Protective Coatings for ATF Claddings
Since 2011, following the tragic Fukushima Daiichi Nuclear accident, great attention has been devoted to the development of a new concept of nuclear fuel to improve the safety of nuclear reactors during normal operation, transient modes and under accident conditions. A concept called “accident tolerant fuel (ATF)” indicates a strategy to prevent/limit the interaction of cladding material with water steam, or hydrogen embrittlement, and to reduce heat generation during cladding oxidation and increase “processing time” under accident conditions before re-flooding of the nuclear core.
In recent decades, zirconium-based alloys have been used in the nuclear industry as the main material for fuel claddings and other structural components of pressurized and boiling water reactors (PWRs and BWRs), Russian water–water energetic reactors (VVERs) and high-power channel-type reactors (RBMK) . During normal operation, zirconium alloys form a protective zirconium oxide layer that reduces cladding corrosion in the coolant. The key factors affecting claddings degradation are radiation swelling and embrittlement caused by oxidation and hydrogenation of the zirconium claddings . However, zirconium alloys demonstrate poor oxidation kinetics at elevated temperatures.
The most crucial conditions for fuel claddings can occur in the case of a loss of coolant accident (LOCA) . LOCA events can be caused by a breakup of the primary cooling system that results in the loss of pressure in the nuclear core and vaporization of the coolant. Under these conditions, the fuel temperature rises, thereby increasing the porosity of the fuel and resulting in its fragmentation. The fuel cladding temperature also increases abruptly. When Zr-based (E110, E635, Zircaloys, M5, ZIRLO, etc.) claddings interact with water steam at a high temperature (above 800 °C), it causes oxidation and embrittlement by releasing additional heat due to the exothermic reaction:
Therefore, the development of ATF cladding material is desperately needed to enhance the robustness of LWRs in normal and possible accident conditions.
Nowadays, two main ATF strategies are considered. The first is to replace current cladding material with SiCf/SiC composites , FeCrAl , molybdenum alloys  or Ni-based stainless steels . Taking into account the duration and possible total costs for developing a new type of cladding material, this way is considered as a long-term strategy. The second ATF concept is the development of protective coatings on the surface of Zr fuel claddings . Protective coatings should strongly improve corrosion and high-temperature (HT) oxidation resistance, wear-resistance and reduce hydrogen absorption of Zr-based alloys used in LWRs. Therefore, coating technology can be added to the technological process of nuclear fuel production that can be achieved in a short-term period. Despite the simplicity of this approach, a large number of possible factors (coating adhesion, thermal conductivity, thermal neutron cross-section, radiation resistance, mechanical properties ) can affect the behavior of coated Zr claddings in both normal and accident conditions. Currently, numerous studies are being performed to improve zirconium alloy performance by the deposition of metallic (Fe-based alloys, Cr, Cr-Al, Y, Ni-Cr, etc.), non-metallic (oxides, nitrides, carbides) or MAX-phase coatings . Among a variety of coatings, the highest performance in LOCA scenarios belongs to materials which can produce oxide phases such as alumina, zirconia, chromia or silica .
2. Protective Coatings for ATF Claddings
Numerous studies have been performed to improve zirconium alloys’ performance under both normal and accident conditions by the deposition of protective coatings . Coated Zr alloys show enhanced corrosion and HT oxidation resistance; although reduced, relatively thin coatings are not influenced by the thermomechanical behavior of Zr-based claddings , the coatings should not noticeably change neutron absorption in LWRs  and heat transfer between the cladding and coolant .
2.1. Metallic Coatings on Zr-Based Alloys
2.1.1. Fe-Based Coatings
Currently, there are two major research institutes focusing on the development of FeCrAl coatings for nuclear fuel claddings. These are the Korea Atomic Energy Research Institute (KAERI) and the University of Illinois Urbana-Champaign (UIUC). Fe-based coatings on Zr necessitate a barrier layer between the coating and substrate to mitigate Zr-Fe eutectic formation at high temperatures . Zhong et al. investigated the oxidation performance of FeCrAl coatings under HT steam conditions . In this study, FeCrAl was deposited on Zry-2 alloy using magnetron sputtering. Based on the experimental results, 1 µm-thick FeCrAl coating prevented the oxidation of the Zry-2 substrate in steam at 700 °C up to 15 h, while a thin FeCrAl (0.3 µm) film did not protect the alloy. Furthermore, investigations performed under autoclave testing (20 days) in a simulated BWR environment demonstrated an adequate performance without the loss of coating integrity. Kim et al. investigated the surface-modified Zr cladding concept: an outer FeCrAl protective layer was chosen as a coating to enhance the oxidation/corrosion resistance of Zr alloy deposited by laser coating and arc ion implantation methods to increase the adhesion strength of the coating and the Zr alloy tubes . The obtained samples demonstrate improved corrosion/oxidation performance, creep and irradiation resistance compared to commercial Zr claddings. Good radiation resistance was also observed for FeCrAl alloys under heavy ion irradiation . There was no void formation or Cr-enriched phases in 10Cr and 13Cr FeCrAl alloys, and irradiation-induced defects only contributed to hardening. Dabney et al. showed that thick FeCrAl coatings obtained by cold spraying demonstrate high oxidation resistance during autoclave tests and HT oxidation in the air . However, a fast diffusion of Fe from FeCrAl coating to Zr alloys caused a thick interlayer composed of (Fe,Cr)2Zr, FeZr3 and FeZr2 Laves phases (Figure 1a,b) .
Figure 1. Cross-section SEM images after 1200 °C air oxidation: (a) FeCrAl-coated Zircaloy-4 and (b) corresponding EDS line scan; (c) FeCrAl/Mo coating and (d) magnified image of the FeCrAl/Mo/Zr interface. Reproduced from  with permission by Elsevier.
Park et al. investigated FeCrAl coatings deposited onto Zr substrates using cold spraying techniques . For the FeCrAl/Zr system, a Mo layer was deposited between the FeCrAl coating and the Zr substrate to inhibit interdiffusion at high temperatures. As a result of mutual diffusion in the FeCrAl/Mo/Zr system, a complex multilayer structure was formed (Figure 1c,d). The FeCrAl coatings improved the oxidation resistance of the Zr alloy when exposed to a steam environment at 1200 °C. The ballooning behavior and mechanical properties of the coated samples studied under simulated LOCA conditions revealed higher burst temperatures, lower circumferential strain and smaller ruptures compared to the bare Zr. In addition, four-point bending and ring compression tests indicated a minimal increase in the maximum load and higher residual ductility of the coated Zr alloy, respectively .
2.1.2. Chromium-Based Coatings
Metallic Cr-based coatings are most promising in that they meet the basic requirements for ATF coating materials for Zr fuel claddings. Chromium has a high melting point, high corrosion resistance in water and steam due to a protective chromia scale, and a coefficient of thermal expansion similar to zirconium alloys. A wide variety of methods were implemented to deposit Cr coatings on Zr alloys such as magnetron sputtering , cathodic arc deposition , cold spraying , 3D laser cladding , electroplating , etc. The high potential of these coatings is also confirmed by the involvement of large industrial/research institutes such as CEA (France), VNIINM (Russia), KIT (Germany) and others. Several full-scale tests with Cr-coated Zr claddings with UO2 fuel in nuclear reactors are being performed at present.
Good protective properties of Cr coatings were demonstrated under normal operation conditions: autoclave testing in PWR and BWR simulated medium . Wei et al. showed high corrosion resistance in Cr-coated Zry-4 alloy in both H3BO3-LiOH and dissolved oxygen containing water , showing only 50–100 nm stable Cr2O3 scale after a 3000 h autoclave test. Chromium coatings deposited on Zry-4 alloy exhibited promising performance under steady-state, power ramp and LOCA transient conditions . Brachet et al. showed good protective properties of a 5–12 µm thick Cr coating under steam oxidation at 1200 °C (Figure 2) . It was demonstrated that the coating enhanced the oxidation resistance of the alloy and ensured the integrity of the cladding for a longer oxidation time (approx. 10 times longer than current LOCA criteria). Ma et al. showed that Cr coatings could protect Zr-1Nb alloy from HT steam oxidation at 1200 and 1300 °C (Figure 3) . It was shown that the thickness of the formed Cr2O3 and residual Cr layers depends on oxidation time, which is probably due to the combined effect of the ongoing redox reactions between chromium oxide and zirconium, also observed by Han et al. .
