2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

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1 PROGRESS OF DEVELOPING ODS MO ALLOY FOR ACCIDENT TOLERANT FUEL CLADDING AT CGN Xing Gong 1, Sigong Li 1, Rui Li 1, Jun Yan 1, Jiaxiang Xue 1, Qisen Ren 1, Tong Liu*,1, Geng An 2, Yuanjun Sun 2 1 Department of ATF R&D, Nuclear Fuel Research and Development Center, China Nuclear Power Technology Research Institute Co., Ltd., China General Nuclear Power Corporation (CGN), Shenzhen, , PR China 2 JINDUICHENG MOLYBDENUM CO.,LTD. Xi an, , PR China (*Corresponding author. address: liutong@cgnpc.com.cn) ABSTRACT: China General Nuclear Power Corporation (CGN) is leading the China accident tolerant fuels (ATF) research program. This program is aimed to develop various innovative cladding materials and fuels for further enhancing accident tolerance of light water reactors (LWRs). In this paper, the progress of developing a nano structured (NS) and oxide dispersion strengthened (ODS) Mo alloy for ATF cladding as well as experimental assessment of its non-irradiated properties is presented. So far, thin-walled tubes have been successfully manufactured at CGN. FeCrAl alloy was adopted as a thin coating to protect the tube surface from high temperature steam oxidation. Experimental assessment on the NS ODS Mo alloy tubes has been performed to test their mechanical properties, autoclave corrosion resistance and performance under simulated LOCA conditions. The results show that the NS ODS Mo alloy maintained sufficient strength even at 1200 and the coating showed good resistance to water corrosion and oxidation. Jointing of tube to endplugs was demonstrated not to be a major issue. KEYWORDS: ATF; ODS Mo; Corrosion; Oxidation; Welding I. INTRODUCTION The nuclear disaster occurred at Fukushima Japan in 2011 has strongly aroused public concerns about nuclear safety. In this context, the accident tolerant fuels (ATF) concept was initiated to further extend safety margins and strengthen nuclear safety of light water reactors (LWRs) under accidents beyond design basis (BDB), and even more importantly to rebuild the public confidence to nuclear energy [1-3]. To date, the major nuclear countries such as the United States, France, UK, China, South Korea, etc. have initiated definitive research programs aimed to develop the ATF technology. This technology involves development of a wide range of promising new cladding materials featuring much better oxidation resistance and other unique advantages compared to the conventional Zircaloy cladding being served in current commercial LWRs. Molybdenum (and its alloys) is one of the new cladding candidates targeted by the ATF research programs, due to its excellent high temperature mechanical properties. This advantage is critical for maintaining core cooling capability as the core temperature under BDB could rise up to 1500 where not too many structural materials could survive. EPRI is a pioneer in developing Mo and oxide dispersion strengthened (ODS) Mo for ATF cladding applications and their materials have been sent to HALDEN reactor for irradiation verification [4-6]. China General Nuclear Power Corporation (CGN) is leading the China accident tolerant fuels (ATF) research program. Under the framework of this program, CGN has also been developing a FeCrAl coated NS ODS Mo cladding. In this paper, the latest progress was reported. The results obtained so far show that the NS ODS Mo cladding is promising for accident tolerance purpose. II. EXPERIMENTAL 1

2 The NS ODS Mo alloy has a nominal chemical composition of Mo-0.1~5%La (wt.%). Its fabrication process has been reported in [7]. This material was then manufactured into thin-walled tubes. The tubes were coated with a thin FeCrAl coating to overcome the drawback that Mo has poor resistance to oxidation at temperature higher than 800. The microstructure of the tubes was examined under scanning electron microscope (SEM) and transmission electron microscope (TEM). A series of experimental tests were performed to characterize the properties of the NS ODS Mo tubes. Tensile mechanical properties were tested at room temperature up to Water corrosion tests were conducted in autoclave for 72h to study the compatibility between the FeCrAl coating with static PWR water. LOCA simulation tests involving first 8 h oxidation at 1200 and subsequent water quenching were also conducted to assess if the coated NS ODS Mo cladding tubes could survive under this extreme condition. III. Results and discussion III.A. As-received microstructure TEM bright-field images in Fig. 1 show that the grains in the NS ODS Mo tubes have been heavily elongated into a fiber-like structure. The added La has been transformed into nanoscale particles which seem to mostly locate at grain interiors. Dislocation tangles can be observed as well. Appropriate heat treatment could be used to tailor the microstructure to obtain optimum mechanical properties, which is being investigated. III.B. FeCrAl coating Fig. 1. TEM bright-field images of the NS ODS Mo tubes The FeCrAl coating is dense and has a thickness of about 40µm (Fig. 2a). The scratch test shows absence of cracks formed in the coating, indicating good interfacial bonding between the FeCrAl coating and the Mo substrate. The coating adhesion was quantitatively determined at a level of 38.6MPa. 2

