Performance degradation of candidate accident-tolerant cladding under corrosive environment

3.1 Coated Zr alloy cladding

The main weakness of Zr alloy cladding is high-temperature oxidation and the subsequent performance degradation that negatively affects accident progression. To minimize the impact of alternative cladding materials to the current reactor core design, the coating of materials that have higher oxidation resistance is one of the best solutions. The coating material should have appropriate thermal conductivity, mechanical properties, weldability, high-temperature mechanical and chemical stabilities, and thermal expansion close to Zr.

Cr coatings aimed at protecting the current Zircaloy and M5 cladding are developed at the French Alternative Energies and Atomic Energy Commission (CEA) in partnership with AREVA and EDF. In their recent studies (Idarraga-Trujillo et al., 2013; Brachet et al., 2015, 2016; Bischoff et al., 2016; Kumar et al., 2016), waterside corrosion has been investigated with coated plate samples at 360°C in autoclave under a representative PWR coolant conditions for about 60 days and at 415°C under 100 bar steam for about 200 days. Corrosion kinetics was significantly reduced compared to uncoated reference materials as shown in Figure 3. The chromium oxide (Cr₂O₃) formed at the outer surface is thinner than 250 nm after the test at 415°C. Because zirconium oxide (ZrO₂) was not observed beneath the coated layer, it was confirmed that the Cr-coated layer kept the protectiveness against oxygen diffusion under examined conditions. Samples with preexisting defects in the coating were also subjected to corrosion tests at 360°C (and oxidation tests at 1100°C). Limited and localized oxidation was observed, which indicated the very small impact of defects on the corrosion resistance and good adhesion to the substrate. The Cr-coated samples were also subjected to steam at higher temperatures assuming DBA-LOCA, and it was shown that weight gain was very small and the Zircaloy substrate was protected from oxygen diffusion up to 1500 s at 1200°C. To confirm the postquench ductility, which is the key performance for safety in a DBA-LOCA, Cr-coated samples were quenched after oxidation (at 1000°C for 15,000 s) and then tested by the ring compression method. As expected from oxidation tests, reduction in strength and ductility were very small after oxidation and quench. Some α-incursions, which are the trace of oxygen diffusion, were observed in the Zircaloy substrate after oxidation in helium-steam mixture at 1300°C for 5600 s, although the integrity of the specimens was maintained by the remaining metallic part of the Zircaloy substrate. For information on the in-pile behavior of the Cr-coated cladding, irradiation experiments of the fuel rod geometry with UO₂ fuel pellets are planned under a representative PWR environment in the Halden reactor.

Figure 3:

Weight gain of uncoated and Cr-coated Zircaloy-4 and M5 after autoclave tests at 415°C under 100 bar steam environment (Brachet et al., 2015; Bischoff et al., 2016).

Xu et al. (2016) conducted oxidation tests of ZIRLO cladding on which Ti₂AlC was coated by cold-spray deposition to examine the integrity under normal operations and DBA conditions in the Westinghouse-led ATF program. Results of corrosion tests at 360°C (150 bar) and 427°C (103 bar) showed improved corrosion resistance of the dense Ti₂AlC-coated cladding, which was formed by the optimization of particle size and cold-spray deposition parameters. The higher corrosion resistance was achieved by the formation of titanium oxide (TiO₂) instead of aluminum oxide (Al₂O₃) on the surface of the coating. On the contrary, it was reported that resistance to steam oxidation at 1200°C was attributed to the formation of Al₂O₃ on the surface. It was noted that oxidation resistance would be improved by increasing the purity of TiAlC powder. The high corrosion resistance at 360°C of the ZIRLO cladding with the TiAlN/TiN multilayer coating was also reported in their study. The TiN layer is expected to have resistance to the corrosion during normal operations, whereas TiAlC is expected to have resistance at higher temperatures in an accident. Unlike the results of Xu et al. (2016), Kumar et al. (2016) reported the poor performance of Zircaloy-4 cladding coated with Ti₂AlC and Zr₂AlC in the oxidation tests, which shows that the corrosion and oxidation resistance of those coating changes depend on the coating method and properties of the coating.

Kim et al. (2015a, 2016a,b) of the KAERI are conducting R&D to select coating materials, such as Cr, Cr-Al alloys, and FrCrAl, and coating methods. They reported that Cr-Al alloys with higher Cr content, which were prepared by vacuum arc melting process, showed superior resistance to corrosion in PWR-simulated conditions at 360°C and steam oxidation at 1200°C. On the contrary, they pointed out that it is necessary to prevent eutectic reaction between Zr and Fe when FeCrAl is used as the coating material deposited on Zr alloys.