Figure 2. Optical images and weight gain evolution of uncoated and Cr-coated (5–12 μm) Zry-4 alloy during steam oxidation at 1200 °C. Reproduced from  with permission by Elsevier.
Figure 3. Optical and SEM images of Cr-coated zirconium alloy cladding after high-temperature steam oxidation at 1200 °C for 0.5 h (a), 1 h (b), 2 h (c), 4 h (d), and 1300 °C for 0.5 h (e) and 1 h (f). Reproduced from .
Results of mechanical tests under ambient and elevated (400 °C) temperatures showed no significant difference between the Cr-coated and the uncoated Zr claddings . Kim et al. investigated the mechanical behavior of Zry-4 alloy with 3D laser-coated Cr coatings (80–120 µm) under ring compression and tensile tests . The results indicate good adhesion and no defects in the Cr coatings up to 4% strain. The generation of cracks occurred at 6% strain without coating spallation or peeling. Ribis et al. analyzed the atomic-scale structure of the Cr/Zry-4 interface where the Cr coating was deposited using PVD method . It was found that a thin layer of Laves Zr(Cr,Fe)2 phase with C14 and C15 polytypes was formed at the interface, while the coherent boundaries ensured good adhesion of the coating.
The adhesion behavior of Cr coatings was also affected by oxidation and diffusion processes at the coating/alloy interface . Jiang et al. performed a comparative analysis of cracking resistance of multi-arc ion plated Cr and CrN coatings on Zry-4 alloy under tensile loading . It was revealed that Cr coatings exhibit a brittle-to-ductile transition under thermomechanical loading, which results in their better cracking resistance compared to CrN coatings.
Several studies presented a good radiation resistance of chromium coatings under ion irradiation . The stabilization of C14 and disappearance of C15 polytypes in the Cr/Zry-4 samples due to continuous incoming Fe flux were found during 20 MeV Kr+ irradiation at 400 °C. Despite this, the samples retained their adhesion and microstructure stability after ion irradiation . An acceptable swelling of ~1.6% was observed at an irradiation temperature of 500 °C under 1.4 MeV heavy ion radiation up to 25 dpa . This is twice lower than the allowable swelling value (~5%) for reactor materials.
Cr-Al alloy coatings show better oxidation resistance than pure chromium . However, a more complex and heterogeneous structure of the oxide layers forms after oxidation , which requires more detailed studies under long-term HT steam oxidation, as well as B-DBA conditions. Ni-Cr coatings exhibit better ductility than pure Cr coatings but, at high temperatures, they show lower oxidation resistance and rapid diffusion of Ni inside the zirconium alloy occurs .
2.1.3. Yttrium-Based Coatings
The resistance of yttrium to oxidation, as well as the formation of stable protective yttrium oxides at very high temperatures, made it very attractive for study. Sridharan et al. investigated surface modified Y-coated Zry-2 alloy and observed a decrease in the oxidation rate in supercritical water at 400 °C for 7 days (168 h) . Kim et al. investigated the surface modification of 2 mm-thick Zry-4 alloy by deposition of Y2O3 (10 µm) using a laser beam scan method to produce an oxide dispersion strengthening (ODS) treatment . It has been shown that the yield strength of Zry-4 with the ODS layer was 65% higher compared to uncoated alloy at 500 °C, therefore enhancing high-temperature strength to defeat the ballooning behavior of fuel cladding during an accident event . However, the limited research on these types of coatings does not allow us to confirm the feasibility of their use as ATF materials.
2.1.4. HEAs Coatings
High entropy alloys (HEA), also called multi-principal component alloys, are solid solutions of at least four elements in near equimolar ratios . HEAs demonstrate a wide range of outstanding properties, such as high thermal stability, corrosion resistance and radiation damage tolerance . Zhang et al. investigated corrosion behavior of a 3 µm-thick AlCrMoNbZr HEA coating deposited on an N36 alloy (Zr-1Sn-1Nb-0.3Fe) in static water at 360 °C and 18.7 MPa for 30 days . It was reported that the oxidation resistance of the N36 alloy increased threefold due to multiphase oxide phases (Nb2Zr6O17, ZrO2 and Cr2O3) formed at the surface. A multilayer AlCrMoNbZr/(AlCrMoNbZr)N coating can be beneficial for preventing Al migration and boehmite phase formation during autoclave corrosion of an AlCrMoNbZr alloy . The multilayer AlCrMoNbZr/(AlCrMoNbZr)N coating with the layer step of 50 nm demonstrated better protective properties compared to 5/5 nm and 10/10 nm multilayers and single-layer AlCrMoNbZr (Figure 4) . Despite the potential application of HEAs as ATF materials for fuel claddings , their application as protective coatings is challenging due to possible low temperature eutectics, with Zr alloys and complex oxide scales formed after HT oxidation.
Figure 6. Multilayer AlCrMoNbZr/(AlCrMoNbZr)N HEA coatings: (a) cross section TEM image of 10/10 nm coating; (b) mass gains for HEA-coated N36 zirconium alloy after an autoclave test at 360 °C, 18.7 MPa for 30 days; (c) depth distribution of elements in 50/50 nm HEA coating after the autoclave test. Reproduced from  with permission by Elsevier.
2.2. Non-Metallic Coatings on Zr-Based Alloys
2.2.1. Nitride Coatings
There are several research groups working on the development of nitride coatings for nuclear fuel Zr claddings. Khatkhatay et al. deposited TiN and Ti0.35Al0.65N coatings on Zry-4 substrates using pulsed laser deposition, and exposed them to supercritical water conditions for 48 h at a temperature of 500 °C and a pressure of 25 MPa . The coated tubes were remarkably intact after exposure, while uncoated tubes demonstrated severe oxidation and breakaway corrosion. The TiN coatings also reduce the hydrogenation of Zr alloys . However, nanocrystalline TiN coatings can dissociate under energetic particle bombardment, forming Ti-enriched zones with low oxidation resistance . Despite good diffusion barrier properties of TiN coatings , they are considered less as a protective coating or barrier interlayer because of their large coefficient of thermal expansion (CTE) difference with Zr and possible cracking during thermal cycling . Alat et al. studied the corrosion resistance of TiAlN and TiN coatings deposited on ZIRLO substrates by cathodic arc physical vapor deposition . The corrosion tests were implemented in static pure water at 360 °C and 18.7 MPa for 72 h. After the tests, a very low weight gain between 1–5 mg/dm2 was observed, whereas the uncoated ZIRLO samples showed an average weight gain of 14.4 mg/dm2. However, aluminum depletion in a high-temperature water environment results in the formation of a boehmite phase that degrades the corrosion resistance of TiAlN coatings (Figure 5). To eliminate boehmite formation, the deposition of an outer TiN layer or multilayer approach based on TiN/TiAlN coating can be effective .
Figure 5. Cross-section SEM images of TiAlN/Ti/ZIRLO™ after autoclave testing at 360 °C for 72 h: (a) secondary electron mode, and (b) backscattered electron mode. Reproduced from  with permission by Elsevier.
Daub et al. provided comparative analyses on corrosion resistance of 2–4 µm-thick CrN-, TiAlN- and AlCrN-coated Zry-4 alloy . It was shown that a CrN coating demonstrates better overall performance in both aqueous and steam environments, as well as twice-reduced hydrogen ingress. Meng et al. showed the high resistance of a 13 µm-thick CrN coating deposited on a Zr-702 alloy in the air up to 1160 °C (Figure 6) . Krejčí et al. demonstrated cracking and local failures of CrN coatings after HT steam oxidation at 1200 °C for 30 min . It was indicated that the cracking was caused by partial decomposition of CrN to Cr2N at high temperatures (typically below 850 °C) .
Figure 6. Weight gain of uncoated and CrN-coated Zr alloys as the function of oxidation temperature in the air (a), and the appearance of the samples after the oxidation tests (b). Reproduced from  with permission by Elsevier.