3 Fig. 2. Morphology of the FeCrAl coating prepared by the magnetron sputtering technique (a) and scratch test (b) III.C. Mechanical properties The tensile mechanical properties of the NS ODS Mo alloy tubes have been tested in air at room temperature, 400 and The results in Fig. 3 show that the NS ODS Mo maintained sufficient strength at 1200, while Zircaloy-4 fully lost its strength at this temperature. This advantage could enable sufficient cooling capability of reactor core under accidental conditions, if the NS ODS Mo alloy is adopted as fuel cladding. The elongation of the NS ODS Mo is comparatively low but stays at acceptable levels over the tested temperature range. Fig. 3. Comparison of mechanical properties of (ODS) Mo and Zircaloy-4 alloy (left: ultimate tensile strength; right: elongation) 3

4 SEM fractographic images of a tube tested at room temperature show intergranular cracking at boundaries of the elongated and fiber-like grains (Fig. 4). Intergranular cracking is a typical failure mode for Mo and its alloys, indicating weak grain boundary cohesion which is likely attributed to grain boundary segregation of deleterious impurities like O, N, etc [8]. Fig. 4. SEM fractography of the NS ODS Mo tube after tensile testing at RT. III.D. Water corrosion testing and LOCA simulation testing The compatibility of the FeCrAl coated NS ODS Mo alloy has been tested either with a PWR water environment to simulate a normal operating condition or with a high temperature steam environment to simulate a LOCA condition. The results in Fig. 5 show that the coating was not subjected to severe corrosion damages after exposed to static PWR water at 360 and 18.6MPa for 72h. The uniform corrosion rate was slower than Zircaloy exposed to the same condition. Fig. 5. Weight gain as a function of exposure time for the FeCrAl coated NS ODS Mo alloy and Zircaloy exposed to a static PWR water environment No obvious mass changes and exfoliations have occurred to the coating after high temperature steam oxidation at 1200 for 8h immediately followed by room temperature water quenching, even though microcracks were observed in the coating likely due to quenching induced thermal stresses (Figs. 6,7). This indicates that the FeCrAl coating is adherent highly to the Mo substrate under this testing condition. In the future, the corrosion resistance will be further tested under longer duration of exposure to the PWR water. 4

5 Fig. 6. Appearances of the FeCrAl coated ODS Mo tubes/rods before and after autoclave testing and high temperature steam oxidation followed by water quenching testing. III.E. Welding Fig. 7. Morphology of the coating after high temperature oxidation and water quenching treatment. Recently, preliminary success has been achieved after many trials on welding of tube to endplugs. The weldment has passed 20 even up to 50 MPa water-pressurized testing at room temperature. The tensile strength of the weldment reached 90% of the base metal (Fig. 8). These results demonstrate that welding of the NS ODS Mo may not be a major issue. Fig. 8. Tensile stress-displacement curves of the weldment. 5

6 III.F Neutron irradiation NS ODS Mo specimens in bulk or/and tube form are being prepared for in-pile neutron irradiation tests, with an emphasis on the effect of irradiation embrittlement on mechanical properties and coating stability. The irradiation tests will start by the end of this year. In order to mitigate potential irradiation embrittlement, microstructural tailoring is being investigated to strengthen grain boundary cohesion and ductilize the Mo matrix of the NS ODS Mo alloy. IV. SUMMARY In this paper, the progress of developing FeCrAl coated NS ODS Mo cladding as well as experimental evaluation is presented. So far, thin-walled NS ODS Mo cladding tubes have been successfully manufactured. The next step is to further thin the thickness to improve neutron economy. The coating has good interfacial bonding strength with the Mo substrate. The out-of-pile experimental results demonstrate that the coated NS ODS Mo cladding can survive the simulated extreme environmental conditions and in this sense it should be a promising material for ATF cladding. Welding of plug to tube is not a big issue. In-pile neutron irradiation tests, however, are necessary to further qualify this material and this work is underway. ACKNOWLEDGEMENTS The authors are grateful to the financial support from China National Science and Technology Major Special Project Research on Accident Tolerant Fuels Key Technologies (Grant No. R-2015ZBNFC003) and China CGN Strategic Special Project for Scientific and Technological Innovation Fundamental Research on Accident Tolerant Fuels (R- 2015ZBNFC001). REFERENCES 1. S.J. Zinkle, K.A. Terrani, J.C. Gehin, L.J. Ott, L.L. Snead, J. Nucl. Mater. 448 (2014) S. Briggs-Sitton, Nucl. News (March) (2014) X. Gong, R. Li, M.Z. Sun, Q.S. Ren, T. Liu, M.P. Short, J. Nucl. Mater. 482 (2016) Bo Cheng, P. Chou, Y-J. Kim, Nucl. Sci. Tech (5): B. Cheng, Y-J. Kim, P. Chou, Nucl. Eng. Tech. 48 (2016) Cheng, B., Program on Technology Innovation: Coated Molybdenum-Alloy Cladding for Accident-Tolerant Fuel Progress Report. 2015, EPRI. 7. G.Liu et al., Nat. Mater. 12(2013) A.V. Krajnikov, A.S. Drachinskiy, V.N. Slyunyaev, Refractory Metals & Hard Materials, 11(1992)