARTIN-MUNOZ, F.J., CRESPO, L.S., Long term corrosion on T91 and AISI1 316L steel in flowing lead alloy and corrosion protection barrier development: Experiments and models [J]. J. Nucl. Mater. 415 (2011) 260-269.

[6] LIU, J., JIANG, Z., TIAN, S., HUANG, Q., LIU, Y., Stress corrosion behavior of T91 steel in static lead–bismuth eutectic at 480 oC [J]. J. Nucl. Mater. 468 (2016) 299-304.

[7] HEINZEL, A., WEISENBURGER, A., MÜLLER, G., Corrosion behavior of austenitic steels in liquid lead bismuth containing 10−6 wt% and 10−8 wt% oxygen at 400-500 °C [J]. J. Nucl. Mater. 448 (2014) 163-171.

[8] KURATA, Y., FUTAKAWA, M., SAITO, S., Corrosion behavior of steels in liquid lead–bismuth with low oxygen concentrations [J]. J. Nucl. Mater. 373 (2008) 164-178.

[9] SCHROER, C., VOSS, Z., WEDEMEYER, O., NOVOTNY, J., KONYS, J., Oxidation of steel T91 in flowing lead bismuth eutectic (LBE) at 550 °C [J]. J. Nucl. Mater. 356 (2006) 189-197.

[10] SCHROER, C., WEDEMEYER, O., NOVOTNY, J., SKRYPNIK, A., Performance of 9% Cr steels in flowing lead-bismuth eutectic at 450 and 550 oC, and 10−6 mass% dissolved oxygen [J]. J. Konys, Nucl. Eng. Des. 280 (2014) 661-672.

[11] BARBIER, F., ET AL. (2001). "Compatibility tests of steels in flowing liquid lead–bismuth." Journal of Nuclear Materials 295(2–3): 149-156.

[12] MULLER, G., ET AL. (2004). "Behavior of steels in flowing liquid pbbi eutectic alloy at 420 -600 degrees c after 4000-7200 h." Journal of Nuclear Materials 335(2): 163-168.

[13] SCHROER, C., ET AL. (2011). "Long-term service of austenitic steel 1.4571 as a container material for flowing lead-bismuth eutectic." Journal of Nuclear Materials 418(1-3): 8-15.

[14] TIAN, S. J., ET AL. (2016). "Oxidation behavior of t91 steel in flowing oxygen -containing lead-bismuth eutectic at 500 degrees c." Materials and Corrosion-Werkstoffe Und Korrosion 67(12): 1274-1285.

[15] MÜLLER G, SCHUMACHER G, ZIMMERMANN F. Investigation on oxygen controlled liquid lead corrosion of surface treated steels [J]. J. Nucl. Mater. 2000, 278(1): 85 -95.

[16] MARTINELLI L, BALBAUD-CÉLÉRIER F, TERLAIN A, ET AL. Oxidation mechanism of an Fe–9Cr–1Mo steel

by liquid Pb–Bi eutectic alloy at 470°C (Part II) [J]. Corrosion Science, 2008, 50(9): 2537-2548.

[17] SHUJIAN T., 2016. Corrosion Behavior and Mechanism of T91 and 15-15Ti Steels in Liquid Lead-Bismuth Eutectic under Oxygen Control at 500 ℃. PhD Thesis, University of Science and Technology of China.

[18] LIANGLIANG S., 2018. Design and Property Analyses on Silicon-containing 9Cr-ODS Steels. PhD Thesis, University of Science and Technology of China.

DEVELOPMENT OF ALUMINA FORMING MATERIALS FOR CORROSION MITIGATION IN HEAVY LIQUID METAL COOLED NUCLEAR REACTORS

Alfons Weisenburger

Karlsruhe Institute of Technology, Institute for Pulsed Power and Microwave Technology

Herman-von-Helmholtz-Platz1, 76344 Eggenstein-Leopoldshafen, Germany Email: Alfons.weisenburger@kit.edu

Adrian Jianu

Karlsruhe Institute of Technology, Institute for Pulsed Power and Microwave Technology

Herman-von-Helmholtz-Platz1, 76344 Eggenstein-Leopoldshafen, Germany Annette Heinzel

Karlsruhe Institute of Technology, Institute for Pulsed Power and Microwave Technology

Herman-von-Helmholtz-Platz1, 76344 Eggenstein-Leopoldshafen, Germany Hao Shi

Karlsruhe Institute of Technology, Institute for Pulsed Power and Microwave Technology