2.2.2. Zirconium Silicide Coatings
Zirconium silicides have high thermal conductivity and favorable mechanical properties . Yeom et al. studied the oxidation behavior of Zr2Si, ZrSi and ZrSi2 coatings deposited on a Zry-4 alloy by magnetron sputtering . Thicker (3.9 µm) ZrSi2 coating exhibited oxidation resistance almost two orders of magnitude higher compared to uncoated Zry-4 in 700 °C air for 20 h. The thicknesses of the oxide layers were 7 and 20 μm for coated Zry-4 at 1000 (1 h) and 1200 °C (10 min) steam, respectively. No cracking or spallation was observed after three cycles of water quenching from 700 °C. However, the brittle nature of silicide coatings and coating volatilization under autoclave testing limit their application for ATF .
2.2.3. Carbide Coatings
Silicon carbide is a promising material for ATF Zr claddings since it has a high melting point, low chemical reactivity, superior oxidation resistance at high temperatures and a lower thermal neutron cross-section than Zr-based alloys . SiC coatings deposited on a Zry-4 alloy using co-sputtering of SiC and Si targets demonstrate enhanced oxidation resistance in 900 °C steam due to the formation of a protective silica layer . SiC coatings obtained by magnetron sputtering also reduce the hydrogenation of zirconium alloys under normal operation temperatures . The effect of deposition parameters on the mechanical properties of SiC coatings deposited on Zircaloy-4 substrate by magnetron sputtering was shown in . Al-Olayyan et al. showed better adhesion of SiC coating on rough surfaces of Zry-4 alloy, which resulted in higher corrosion resistance . Nevertheless, unstable oxide growth and coating volatilization during autoclave corrosion tests must be considered carefully when designing SiC-based coatings for LWRs .
Zr-Al-C coatings obtained on Zry-4 alloy by magnetron sputtering demonstrated poor protective properties during oxidation in steam at temperatures beyond 800 °C due to formation of ZrO2 and Al2O3 oxides scales . Michau et al. studied the protective properties of Cr-based carbide (CrC, CrSiC) coatings deposited on the inner surface of Zr cladding tubes using DLI-MOCVD . Good adhesion and better oxidation resistance of amorphous CrxCy coating in air and steam at 1200 °C were shown (Figure 7), while the addition of Si was ineffective in improving oxidation resistance of chromium carbide coatings. It was also shown that an optimized DLI-MOCVD process can be successfully used to deposit CrCx coatings inside Zr cladding tubes 1 m in length .
Figure 7. Mass gains measured by TGA of uncoated and CrxCy-coated Zircaloy-4 under air oxidation isothermal air oxidation at 1200 °C (a); BSE SEM image and oxygen profiles of CrxCy-coated clad segment after oxidation in steam at 1200 °C for 10 min, followed by water quenching (b); and the detailed SEM + EBSD image of the inner surface of CrxCy-coated clad segment after the oxidation test (c). Reproduced from  with permission by Elsevier.
Jin et al. deposited 75Cr3C2-25NiCr (wt.%) on a Zr-2.5Nb alloy using a high velocity oxygen fuel (HVOF) technique and showed improved oxidation resistance in air and steam at 700–1000 °C . However, the coated samples showed poor corrosion resistance in supercritical water at 400 °C and 10.3 MPa, caused by a bimetallic effect (this coating can build up bimetallic couple with the Zr-2.5Nb alloy, resulting in the acceleration of the corrosion rate). Yang et al. demonstrated high oxidation resistance of 120 μm-thick HVOF Cr3C2-NiCr coating at 1200 °C in steam for 1 h due to the formation of a dense chromia scale . However, the high coating thickness and heterogeneous/porous microstructure formed by the HVOF method may limit the usefulness of such coatings for ATF Zr claddings.
2.2.4. MAX-Phase Coatings
MAX phases are layered hexagonal carbides and nitrides referenced to the general formula Mn+1AXn, (MAX), where “n” ranges between 1 and 4, “M” represents a transition metal (Sc, Ti, V, Cr, etc.), “A” is an A-group (mainly IIIA and IVA, or groups 13 and 14) element and “X” is either C and/or N . MAX phase materials exhibit excellent properties of both metals and ceramics such as high electrical and thermal conductivity, high melting points and excellent oxidation resistance .
Alumina-forming MAX phases demonstrate superior oxidation resistance at high temperatures and exhibit self-healing ability . Li et al. investigated high-temperature oxidation resistance of 12 μm-thick Ti2AlC coating on ZIRLO alloy in pure steam at 1000–1200 °C . The coating remains intact and demonstrates good protective properties up to 1200 °C during 5 min oxidation time. The improved oxidation resistance was due to the dense columnar-free microstructure of Ti2AlC coating and triple oxide scale α-Al2O3 + rutile-TiO2/α-Al2O3/TiO2. Maier et al. showed the oxidation resistance of cold sprayed Ti2AlC coating (90 μm) in Ar/steam mixture at 1005 °C for 20 min . The Ti2AlC coating protected the Zry-4 alloy from oxidation and had high hardness and wear resistance. Thinner Ti2AlC/TiC coatings (5/0.5 μm) deposited by magnetron sputtering demonstrate high oxidation resistance in steam at 800 °C, forming a triple-layered scale (θ-Al2O3 + TiO2/θ-Al2O3/TiO2) . It was also shown that the TiC barrier layer mitigates the inward diffusion of Al into the Zry-4 alloy. However, rapid oxidation at 1000 °C resulted in cracking and spallation of the Ti2AlC coatings.
The oxidation resistance of Ti-Al-C and Cr-Al-C coatings deposited by magnetron sputtering on ZIRLO alloy was investigated in . Results of the autoclave test at 360 °C and 18.6 MPa showed poor protective properties of Ti-Al-C coatings due to the formation of multiple oxide scale and hydroxide phases. The Cr2AlC coating demonstrates better oxidation performance; however, partial spallation and formation of volatilized AlOOH was also revealed. Tang et al. showed the excellent oxidation resistance and self-healing ability of thin Cr2AlC coatings in steam at 1000 °C (Figure 8) .
Figure 8. Cross-section images of three types of MAX-phase coatings with barrier layers on Zircaloy-4 after annealing: (a) Ti2AlC, (b) Zr(Al)C, (c) Cr2AlC, and their oxidation kinetics in 1000 °C steam (d). Reproduced from  with permission from the authors.
Imtyazuddin et al. analyzed the radiation resistance of magnetron sputtered Cr2AlC films under 320 keV Xe ions at 300 and 623 K . It was found that Cr2AlC films are amorphized at room temperature even at low doses; however, no amorphous phase was found up to 90 dpa at 623 K, indicating good radiation tolerance at elevated temperatures (Figure 9). Furthermore, both Ti2AlC and Cr2AlC coatings reduce hydrogen uptake effectively . Wang et al. also demonstrated good oxidation resistance of 10 μm-thick Cr2AlC coatings in air at 1100 °C .
Figure 9. Evolution of microstructure of magnetron-deposited Cr2AlC thin film under 320 keV Xe irradiation at 300 and 623 K. Reproduced from  with permission by Elsevier.
A literature review indicates the rapid diffusion of “A” element from the MAX phases into the Zr alloys with the formation of an interdiffusion layer, thereby decreasing the protective properties of the coating. Thus, various diffusion barrier layers between the coating and the substrate should be considered when developing protective coatings for ATF based on MAX-phases.
The entry is from 10.3390/coatings11050557
- Azhazha, V.; Borts, B.; Butenko, I.; Voevodin, V.; V’yugov, P.; Gritsina, V.; Krasnorutskij, V.; Lavrinenko, S.; Levenets, V.; Neklyudov, I. Zirconium-Niobium Alloys for NPP; Splav Tsirkonij-Niobij Dlya AehS: Alushta, Ukraine, 2012.
- Zieliński, A.; Sobieszczyk, S. Hydrogen-enhanced degradation and oxide effects in zirconium alloys for nuclear applications. Int. J. Hydrog. Energy 2011, 36, 8619–8629.
- Charit, I. Accident tolerant nuclear fuels and cladding materials. JOM 2018, 70, 173–175.