3.2 FeCrAl alloys

FeCrAl alloys for ATF fuel cladding are mainly developed in United States (Yan et al., 2014; Field et al., 2016; Rebak et al., 2016; Terrani et al., 2016) and Japan (Sakamoto et al., 2015, 2016, 2017; Ukai et al., 2016, 2017), expecting the performance of the alloys at high temperatures during accidents. Higher thermal neutron absorption coefficients of FeCrAl alloys should be solved by reducing cladding wall thickness to introduce the alloys into the LWRs. Therefore, the strengthening of the alloys is a key requirement. Because the status of the development and characterization including the corrosion behavior of the alloys in the United States have been well summarized recently (Yan et al., 2014; Field et al., 2016; Rebak et al., 2016; Terrani et al., 2016), the recent R&D conducted in Japan is briefly reviewed below.

Oxide dispersion strengthened (ODS) alloys (FeCrAl-ODS alloys) have been employed as candidate accident-tolerant FeCrAl alloys for LWR use in Japan. Ukai et al. (2016, 2017) conducted the optimization of alloy design by modifying contents of the main elements (Cr and Al) and the minor elements (Zr and O). The modification of contents of the main elements was intended to obtain good oxidation resistance to high-temperature steam under DBA and beyond DBA (BDBA) conditions. The content of Cr was also adjusted to prevent α′ embrittlement under normal operating condition, which is a representative brittle manner of ferritic steels. The modification of the minor elements was aimed to increase strength in a wide temperature range. The measured ultimate tensile strength of the developed Ce-type FeCrAl-ODS alloy at 973 K was more than twice as high as that of unirradiated recrystallized Zircaloy-2 (Sakamoto et al., 2016). This high strength allows a thin wall-thickness design to reduce the penalty in neutron economics.

The oxidation tests in high-temperature steam at 1473 K showed 2–3 orders of magnitude smaller oxidation kinetics than representative Zr alloys. This good resistance to steam oxidation resulted in the preservation of the integrity of the cladding under simulated LOCA conditions. Actually, FeCrAl-ODS alloys did not show loss of geometry in the LOCA simulation test at 1473 K for up to 2 h (Sakamoto et al., 2017) as well as FeCrAl alloys (Ukai et al., 2016), which proved the much higher safety performance of the alloys after high-temperature oxidation. Although no irradiation test was conducted, α′ embrittlement could be prevented by controlling the Cr content. Sakamoto et al. (2015, 2016, 2017) reported the compatibility of FeCrAl-ODS alloys with neighboring materials in various conditions. The out-of-pile corrosion test in coolant water under normal operating conditions was conducted at 360°C in pure water environment (Sakamoto et al., 2015). A stable corrosion rate of less than 40 mg/dm²/year was found in the range examined. This good corrosion resistance is also reported on FeCrAl alloys by the Oak Ridge National Laboratory (ORNL; Terrani et al., 2016). Chemical stabilities were also examined with fuel pellet (UO₂; Sakamoto et al., 2015, 2016, 2017, neutron absorber (B₄C; Sakamoto et al., 2015, 2016), and Zircaloy (Sakamoto et al., 2015) at 1573 and 1673 K for 1 h. In all cases, the formation of alumina scale prevented significant chemical interactions between FeCrAl-ODS alloys and the neighboring materials. Figure 4 shows the interfaces between FeCrAl-ODS alloys and UO₂ pellet after the material reaction test at 1673 K for 1 h in He gas atmosphere (Sakamoto et al., 2016). The alumina layer prevented the reaction with UO₂ pellet unlike the Zircaloy-UO₂ reaction (Figure 5; Sakamoto et al., 2016). According to those high performances at high temperatures, the FeCrAl-ODS fuel cladding is expected to have advantages under DBA and BDBA conditions compared to the representative Zr alloys up to its melting point (~1800 K).

Figure 4:

Microstructure and elemental distribution near the surface of Ce-type FeCrAl-ODS alloy heated with UO₂ at 1673 K for 1 h (Sakamoto et al., 2016).

Figure 5:

Microstructure and elemental distribution near the surface of Zircaloy heated with UO₂ at 1673 K for 1 h (Sakamoto et al., 2016).