Herman-von-Helmholtz-Platz1, 76344 Eggenstein-Leopoldshafen, Germany Sabine Schlabach

Karlsruhe Institute of Technology, Institute for Applied Materials

Herman-von-Helmholtz-Platz1, 76344 Eggenstein-Leopoldshafen, Germany Vinga Szabo

Karlsruhe Institute of Technology, Institute for Applied Materials

Herman-von-Helmholtz-Platz1, 76344 Eggenstein-Leopoldshafen, Germany Georg Müller

Karlsruhe Institute of Technology, Institute for Pulsed Power and Microwave Technology

Herman-von-Helmholtz-Platz1, 76344 Eggenstein-Leopoldshafen, Germany Abstract

Liquid Pb and PbBi are promising coolants for nuclear reactors operating in fast neutron spectrum.

Beside its excellent neutronic properties, Pb and PbBi offer improved safety features. But, deficient compatibility of structural materials with the heavy liquid metals (HLM) especially at higher temperatures is a major challenge to be solved for the use of Pb and PbBi as coolant or heat storage material.

The major driving force for corrosion attack is the solubility of steel alloying elements like Ni, Fe and Cr in the HLM. Addition of oxygen to the HLM to provide in-situ formation of protective oxide scales is the mitigation mechanism of choice, but in a limited operating temperature window of up to 500°C. To enlarge the operation window of HLM cooled nuclear reactors advanced mitigation measures are under investigation since several years. Alloying of strong oxide formers like Al into steel surfaces can shift the operating temperature window to 600°C and even higher. FeCrAlY surface layers and bulk

materials are one further option to increase the corrosion resistance in oxygen-containing liquid metals.

Both methods rely on the in-situ oxidation capability of such material. Recently, new classes of bulk materials, alumina-forming austenitic steels (AFA) and alumina forming high entropy alloys (HEA), are investigated. The paper will briefly discuss the involved corrosion (dissolution and oxidation) mechanisms and will then focus on the advanced mitigation strategies that rely on in-situ alumina formation (surface alloying and mainly the new advanced alumina forming bulk materials (HEA and AFA) required for reliable long term operation at higher temperatures for HLM cooled fast reactors.

1. INTRODUCTION

To limit global warming, a transition to low-carbon economy by reducing the CO2 footprint of electricity production is required. However, most of the renewable energy sources have no base-load capability and are intermittent, like wind and PV. Therefore, as a base base-load capable energy source with low CO2 footprint, nuclear energy, which contributes 11% of the world’s energy supply today will be part of the energy mix in the future [1-2].

The research and design of the new nuclear reactor generation (“Generation IV International Forum”) targets highly improved safety standards in combination with higher temperatures and advanced heat transfer fluids. According to the “Generation IV International Forum”, six new types of reactors, including gas-cooled fast reactor, lead cooled fast reactor, molten salts reactor, sodium-cooled fast reactor, supercritical water-cooled reactor (SCWR), and very high-temperature gas reactor (VHTR) are selected as the next generation reactor designs [3 -4]. These systems often involve high temperature, mechanical loads and high irradiation doses (500-1000

°C, neutron dose: 10-150 dpa) [5]. The compatibility between the structural materials (e.g.

container, pipes, heat exchanger, cladding materials) and these aggressive mediums, in particular the high corrosiveness at elevated temperatures, challenge their commercial applications [6-11]. Therefore, developing advanced materials is essential to improve the high temperature compatibility between structural materials and aggressive heat transfer fluids.

Alumina-forming alloys (e.g. Ni-based, Fe-based) have received a great attention for high temperature applications because of their excellent oxidation resistance and improved mechanical properties [12-15]. By adding appropriate concentrations of Al and Cr, the alloys can form protective alumina-rich oxide scale in oxygen containing environmental conditions.

Ferritic FeCrAl alloys have been successfully developed for use in liquid Pb environment.

However, ferritic steels as well as ferritic FeCrAl alloys suffer from liquid metal embrittlement in liquid Pb [16-17], while AFA not. The high entropy alloy concept, which was first reported in 2004, has attracted attention to many research groups [18-19]. These alloys contain five or more principle elements in equal or near-equal atomic ratio to form solid solutions. Some of these combinations show superior mechanical, physical properties in comparison with the traditional materials. Recently, the high entropy alloys are proposed as candidates for high temperature structural materials considering their excellent corrosion resistance, structural stability and mechanical properties [20-21]. This paper will give an overview on the recent work at Karlsruhe Institute of Technology to develop corrosion resistant materials for liquid Pb cooled nuclear reactors [22].