- Motta, A.T.; Capolungo, L.; Chen, L.Q.; Cinbiz, M.N.; Daymond, M.R.; Koss, D.A.; Lacroix, E.; Pastore, G.; Simon, P.-C.A.; Tonks, M.R.; et al. Hydrogen in zirconium alloys: A review. J. Nucl. Mater. 2019, 518, 440–460.
- Terrani, K.A.; Zinkle, S.J.; Snead, L.L. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding. J. Nucl. Mater. 2014, 448, 420–435.
- Kim, D.; Lee, H.-G.; Park, J.Y.; Kim, W.-J. Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications. J. Nucl. Mater. 2015, 458, 29–36.
- Katoh, Y.; Snead, L.L.; Henager Jr, C.H.; Nozawa, T.; Hinoki, T.; Iveković, A.; Novak, S.; De Vicente, S.G. Current status and recent research achievements in SiC/SiC composites. J. Nucl. Mater. 2014, 455, 387–397.
- Zhou, X.; Wang, H.; Zhao, S. Progress of SiCf/SiC composites for nuclear application. Adv. Ceram. 2016, 37, 151–167.
- Qiu, B.; Wang, J.; Deng, Y.; Wang, M.; Wu, Y.; Qiu, S. A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding. Nucl. Eng. Technol. 2020, 52, 1–13.
- Liu, J.; Zhang, X.; Yun, D. A complete review and a prospect on the candidate materials for accident-tolerant fuel claddings. Mater. Rev. 2018, 32, 1757–1778.
- Bragg-Sitton, S.M.; Todosow, M.; Montgomery, R.; Stanek, C.R.; Montgomery, R.; Carmack, W.J. Metrics for the technical performance evaluation of light water reactor accident-tolerant fuel. Nucl. Technol. 2016, 195, 111–123.
- Chen, H.; Wang, X.; Zhang, R. Application and development progress of Cr-based surface coatings in nuclear fuel element: I. selection, preparation, and characteristics of coating materials. Coatings 2020, 10, 808.
- Chen, H.; Wang, X.; Zhang, R. Application and Development Progress of Cr-Based Surface Coating in Nuclear Fuel Elements: II. Current Status and Shortcomings of Performance Studies. Coatings 2020, 10, 835.
- Terrani, K.A. Accident tolerant fuel cladding development: Promise, status, and challenges. J. Nucl. Mater. 2018, 501, 13–30.
- Cheng, B. Fuel behavior in severe accidents and Mo-alloy based cladding designs to improve accident tolerance. Atw. Int. Z. Fuer Kernenerg. 2013, 59, 158–160.
- Nelson, A.; Sooby, E.; Kim, Y.-J.; Cheng, B.; Maloy, S. High temperature oxidation of molybdenum in water vapor environments. J. Nucl. Mater. 2014, 448, 441–447.
- Cheng, B.; Chou, P.; Kim, Y.-J. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance. EPJ Nucl. Sci. Technol. 2016, 2, 5.
- Cheng, B.; Chou, P.; Kim, Y.-J. Development of Mo-Based Accident Tolerant LWR Fuel Cladding; International Atomic Energy Agency: Vienna, Austria, 2016.
- Karpyuk, L.; Savchenko, A.; Leont’eva-Smirnova, M.; Kulakov, G.; Konovalov, Y.V. Steel Cladding for VVER Fuel Pins in the Context of Accident-Tolerant Fuel: Prospects. At. Energy 2020, 128, 218–222.
- Guanghai, B.; Zhilin, C.; Yanwei, Z.; Erwei, L.; Jiaxiang, X.; Weiwei, Y.; Rongshan, W.; Rui, L.; Tong, L. Research progress of coating on zirconium alloy for nuclear fuel cladding. Rare Met. Mater. Eng. 2017, 46, 2035–2040.
- Maier, B.; Yeom, H.; Johnson, G.; Dabney, T.; Walters, J.; Romero, J.; Shah, H.; Xu, P.; Sridharan, K. Development of cold spray coatings for accident-tolerant fuel cladding in light water reactors. JOM 2018, 70, 198–202.
- Lustman, B. Zirconium technology—twenty years of evolution. In Zirconium in the Nuclear Industry; ASTM International: West Conshohocken, PA, USA, 1979.
- Baczynski, J. High Temperature Steam Oxidation of Titanium-Coated Zircaloy-2 and Titanium-Zirconium Alloys. Master’s Thesis, University of Illinois at Urbana-Champaign, Urbana, IL, USA, 2014.
- Kim, H.-G.; Kim, I.-H.; Jung, Y.-I.; Park, D.-J.; Yang, J.-H.; Koo, Y.-H. Development of surface modified Zr cladding by coating technology for ATF. Proc. Top Fuel 2016, 1157–1163.
- Sagiroun, M.I.; Xinrong, C. Zirconium-Based Cladding Coating Technique for Oxidation, Corrosion and Embrittlement Reduction at High-Temperature: An Overview. IOP Conf. Ser. Mater. Sci. Eng. 2019, 649, 012008.
- Barrett, F.; Huang, X.; Guzonas, D. Characterization of TiO2-doped yttria-stabilized zirconia (YSZ) for supercritical water-cooled reactor insulator application. J. Therm. Spray Technol. 2013, 22, 734–743.
- Kane, K.A.; Stack, P.; Mouche, P.A.; Pillai, R.R.; Pint, B.A. Steam oxidation of chromium corrosion barrier coatings for sic-based accident tolerant fuel cladding. J. Nucl. Mater. 2021, 543, 152561.
- Chen, Q.; Liu, C.; Zhang, R.; Yang, H.; Wei, T.; Wang, Y.; Li, Z.; He, L.; Wang, J.; Wang, L. Microstructure and high-temperature steam oxidation properties of thick Cr coatings prepared by magnetron sputtering for accident tolerant fuel claddings: The role of bias in the deposition process. Corros. Sci. 2020, 165, 108378.
- Rickover, H.G.; Geiger, L.D.; Lustman, B. History of the Development of Zirconium Alloys for Use in Nuclear Reactors; Energy Research and Development Administration: Washington, DC, USA, 1975.
- Gadiyar, H. Corrosion of zirconium base alloys-an overview. In Proceedings of Symposium on Zirconium Alloys for Reactor Components; Bhabha Atomic Research Centre: Bombay, India, 1992.
- Kass, S. The development of the zircaloys. In Corrosion of Zirconium Alloys; ASTM International: West Conshohocken, PA, USA, 1964.
- Motta, A.T.; Couet, A.; Comstock, R.J. Corrosion of zirconium alloys used for nuclear fuel cladding. Annu. Rev. Mater. Res. 2015, 45, 311–343.
- Cox, B. Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys. J. Nucl. Mater. 2005, 336, 331–368.
- Alam, T.; Khan, M.K.; Pathak, M.; Ravi, K.; Singh, R.; Gupta, S. A review on the clad failure studies. Nucl. Eng. Des. 2011, 241, 3658–3677.
- Moorthy, K.B. Current trends in the Use of Zirconium Alloys; National Metallurgical Laboratory: Jamshedpur, India, 1969.
- Shimada, S.; Cheng, B.; Lutz, D.; Kubota, O.; Ichikawa, N.; Ibe, H. In-core tests of effects of BWR water chemistry impurities on zircaloy corrosion. In Zirconium in the Nuclear Industry: Fourteenth International Symposium; ASTM International: West Conshohocken, PA, USA, 2005.
- Krishnan, R.; Asundi, M. Zirconium alloys in nuclear technology. Proc. Indian Acad. Sci. Sect. C Eng. Sci. 1981, 4, 41–56.
- The Fukushima Daiichi Nuclear Power Plant Accident; OECD: Paris, France, 2013.
- Nikulina, A. Zirconium alloys in nuclear power engineering. Met. Sci. Heat Treat. 2004, 46, 458–462.
- Nikulina, A.V.; Markelov, V.; Peregud, M.; Bibilashvili, Y.K.; Kotrekhov, V.; Lositsky, A.; Kuzmenko, N.; Shevnin, Y.P.; Shamardin, V.; Kobylyansky, G. Zirconium alloy E635 as a material for fuel rod cladding and other components of VVER and RBMK cores. In Zirconium in the Nuclear Industry: Eleventh International Symposium; ASTM International: West Conshohocken, PA, USA, 1996.