3.3 SiC

The material properties of SiC have been well studied and the basic data regarding corrosion and oxidation behavior are often reviewed as represented by references (Jacobson, 1993; Munro & Dapkunas, 1993; Narushima et al., 1997; Snead et al., 2007; Roy et al., 2014). However, SiC is generally used as coating or composite for industrial purposes and the properties of coating and composite are altered depending on the impurity, composition, microstructure, and fabrication method. Because the SiC/SiC composite used for cladding and composition, microstructure, and fabrication method differ by developers, the properties of the composites are not always the same and studies are necessary for each product.

SiC is known for its chemical and thermal stability, high strength, and high thermal conductivity. Meanwhile, SiC preferentially reacts with oxygen at high temperatures to form SiO₂. The oxidation behavior differs with impurity, microstructure, temperature, and oxygen potential (Roy et al., 2014). However, two basic oxidation processes are possible: one is passive oxidation at high oxygen potentials and high temperatures and the other is active oxidation at lower oxygen potentials.

Passive oxidation is considered dominant under LWR coolant conditions (Schneider et al., 1998). In that environment, the SiO₂ layer forms on the surface and the oxidation rate decreases due to the lower diffusion rate of oxygen in the oxide layer.

In addition to previous studies on the basic material properties of SiC (Jacobson, 1993; Munro & Dapkunas, 1993; Narushima et al., 1997; Snead et al., 2007; Roy et al., 2014), oxidation studies have been conducted with SiC fabricated with various methods and SiC/SiC composite, considering the application to LWR in recent years. Some recent investigations and results are introduced below.

Park et al. (2013) of the KAERI conducted the corrosion test with SiC coupons in the static autoclave and loop at 360°C. SiC coupons were prepared by chemical vapor deposition (CVD). The weight loss measured after the corrosion tests was obviously greater in the statistic autoclave than in the loop. The oxygen concentration was controlled at about 5 ppb in the loop tests, whereas it was not controlled and therefore expected to be much higher in the autoclave tests. An enhanced oxidation of SiC in high oxygen concentration was already reported (Hirayama et al., 1989; Kim et al., 2003; Barringer et al., 2007) and the KAERI results agree with those of the previous studies. Scanning electron microscopy (SEM) observation showed that corrosion occurred with the preferential dissolution of grain boundaries and subsequent dispatch of grains of SiC. They also reported the negligible influence of water chemistry simulating PWR coolant (Li/B=2.2 ppm/650 ppm) and the small protective effect of the preoxidation of SiC coupons.

Okonogi et al. (2015) conducted the autoclave corrosion tests for SiC coupons fabricated by three different methods: CVD, liquid-phase sintering (LPS), and reaction sintering. The tests were conducted at 290°C, 320°C, and 360°C for 168 h in 20 MPa water with an oxygen concentration of 8 ppm. It was shown that weight change by corrosion is small regardless of temperature in CVD SiC, whereas it is obviously greater and increases with temperature in LPS SiC. The test results also showed an increase in the corrosion rate with an increase in the amount of sintering additives, such as Al₂O₃ and Y₂O₃, which were possibly caused by the dissolution of the additives from the grain boundary of SiC. To estimate irradiation effects, high-purity samples of CVD SiC were corroded at 320°C for 165 h after irradiation of 5.1 MeV Si ions to 2.5 dpa. The test results suggested that irradiation swelling is insignificant and the corrosion rate is increased by irradiation.

Lorrette et al. (2015) of the CEA fabricated tubular SiC/SiC composites and conducted autoclave tests to evaluate the corrosion behavior under LWR operating conditions in partnership with AREVA and EDF. The tubular fibrous performs of Hi-Nicalon type S were chemically vapor infiltrated with a single pyrocarbon and the dimension of the specimen was about OD 10 mm×t 0.85–0.90 mm×L 60 mm. The corrosion tests were conducted at 360°C and 180 bar in a simulated PWR coolant condition (2 ppm LiOH and 1000 ppm H₂BO₃). Weight decrease was very small, except for the specimens highly predamaged with uniaxial loading that possibly caused large crack opening. The X-ray photoelectron spectroscopy (XPS) analyses of the corroded specimens showed that the SiO₂ layer, which should have been formed by the active oxidation of SiC described by Equation (1), did not form on the surface. They discussed the mechanism of the dissolution of SiO₂. The result showing no evidence of SiO₂ formation agreed with those obtained by previous corrosion tests under simulated reactor coolant conditions with the similar nuclear-grade SiC (Kim et al., 2003; Terrani et al., 2015).