- Markelov, V.; Novikov, V.; Nikulina, A.; Konkov, V.; Sablin, M.; Novoselov, A.; Kobylyansky, J. Application of E635 alloy as structural components of WWER-1000 fuel assemblies. In Proceedings of the International Conference on WWER fuel Performance, Modelling and Experimental Support, Albena Bulgaria, 19–23 September 2005.
- Hózer, Z.; Győri, C.; Matus, L.; Horváth, M. Ductile-to-brittle transition of oxidised Zircaloy-4 and E110 claddings. J. Nucl. Mater. 2008, 373, 415–423.
- Novikov, V.; Nesterov, B.; Troyanov, V.; Izhutov, A.; Burukin, A.; Shahmut, E. Out of reactor investigation corrosive characteristics cladding of new Zirconium alloys as applied to conditions reactor-plant WWER-1200 (AES-2006) and program irradiation tests of this alloys. In Proceedings of the International Conference on WWER Fuel Performance, Modelling and Experimental Support, Albena, Bulgaria, 17–24 September 2011.
- Novikov, V.; Markelov, V.; Gusev, A.; Malgin, A.; Kabanov, A.; Pimenov, Y. Some results on the properties in-vestigations of zirconium alloys for WWER-1000 fuel cladding. In Proceedings of the International Conference on WWER Fuel Performance, Modelling and Experimental Support, Albena, Bulgaria, 17–24 September 2011.
- Foster, J.P.; Yueh, H.K.; Comstock, R.J. ZIRLO TM cladding improvement. J. Astm Int. 2008, 5, 1–13.
- Le Saux, M.; Vandenberghe, V.; Crébier, P.; Brachet, J.; Gilbon, D.; Mardon, J.; Jacques, P.; Cabrera, A. Influence of steam pressure on the high temperature oxidation and post-cooling mechanical properties of Zircaloy-4 and M5 cladding (LOCA conditions). In Zirconium in the Nuclear Industry: 17th Volume; ASTM International: West Conshohocken, PA, USA, 2015.
- Brachet, J.-C.; Portier, L.; Forgeron, T.; Hivroz, J.; Hamon, D.; Guilbert, T.; Bredel, T.; Yvon, P.; Mardon, J.-P.; Jacques, P. Influence of hydrogen content on the α/β phase transformation temperatures and on the thermal-mechanical behavior of Zy-4, M4 (ZrSnFeV), and M5™(ZrNbO) alloys during the first phase of LOCA transient. In Zirconium in the Nuclear Industry: Thirteenth International Symposium; ASTM International: West Conshohocken, PA, USA, 2002.
- Wakamatsu, A.; Nunokawa, K.; Nakano, M.; Hamasaki, M.; Uno, Y.; Kawagoe, T. Development of advanced fuel and core for high reliability and high performance. Mitsubishi. Juko Giho. 2006, 43, 20–24.
- Malgin, A.G.; Markelov, V.A.; Novikov, V.V.; Shelepov, I.A.; Donnikov, V.E.; Latunin, V.I.; Krejci, J. Research of high-temperature oxidation behavior of E110 opt and E110М sponge based zirconium alloys. Proc. Top Fuel 2018, 239, 1–10.
- Pawel, R.E.; Cathcart, J.V.; McKee, R.A. The kinetics of oxidation of Zircaloy-4 in steam at high temperatures. J. Electrochem. Soc. 1979, 126, 1105.
- Yan, Y.; Garrison, B.E.; Howell, M.; Bell, G.L. High-temperature oxidation kinetics of sponge-based E110 cladding alloy. J. Nucl. Mater. 2018, 499, 595–612.
- NRC. 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors; US NRC: Washington, DC, USA, 2017.
- Le Saux, M.; Brachet, J.-C.; Vandenberghe, V.; Ambard, A.; Chosson, R. Breakaway oxidation of zirconium alloys exposed to steam around 1000 °C. Corros. Sci. 2020, 176, 108936.
- Steinbrueck, M.; da Silva, F.O.; Grosse, M. Oxidation of Zircaloy-4 in steam-nitrogen mixtures at 600–1200 °C. J. Nucl. Mater. 2017, 490, 226–237.
- Steinbrück, M. Prototypical experiments relating to air oxidation of Zircaloy-4 at high temperatures. J. Nucl. Mater. 2009, 392, 531–544.
- Steinbrück, M.; Böttcher, M. Air oxidation of Zircaloy-4, M5® and ZIRLO™ cladding alloys at high temperatures. J. Nucl. Mater. 2011, 414, 276–285.
- Tang, C.; Stueber, M.; Seifert, H.J.; Steinbrueck, M. Protective coatings on zirconium-based alloys as accident-tolerant fuel (ATF) claddings. Corros. Rev. 2017, 35, 141–165.
- Bischoff, J.; Vauglin, C.; Delafoy, C.; Barberis, P.; Perche, D.; Guerin, B.; Vassault, J.; Brachet, J. Development of Cr-coated zirconium alloy cladding for enhanced accident tolerance. In Proceedings of the Top Fuel 2016, Boise, ID, USA, 11–16 September 2016; pp. 1165–1171.
- Maier, B.; Yeom, H.; Johnson, G.; Dabney, T.; Walters, J.; Xu, P.; Romero, J.; Shah, H.; Sridharan, K. Development of cold spray chromium coatings for improved accident tolerant zirconium-alloy cladding. J. Nucl. Mater. 2019, 519, 247–254.
- Wei, T.; Zhang, R.; Yang, H.; Liu, H.; Qiu, S.; Wang, Y.; Du, P.; He, K.; Hu, X.; Dong, C. Microstructure, corrosion resistance and oxidation behavior of Cr-coatings on Zircaloy-4 prepared by vacuum arc plasma deposition. Corros. Sci. 2019, 158, 108077.
- Bryan, W.J.; Jones, D. Wear resistant coating for fuel cladding. Patent 5,268,946, 7 December 1993.
- Gray, D.M.; White, D.W.; Andresen, P.L.; Kim, Y.J.; Lin, Y.P.; Curtis, T.C.; Patterson, C.B. Fuel Rod with Wear-Inhibiting Coating. U.S. Patent 11/780,537, 22 January 2009.
- Kashkarov, E.B.; Nikitenkov, N.; Sutygina, A.; Obrosov, A.; Manakhov, A.; Polčak, J.; Weiß, S. Hydrogen absorption by Ti-implanted Zr-1Nb alloy. Int. J. Hydrogen Energy 2018, 43, 2484–2491.
- Kashkarov, E.; Nikitenkov, N.; Sutygina, A.; Laptev, R.; Bordulev, Y.; Obrosov, A.; Liedke, M.O.; Wagner, A.; Zak, A.; Weiβ, S. Microstructure, defect structure and hydrogen trapping in zirconium alloy Zr-1Nb treated by plasma immersion Ti ion implantation and deposition. J. Alloys Compd. 2018, 732, 80–87.
- Kashkarov, E.; Sidelev, D.; Rombaeva, M.; Syrtanov, M.; Bleykher, G. Chromium coatings deposited by cooled and hot target magnetron sputtering for accident tolerant nuclear fuel claddings. Surf. Coat. Technol. 2020, 389, 125618.
- Lee, Y.; Lee, J.I.; No, H.C. Mechanical analysis of surface-coated zircaloy cladding. Nucl. Eng. Technol. 2017, 49, 1031–1043.
- Younker, I.; Fratoni, M. Neutronic evaluation of coating and cladding materials for accident tolerant fuels. Prog. Nucl. Energy 2016, 88, 10–18.
- Kam, D.H.; Lee, J.H.; Lee, T.; Jeong, Y.H. Critical heat flux for SiC-and Cr-coated plates under atmospheric condition. Ann. Nucl. Energy 2015, 76, 335–342.
- Birks, N.; Meier, G.H.; Pettit, F.S. Introduction to the High Temperature Oxidation of Metals, 2nd ed.; Cambridge University Press: Cambridge, UK, 2006.
- Young, D.J. High Temperature Oxidation and Corrosion of Metals, 2nd ed.; Elsevier: Amsterdam, The Netherlands, 2008.
- Kofstad, P. High Temperature Corrosion; Elsevier Applied Science Publishers Ltd.: Amsterdam, The Netherlands, 1988.
- Sarrazin, P.; Galerie, A.; Fouletier, J. Mechanisms of High Temperature Corrosion; Trans Tech Publications Ltd.: Bäch, Switzerland, 2008.
- Pasamehmetoglu, K.; Massara, S.; Costa, D.; Bragg-Sitton, S.; Moatti, M.; Kurata, M.; Iracane, D.; Ivanova, T.; Bischoff, J.; Delafoy, C. State-of-the-Art Report on Light Water Reactor Accident-Tolerant Fuels; Organisation for Economic Co-Operation and Development: Paris, France, 2018.
- Zhong, W.; Mouche, P.A.; Han, X.; Heuser, B.J.; Mandapaka, K.K.; Was, G.S. Performance of iron–chromium–aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions. J. Nucl. Mater. 2016, 470, 327–338.
- Kim, H.-G.; Kim, I.-H.; Jung, Y.-I.; Park, D.-J.; Park, J.-H.; Yang, J.-H.; Koo, Y.-H. Progress of surface modified Zr cladding development for ATF at KAERI. In Proceedings of the 2017 Water Reactor Fuel Performance Meeting, Ramada Plaza Jeju, Jeju Island, Korea, 10–14 September 2017; pp. 10–14.
- Aydogan, E.; Weaver, J.S.; Maloy, S.A.; El-Atwani, O.; Wang, Y.Q.; Mara, N.A. Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations. J. Nucl. Mater. 2018, 503, 250–262.
- Dabney, T.; Johnson, G.; Yeom, H.; Maier, B.; Walters, J.; Sridharan, K. Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding. Nucl. Mater. Energy 2019, 21, 100715.
- Gigax, J.G.; Kennas, M.; Kim, H.; Maier, B.R.; Yeom, H.; Johnson, G.O.; Sridharan, K.; Shao, L. Interface reactions and mechanical properties of FeCrAl-coated Zircaloy-4. J. Nucl. Mater. 2019, 519, 57–63.
- Park, D.J.; Kim, H.G.; Jung, Y.I.; Park, J.H.; Yang, J.H.; Koo, Y.H. Behavior of an improved Zr fuel cladding with oxidation resistant coating under loss-of-coolant accident conditions. J. Nucl. Mater. 2016, 482, 75–82.
- Brachet, J.-C.; Idarraga-Trujillo, I.; Le Flem, M.; Le Saux, M.; Vandenberghe, V.; Urvoy, S.; Rouesne, E.; Guilbert, T.; Toffolon-Masclet, C.; Tupin, M. Early studies on Cr-Coated Zircaloy-4 as enhanced accident tolerant nuclear fuel claddings for light water reactors. J. Nucl. Mater. 2019, 517, 268–285.
- Wang, X.; Guan, H.; Liao, Y.; Zhu, M.; Xu, C.; Jin, X.; Liao, B.; Xue, W.; Zhang, Y.; Bai, G.; et al. Enhancement of high temperature steam oxidation resistance of Zr–1Nb alloy with ZrO2/Cr bilayer coating. Corr. Sci. 2021, 187, 109494.
- Yeom, H.; Sridharan, K. Cold spray technology in nuclear energy applications: A review of recent advances. Ann. Nucl. Energy 2021, 150, 107835.
- Kim, H.-G.; Kim, I.-H.; Jung, Y.-I.; Park, D.-J.; Park, J.-Y.; Koo, Y.-H. Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating. J. Nucl. Mater. 2015, 465, 531–539.
- Huang, M.; Li, Y.; Ran, G.; Yang, Z.; Wang, P. Cr-coated Zr-4 alloy prepared by electroplating and its in situ He+ irradiation behavior. J. Nucl. Mater. 2020, 538, 152240.
- Brachet, J.; Le Saux, M.; Le Flem, M.; Urvoy, S.; Rouesne, E.; Guilbert, T.; Cobac, C.; Lahogue, F.; Rousselot, J.; Tupin, M. On-going studies at CEA on chromium coated zirconium based nuclear fuel claddings for enhanced accident tolerant LWRs fuel. In Proceedings of the TopFuel, Zurich, Switzerlan, 13–19 September 2015; pp. 13–19.
- Wagih, M.; Spencer, B.; Hales, J.; Shirvan, K. Fuel performance of chromium-coated zirconium alloy and silicon carbide accident tolerant fuel claddings. Ann. Nucl. Energy 2018, 120, 304–318.
- Holzwarth, U.; Stamm, H. Mechanical and thermomechanical properties of commercially pure chromium and chromium alloys. J. Nucl. Mater. 2002, 300, 161–177.
- Ma, H.-B.; Yan, J.; Zhao, Y.-H.; Liu, T.; Ren, Q.-S.; Liao, Y.-H.; Zuo, J.-D.; Liu, G.; Yao, M.-Y. Oxidation behavior of Cr-coated zirconium alloy cladding in high-temperature steam above 1200 °C. Npj Mater. Degrad. 2021, 5, 1–11.
- Han, X.; Xue, J.; Peng, S.; Zhang, H. An interesting oxidation phenomenon of Cr coatings on Zry-4 substrates in high temperature steam environment. Corros. Sci. 2019, 156, 117–124.
- Le Saux, M.; Brachet, J.; Vandenberghe, V.; Gilbon, D.; Mardon, J.; Sebbari, B. Influence of pre-transient oxide on LOCA high temperature steam oxidation and post-quench mechanical properties of zircaloy-4 and M5™ cladding. In Proceedings of the Water Reactor Fuel Performance Meeting, paper T3-040, Chengdu, China, 11–14 September 2011.
- Ribis, J.; Wu, A.; Brachet, J.-C.; Barcelo, F.; Arnal, B. Atomic-scale interface structure of a Cr-coated Zircaloy-4 material. J. Mater. Sci. 2018, 53, 9879–9895.
- Jiang, J.; Zhan, D.; Lv, J.; Ma, X.; He, X.; Wang, D.; Hu, Y.; Zhai, H.; Tu, J.; Zhang, W. Comparative study on the tensile cracking behavior of CrN and Cr coatings for accident-tolerant fuel claddings. Surf. Coat. Technol. 2021, 409, 126812.
- Wu, A.; Ribis, J.; Brachet, J.C.; Clouet, E.; Leprêtre, F.; Bordas, E.; Arnal, B. HRTEM and chemical study of an ion-irradiated chromium/zircaloy-4 interface. J. Nucl. Mater. 2018, 504, 289–299.
- Kuprin, A.S.; Belous, V.A.; Voyevodin, V.N.; Vasilenko, R.L.; Ovcharenko, V.D.; Tolstolutskaya, G.D.; Kopanets, I.E.; Kolodiy, I.V. Irradiation resistance of vacuum arc chromium coatings for zirconium alloy fuel claddings. J. Nucl. Mater. 2018, 510, 163–167.
- Kuprin, A.S.; Vasilenko, R.L.; Tolstolutskaya, G.D.; Voyevodin, V.N.; Belous, V.A.; Ovcharenko, V.D.; Kopanets, I.E. Irradiation resistance of chromium coatings for ATFC in the temperature range 300–550 °C. J. Nucl. Mater. 2021, 549, 152908.
- Chen, S.-L.; He, X.-J.; Yuan, C.-X. Recent studies on potential accident-tolerant fuel-cladding systems in light water reactors. Nucl. Sci. Tech. 2020, 31, 1–30.
- Kim, J.-M.; Ha, T.-H.; Kim, I.-H.; Kim, H.-G. Microstructure and oxidation behavior of CrAl laser-coated Zircaloy-4 alloy. Metals 2017, 7, 59.
- Park, D.J.; Jung, Y.I.; Park, J.H.; Lee, Y.H.; Choi, B.K.; Kim, H.G. Microstructural characterization of accident tolerant fuel cladding with Cr–Al alloy coating layer after oxidation at 1200° C in a steam environment. Nucl. Eng. Technol. 2020, 52, 2299–2305.