In-pile corrosion studies are still limited, although some irradiation test plans are indicated by some groups. Currently, the irradiation test at the Halden reactor conducted by the Muroran Institute of Technology is the only project that has provided information on in-pile corrosion. According to the paper presented by Park et al. (2016) of the Muroran Institute of Technology, coupons, cladding specimens, and segment rods of SiC/SiC composites fabricated by the nanopowder infiltration and transient eutectic phase (NITE) method were irradiated in three irradiation campaigns. In the first irradiation of the segment rods with EB-welded Zircaloy end-caps, the weight loss of SiC was estimated from the measured concentration of Si dissolved into the coolant of the loop. Unfortunately, the reliability of estimation is considered low due to the unknown surface area, unknown effect of galvanic corrosion with Zircaloy, and effect of the fractured segment; nevertheless, it was shown that corrosion is obviously accelerated by irradiation. The second irradiation of various types of specimens was conducted under two different water chemistries: oxygen water chemistry (200–300 ppb) and hydrogen water chemistry (1.8–2.2 ppm). Although it was noted by the authors that further analysis and refinement are necessary, they successfully provided the preliminary quantitative results on the SiC weight loss of 99–120 and 28–32 mg/dm²/day in the oxygen and hydrogen water chemistries, respectively. Accordingly, hydrogen chemistry can drastically reduce the corrosion rate of SiC under irradiation conditions. Postirradiation tests of the segments confirmed the integrity of the joint part with Zircaloy end-caps and no change in the microstructure of the CVD SiC coating.

Lahoda et al. (2016) reported that the corrosion rate of SiC is changed by the addition of H₂O₂ and H₂ in the autoclave tests. The addition of H₂O₂ increases and the addition of H₂ decreases the corrosion rate, which agrees with the results of the above irradiation test at the Halden reactor. The observations and theoretical considerations indicate that the SiO₂ formed on the surface of SiC is dissolved more at lower H₂ concentrations. Their results may be available to understanding the corrosion mechanism and measures to reduce SiC corrosion under LWR coolant conditions.

Figure 6 summarizes some recent results from corrosion tests for SiC composites (Munro & Dapkunas, 1993; Roy et al., 2014; Sakamoto et al., 2015). It should be noted that variations in corrosion rate are caused by three factors at least: differences in specimen preparation including fabrication method, corrosion environment including water chemistry, and irradiation condition. However, it is evident first that SiC composites have higher corrosion resistance compared to Zr alloys if the water chemistry is properly controlled. The figure shows that the corrosion resistance of SiC composites likely decreases with irradiation in the reactor coolant, which are already suggested by Park et al. (2016), even if the differences in specimen preparation and corrosion environment are considered. The behavior of SiC cladded fuel, such as the decrease and mechanism of the corrosion and the interactions with the fuel pellets under reactor operating conditions, should be carefully investigated to demonstrate the feasibility of practical use in LWRs.

Figure 6:

Weight loss of SiC composites measured in-pile and out-of-pile corrosion tests (Munro & Dapkunas, 1993; Roy et al., 2014; Sakamoto et al., 2015).

As for the behavior of SiC composites under DBA (LOCA) conditions, Bacalski et al. (2016) of the GA conducted test with SiC/SiC tubes made of Hi Nicalon type S fiber. The nominal dimension of the tubes was OD 10 mm×ID 8 mm×L 25 mm. A pyrolytic carbon layer was formed at the OD by infiltration with SiC via chemical vapor infiltration (CVI). The layer of monolithic SiC encapsulated the SiC-SiC to reduce permeability through the base layer. The tubes were heated at 200°C–1000°C and water quenched to ensure the basic performance to maintain the integrity in an LOCA. They confirmed by helium leak tests and mechanical tests that neither leakage nor cracking occurred during heating and quenching. Cheng et al. (2012) of the ORNL conducted oxidation tests at 800°C–1000°C for coupon specimens of CVD-SiC, NITE-SiC, and FeCrAl. NITE-SiC coupons showed very small weight change at 800°C and 1000°C; however, porous SiO₂ formed on the surface and weight gain was observed at 1200°C. CVD-SiC coupons showed lower weight loss under examined conditions than NITE-SiC. Based on the pressure dependency, they estimated that loss of Si is controlled by diffusion of Si(OH)₄ to the gas phase. Okonogi et al. (2015) conducted the steam oxidation tests (1200°C and 1400°C for 72 h) with their CVD, LPS, and reaction sintering SiC specimens. It was shown that CVD and LPS SiC specimens exhibit 1000th of weight change of Zircaloy-2. In the oxidation tests with CVI-SiC/SiC tubular composites by Lorrette et al. (2015), weight gains of 0.82%–1.32% were measured after exposure to the mixed air/H₂O and O₂/H₂O atmospheres at 1200°C and 1400°C for 110 h. They gave consideration about the oxidation mechanism that consists of SiO₂ formation and volatile hydroxides or oxohydroxides and induced oxidation rate constants based on the measured oxidation weight gain and the theoretical consideration.