- Sidelev, D.V.; Kashkarov, E.B.; Syrtanov, M.S.; Krivobokov, V.P. Nickel-chromium (Ni-Cr) coatings deposited by magnetron sputtering for accident tolerant nuclear fuel claddings. Surf. Coat. Technol. 2019, 369, 69–78.
- Jin, D.; Yang, F.; Zou, Z.; Gu, L.; Zhao, X.; Guo, F.; Xiao, P. A study of the zirconium alloy protection by Cr3C2–NiCr coating for nuclear reactor application. Surf. Coat. Technol. 2016, 287, 55–60.
- Sridharan, K.; Harrington, S.; Johnson, A.; Licht, J.; Anderson, M.; Allen, T. Oxidation of plasma surface modified zirconium alloy in pressurized high temperature water. Mater. Des. 2007, 28, 1177–1185.
- Kim, H.-G.; Kim, I.-H.; Jung, Y.-I.; Park, D.-J.; Park, J.-Y.; Koo, Y.-H. Microstructure and mechanical strength of surface ODS treated Zircaloy-4 sheet using laser beam scanning. Nucl. Eng. Technol. 2014, 46, 521–528.
- Kim, H.-G.; Yang, J.-H.; Kim, W.-J.; Koo, Y.-H. Development status of accident-tolerant fuel for light water reactors in Korea. Nucl. Eng. Technol. 2016, 48, 1–15.
- Yeh, J.W.; Chen, S.K.; Lin, S.J.; Gan, J.Y.; Chin, T.S.; Shun, T.T.; Tsau, C.H.; Chang, S.Y. Nanostructured high-entropy alloys with multiple principal elements: Novel alloy design concepts and outcomes. Adv. Eng. Mater. 2004, 6, 299–303.
- Fereiduni, E.; Ghasemi, A.; Elbestawi, M. Characterization of composite powder feedstock from powder bed fusion additive manufacturing perspective. Materials 2019, 12, 3673.
- Do, H.-S.; Lee, B.-J. Origin of radiation resistance in multi-principal element alloys. Sci. Rep. 2018, 8, 1–9.
- Gao, L.; Liao, W.; Zhang, H.; Surjadi, J.U.; Sun, D.; Lu, Y. Microstructure, mechanical and corrosion behaviors of CoCrFeNiAl0. 3 high entropy alloy (HEA) films. Coatings 2017, 7, 156.
- Zhang, W.; Tang, R.; Yang, Z.; Liu, C.; Chang, H.; Yang, J.; Liao, J.; Yang, Y.; Liu, N. Preparation, structure, and properties of an AlCrMoNbZr high-entropy alloy coating for accident-tolerant fuel cladding. Surf. Coat. Technol. 2018, 347, 13–19.
- Zhang, W.; Wang, M.; Wang, L.; Liu, C.; Chang, H.; Yang, J.; Liao, J.; Yang, Y.; Liu, N. Interface stability, mechanical and corrosion properties of AlCrMoNbZr/(AlCrMoNbZr) N high-entropy alloy multilayer coatings under helium ion irradiation. Appl. Surf. Sci. 2019, 485, 108–118.
- Zhang, W.; Tang, R.; Yang, Z.; Liu, C.; Chang, H.; Yang, J.; Liao, J.; Yang, Y.; Liu, N. Preparation, structure, and properties of high-entropy alloy multilayer coatings for nuclear fuel cladding: A case study of AlCrMoNbZr/(AlCrMoNbZr) N. J. Nucl. Mater. 2018, 512, 15–24.
- Tao, Z.; Wang, P.; Wang, C.; Ma, Z.; Zhang, Y.; Xue, F.; Bai, G.; Yuan, Y.; Lan, R. Design and characterisation of AlCrFeCuNbx alloys for accident-tolerant fuel cladding. J. Alloys Compd. 2021, 859, 157805.
- Zhang, Z.; Han, E.-H.; Xiang, C. Irradiation behaviors of two novel single-phase bcc-structure high-entropy alloys for accident-tolerant fuel cladding. J. Mater. Sci. Technol. 2021, 84, 230–238.
- Pickering, E.J.; Carruthers, A.W.; Barron, P.J.; Middleburgh, S.C.; Armstrong, D.E.; Gandy, A.S. High-Entropy Alloys for Advanced Nuclear Applications. Entropy 2021, 23, 98.
- Yun, D.; Lu, C.; Zhou, Z.; Wu, Y.; Liu, W.; Guo, S.; Shi, T.; Stubbins, J.F. Current state and prospect on the development of advanced nuclear fuel system materials: A review. Mater. Rep. Energy 2021.
- Khatkhatay, F.; Jiao, L.; Jian, J.; Zhang, W.; Jiao, Z.; Gan, J.; Zhang, H.; Zhang, X.; Wang, H. Superior corrosion resistance properties of TiN-based coatings on Zircaloy tubes in supercritical water. J. Nucl. Mater. 2014, 451, 346–351.
- Kashkarov, E.; Nikitenkov, N.; Sutygina, A.; Bezmaternykh, A.; Kudiiarov, V.; Syrtanov, M.; Pryamushko, T. Hydrogenation behavior of Ti-implanted Zr-1Nb alloy with TiN films deposited using filtered vacuum arc and magnetron sputtering. Appl. Surf. Sci. 2018, 432, 207–213.
- Tunes, M.A.; Da Silva, F.C.; Camara, O.; Schön, C.G.; Sagás, J.C.; Fontana, L.C.; Donnelly, S.E.; Greaves, G.; Edmondson, P.D. Energetic particle irradiation study of TiN coatings: Are these films appropriate for accident tolerant fuels? J. Nucl. Mater. 2018, 512, 239–245.
- Khatkhatay, F.; Jian, J.; Jiao, L.; Su, Q.; Gan, J.; Cole, J.I.; Wang, H. Diffusion barrier properties of nitride-based coatings on fuel cladding. J. Alloys Compd. 2013, 580, 442–448.
- Xiao, W.; Chen, H.; Liu, X.; Tang, D.; Deng, H.; Zou, S.; Ren, Y.; Zhou, X.; Lei, M. Thermal shock resistance of TiN-, Cr-, and TiN/Cr-coated zirconium alloy. J. Nucl. Mater. 2019, 526, 151777.
- Alat, E.; Motta, A.T.; Comstock, R.J.; Partezana, J.M.; Wolfe, D.E. Ceramic coating for corrosion (c3) resistance of nuclear fuel cladding. Surf. Coat. Technol. 2015, 281, 133–143.
- Alat, E.; Motta, A.T.; Comstock, R.J.; Partezana, J.M.; Wolfe, D.E. Multilayer (TiN, TiAlN) ceramic coatings for nuclear fuel cladding. J. Nucl. Mater. 2016, 478, 236–244.
- Alat, E.; Brova, M.; Younker, I.; Motta, A.; Fratoni, M.; Wolfe, D. Neutronic and mechanical evaluation of rare earth doped and undoped nitride-based coatings for accident tolerant fuels. J. Nucl. Mater. 2019, 518, 419–430.
- Daub, K.; Van Nieuwenhove, R.; Nordin, H. Investigation of the impact of coatings on corrosion and hydrogen uptake of Zircaloy-4. J. Nucl. Mater. 2015, 467, 260–270.
- Van Nieuwenhove, R.; Andersson, V.; Balak, J.; Oberländer, B. In-Pile testing of CrN, TiAlN, and AlCrN coatings on zircaloy cladding in the Halden reactor. ASTM Int. 2018, 965–982.
- Meng, C.; Yang, L.; Wu, Y.; Tan, J.; Dang, W.; He, X.; Ma, X. Study of the oxidation behavior of CrN coating on Zr alloy in air. J. Nucl. Mater. 2019, 515, 354–369.
- Krejčí, J.; Ševeček, M.; Cvrcek, L.; Kabátová, J.; Manoch, F. Chromium and chromium nitride coated cladding for nuclear reactor fuel. In Proceedings of the 20th International Corrosion Congress, EUROCORR, Prague, Czech Republic, 3–7 September 2017.