The data are still limited; however, SiC/SiC composites most likely have much lower oxidation rate and maintain the geometry under assumed accidental conditions. Nevertheless, the database should be expanded to confirm the performance of the SiC cladded fuel under DBA and BDBA conditions.

3.4 Coated and lined Mo alloy cladding

Metals and alloys adoptable to LWR fuel cladding are very limited in terms of neutron, corrosion, and oxidation resistance and mechanical properties that are required for the fuel design. Mo alloys have moderate thermal neutron absorption cross-section similar to those of steels, the high melting temperature of approximately 2600°C, and the high strength at high temperatures, which can be the motivation for various engineering applications. If the practical use of Mo alloys in LWR is assumed, R&D is necessary to solve subjects such as susceptible to corrosion and oxidation and formation of volatile MoO₃ at high temperatures. The recent progress of the R&D of Mo alloys for LWR application is summarized below.

As the technical solution for protecting Mo cladding from corrosion and oxidation environments expected in LWRs, Cheng et al. (2016a,b) and Kim et al. (2015b) of the EPRI and GE have been developing and testing Mo cladding coated with Zircaloy and FeCrAl. Coated Zircaloy is expected to transfer to ZrO₂ during the heat-up period below 1000°C to protect the substrate at higher temperatures. Coated FeCrAl is expected to have Cr₂O₃ layer on the surface in the reactor operating conditions and Al₂O₃ at higher temperatures in accidents to protect the substrate. In the corrosion test at 290°C for 4 months in simulated BWR conditions (0.3 ppm H₂), Zircaloy coating was mostly oxidized possibly due to the loss of alloying elements during the coating (PVD) process. Accordingly, corrosion resistance of Zircaloy-coated cladding appears to be currently similar or lower than that of Zircaloy cladding. FeCrAl coating was intact and showed high corrosion resistance with the formation of the protective layer of Cr₂O₃; however, separation from the substrate occurred during the corrosion test, which is likely caused by the difference in thermal expansion.

Mo cladding coated with Zircaloy and FeCrAl was able to maintain the geometry even after steam oxidation at 1000°C, although Zircaloy coating was oxidized to ZrO₂ and damage was seen in the oxidized coating. FeCrAl coating still showed high resistance to oxidation at 1200°C for 24 h. Zircaloy coating was significantly damaged and the protective effect was partly lost, which was also caused by the decrease in contents of alloying elements during the coating process. Microstructure observation showed that the alumina layer formed on the surface of FeCrAl coating at 1000°C–1350°C, where the protective effect was seen. On the contrary, the alumina layer was not seen in the specimen oxidized at 1500°C. An observation of the interface between the coatings and substrates showed the formation of the diffusion layer of Mo/Fe/Cr between FeCrAl and Mo at temperatures higher than 1200°C. The formation of the layer may reduce the protective effect of the coating. The authors concluded that the protective effect of FeCrAl coating for Mo cladding was currently confirmed up to 1200°C for 24 h. Kumar et al. (2016) of the AREVA reported similar results on the corrosion and oxidation behavior of coated Mo cladding.

As the next challenge to use Mo alloys to LWR cladding, the EPRI has started the development to form Mo alloy tubes lined with Zr alloy or FeCrAl, like duplex or triplex, via hot isostatic pressuring (HIPing). The stability in high-temperature steam conditions was confirmed and the fabricability of short segments with high quality was verified. Fabrication of longer tubes, characterization of relevant properties, and mechanical/thermomechanical/neutronic evaluations are anticipated (In use, 2017).