- Krejčí, J.; Ševeček, M.; Kabátová, J.; Manoch, F.; Kočí, J.; Cvrček, L.; Málek, J.; Krum, S.; Šutta, P.; Bublíková, P. Experimental behavior of chromium-based coatings. In Proceedings of the TopFuel, Prague, Czech Republic, 30 September–4 October 2018; pp. 1–13.
- Song, J.; Huang, Z.; Qin, Y.; Wang, H.; Shi, M. Effects of Zirconium Silicide on the Vulcanization, Mechanical and Ablation Resistance Properties of Ceramifiable Silicone Rubber Composites. Polymers 2020, 12, 496.
- Yeom, H.; Maier, B.; Mariani, R.; Bai, D.; Fronek, S.; Xu, P.; Sridharan, K. Magnetron sputter deposition of zirconium-silicide coating for mitigating high temperature oxidation of zirconium-alloy. Surf. Coat. Technol. 2017, 316, 30–38.
- Yeom, H.; Lockhart, C.; Mariani, R.; Xu, P.; Corradini, M.; Sridharan, K. Evaluation of steam corrosion and water quenching behavior of zirconium-silicide coated LWR fuel claddings. J. Nucl. Mater. 2018, 499, 256–267.
- Terrani, K.A.; Pint, B.A.; Parish, C.M.; Silva, C.M.; Snead, L.L.; Katoh, Y. Silicon carbide oxidation in steam up to 2 MPa. J. Am. Ceram. Soc. 2014, 97, 2331–2352.
- Lee, K.; Kim, D.; Yoon, Y.S. SiC/Si thin film deposited on zircaloy to improved accident tolerant fuel cladding. Thin Solid Film. 2018, 660, 221–230.
- Kashkarov, E.B.; Syrtanov, M.S.; Murashkina, T.L.; Kurochkin, A.V.; Shanenkova, Y.; Obrosov, A. Hydrogen sorption kinetics of SiC-coated Zr-1Nb alloy. Coatings 2019, 9, 31.
- Usui, T.; Sawada, A.; Amaya, M.; Suzuki, A.; Chikada, T.; Terai, T. SiC coating as hydrogen permeation reduction and oxidation resistance for nuclear fuel cladding. J. Nucl. Sci. Technol. 2015, 52, 1318–1322.
- Bao, W.; Xue, J.; Liu, J.-X.; Wang, X.; Gu, Y.; Xu, F.; Zhang, G.-J. Coating SiC on Zircaloy-4 by magnetron sputtering at room temperature. J. Alloys Compd. 2018, 730, 81–87.
- Al-Olayyan, Y.; Fuchs, G.; Baney, R.; Tulenko, J. The effect of Zircaloy-4 substrate surface condition on the adhesion strength and corrosion of SiC coatings. J. Nucl. Mater. 2005, 346, 109–119.
- Baney, R.H.; Butt, D.; Demkowicz, P.; Tulenko, J.S. An Innovative Ceramic Corrosion Protection System for Zircaloy Cladding; University of Florida (US): Gainesville, FL, USA, 2003.
- Tang, C.; Steinbrueck, M.; Grosse, M.; Ulrich, S.; Stueber, M.; Seifert, H.J. Evaluation of magnetron sputtered protective Zr-C-Al coatings for accident tolerant Zircaloy claddings. In Proceedings of the Water Reactor Fuel Performance Meeting, Jeju Island, Korea, 10–14 September 2017.
- Michau, A.; Maury, F.; Schuster, F.; Lomello, F.; Brachet, J.-C.; Rouesne, E.; Le Saux, M.; Boichot, R.; Pons, M. High-temperature oxidation resistance of chromium-based coatings deposited by DLI-MOCVD for enhanced protection of the inner surface of long tubes. Surf. Coat. Technol. 2018, 349, 1048–1057.
- Michau, A.; Maury, F.; Schuster, F.; Nuta, I.; Gazal, Y.; Boichot, R.; Pons, M. Chromium carbide growth by direct liquid injection chemical vapor deposition in long and narrow tubes, experiments, modeling and simulation. Coatings 2018, 8, 220.
- Michau, A.; Gazal, Y.; Addou, F.; Maury, F.; Duguet, T.; Boichot, R.; Pons, M.; Monsifrot, E.; Maskrot, H.; Schuster, F. Scale up of a DLI-MOCVD process for the internal treatment of a batch of 16 nuclear fuel cladding segments with a CrCx protective coating. Surf. Coat. Technol. 2019, 375, 894–902.
- Yang, Z.; Niu, Y.; Xue, J.; Liu, T.; Chang, C.; Zheng, X. Steam oxidation resistance of plasma sprayed chromium-containing coatings at 1200 °C. Mater. Corros. 2019, 70, 37–47.
- Lyu, J.; Kashkarov, E.; Travitzky, N.; Syrtanov, M.; Lider, A. Sintering of MAX-phase materials by spark plasma and other methods. J. Mater. Sci. 2020, 56, 1–36.
- Barsoum, M.W. The MN+ 1AXN phases: A new class of solids: Thermodynamically stable nanolaminates. Prog. Solid State Chem. 2000, 28, 201–281.
- Barsoum, M.W.; El-Raghy, T. The MAX phases: Unique new carbide and nitride materials: Ternary ceramics turn out to be surprisingly soft and machinable, yet also heat-tolerant, strong and lightweight. Am. Sci. 2001, 89, 334–343.
- Garcia-Diaz, B.; Olson, L.; Verst, C.; Sindelar, R.; Hoffman, E.; Hauch, B.; Maier, B.; Sridharan, K. MAX phase coatings for accident tolerant nuclear fuel. Trans. Am. Nucl. Soc 2014, 110, 994–996.
- Tallman, D.J.; Anasori, B.; Barsoum, M.W. A critical review of the oxidation of Ti2AlC, Ti3AlC2 and Cr2AlC in air. Mater. Res. Lett. 2013, 1, 115–125.
- Li, W.; Wang, Z.; Shuai, J.; Xu, B.; Wang, A.; Ke, P. A high oxidation resistance Ti2AlC coating on Zirlo substrates for loss-of-coolant accident conditions. Ceram. Int. 2019, 45, 13912–13922.
- Maier, B.R.; Garcia-Diaz, B.L.; Hauch, B.; Olson, L.C.; Sindelar, R.L.; Sridharan, K. Cold spray deposition of Ti2AlC coatings for improved nuclear fuel cladding. J. Nucl. Mater. 2015, 466, 712–717.
- Tang, C.; Steinbrueck, M.; Stueber, M.; Grosse, M.; Yu, X.; Ulrich, S.; Seifert, H.J. Deposition, characterization and high-temperature steam oxidation behavior of single-phase Ti2AlC-coated Zircaloy-4. Corros. Sci. 2018, 135, 87–98.
- Roberts, D.A. Magnetron Sputtering and Corrosion of Ti-Al-C and Cr-Al-C Coatings for Zr-Alloy Nuclear Fuel Cladding. Master’s Thesis, University of Tennessee, Knoxville, TN, USA, 2016.
- Tang, C.; Steinbrueck, M.; Grosse, M.; Ulrich, S.; Stueber, M.; Seifert, H.J. Improvement of the high-temperature oxidation resistance of Zr alloy cladding by surface modification with aluminum-containing ternary carbide coatings. In Proceedings of the 2018 International Congress on Advances in Nuclear Power Plants, Charlotte, NC, USA, 8–11 April 2018.
- Imtyazuddin, M.; Mir, A.H.; Tunes, M.A.; Vishnyakov, V.M. Radiation resistance and mechanical properties of magnetron-sputtered Cr2AlC thin films. J. Nucl. Mater. 2019, 526, 151742.
- Tang, C.; Grosse Karl, M.; Trtik, P.; Steinbrück, M.; Stüber, M.; Seifert, H.J. H2 Permeation Behavior of Cr2AlC and Ti2AlC MAX Phase Coated Zircaloy4 by Neutron Radiography. Acta Polytech. 2018, 58, 69–76.
- Wang, Z.; Ma, G.; Liu, L.; Wang, L.; Ke, P.; Xue, Q.; Wang, A. High-performance Cr2AlC MAX phase coatings: Oxidation mechanisms in the 900–1100 C temperature range. Corros. Sci. 2020, 167, 108492.
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