1. Introduction

The growing energy demands worldwide together with the urgent need to manage atmospheric greenhouse gas emissions drive an increasing necessity for sustainable electricity production. Nuclear energy provides a reliable electricity supply, characterized by its very low carbon emissions, and the relative small waste generation, which could safely be managed, stored and disposed. During the past 40 years, nuclear fuel, which is highly complex material, has subjected to a continuous improvement, leading to a stage of development that enabled its safe irradiation to up to 65 GWd·MTU⁻¹ in commercial nuclear power reactors. During this time there have been many improvements to the original designs and materials used in nuclear power plants. However, the basic concept of the uranium oxide fuel pellets claded by zirconium (Zr) based alloy tubes has remained the fuel of choice for the vast majority of commercial nuclear power plants [1]. Today, Light Water Reactors (LWRs) commercial operation around the world relies on this fuel system (i.e. uranium dioxide (UO₂) encapsulated within a Zr-based alloy cladding) of which, an example is shown in Fig. 1. Uraniumplutonium oxide fuel, widely known as mixed oxide (MOX), is also used in some types or reactors.

The reactor core environment is extremely harsh, nuclear materials inside the core are exposed to a combination of high stresses (both thermal and mechanical), aggressive coolant medium, and extremely high radiation doses. Accordingly, nuclear fuel claddings are key reactor components for fuel integrity and plant safety as a whole. Due to favorable characteristics such as small cross section for neutrons absorption, and good mechanical performance, Zr-based alloys have been chosen as fuel cladding material, and it has adequately served the nuclear industry over the past decades. As the industry expanded, failure and damage rates of nuclear fuel claddings decreased significantly “from an initial rates of defect of ~1/100 fuel rods in late 1960s to less than 1/100,000 in 2005” [2]. This contributed to the optimization of nuclear fuel cycle and cladding design.

Technological improvements, with relation to nuclear materials and fuels, remain a key to the industry’s safety, feasibility and profitability. Previously, the emphasis of LWRs fuel Research and Development (R&D) was focusing on the improvement of fuel performance in terms of increased fuel burn up values, spent fuel waste minimization, power density and power upgrading, and operational life extension to achieve the economical feasibility.

The Fukushima accident in Japan in 2011, triggered by a 9.0 magnitude earthquake and followed by a tsunami, had worldwide devastating consequences on the status of nuclear power utilization potentials. Following to this accident, a renewed interest was generated to address flaws and weaknesses of the traditional UO₂-Zr-based fuel claddings system under accidental conditions. Thus, the post Fukushima research has shifted towards the concept of “Accident tolerant Fuel (ATF)” [3-5], a nuclear fuel concept which is, when compared to the traditional UO₂-Zr nuclear fuel system, can considerably tolerate the loss of reactor’s core active cooling for longer times during accidents scenarios, with no need for the swift operator intervention. In addition, this fuel concept is also expected to maintain and improve normal reactor operation performance [6] and therefore, ensuring safety and economic viability of this proposed new fuels [3]. Among others, the improvements in cladding performance within the ATF frame shall include the reduction of oxidation rates and hydrogen production at high temperatures, prevention of clad fracture, and hydrogen embrittlement. Cladding internal oxidation is also among the issues that need to be addressed in establishing the ATF attributes [3]. Improvement of nuclear fuel itself is also under investigation; areas such as the better retention of fission gases and the increased margin for fuel melting temperature are being researched. Currently, many active programs on ATFs exist and are being carried out by companies, institutions and organizations worldwide. Different concepts were proposed and are under investigation by these programs, such as ATF fuel concepts, cladding concepts and non-fuel materials related concepts.

Numerous ATF cladding concepts have been proposed at present, with a primary purpose of Zr-based alloys performance improvement. These concepts are mainly categorized in two groups: replacing Zr-based cladding with new materials, and the use of protective coatings. Based on the anticipated time required and total costs for development of new clad materials, the first option is being seen as a long-term strategy. Protective coatings on the other hand are expected to offer better corrosion and wear resistance, and to reduce Zr-based alloys hydrogen absorption as well, with no significant modifications introduced to the original design, thus, coated claddings could be put into commercial application within a relatively short time frame.

This review focuses on the various aspects associated with the proposed ATF coatings concepts, including the development status, their performance under normal operation conditions and accident scenarios their challenges and deposition technologies.

2. Zirconium Based Alloys

Zirconium-based alloys were first developed in the United States Navy during the 1950s [8,9]. Zr-alloys used in nuclear industry today contain more that 95% Zr, and hence their characteristics are generally close to those of pure Zr metal. Pure Zr has poor corrosion resistance, resulting in high cracking and spallation of its oxide [10]. Traces of alloying elements are optimized over decades to improve the performance of these alloys. Currently, Zr is mainly alloyed with niobium (Nb) and tin (Sn) to improve corrosion resistance. Iron (Fe), chromium (Cr), and nickel (Ni) are also added to improve mechanical and functional properties. Zr-Sn based alloy system is favored by the US industry, while in Russia, Canada and some other countries Zr-Nb system was chosen as the preferred [11-13]. Table 1 summarizes main compositions for Zr alloys currently exists in the nuclear industry.

Zircaloy-2 have been utilized in the Boiling Water Reactors type (BWRs) as a fuel cladding material and it belongs to the Zr-Sn alloying system, which shows good balance between enhanced corrosion resistance and good mechanical properties. Fe, Cr and Ni utilization resulted in an improved oxidation resistance in zircaloy-2 alloys, however, Ni presence induced higher hydrogen uptakes, therefore, the addition of Ni is avoided in developing zircaloy-4, and instead, the Fe content was increased.

Currently, the main types of zirconium alloys utilized in nuclear industry are: zircaloy-2 in BWRs and zircaloy-4, used as Pressurized Water Reactors (PWRs) fuel cladding material [14]. E110, E635 and E125 alloys are mainly characterized by the use of Nb and they are generally used as a material for the fabrication of fuel claddings, spacer grids, and guide tubes in the Russian VVER and RBMK types of reactors. Both E110 and E635 alloys have been developed with a good creep and irradiation growth resistance. The E110 alloy is characterized by its superior corrosion resistance in PWRs, while E635 has high strength, resistance to creep and growth in BWRs environment [15,16]. E125 alloy is usually used in RBMK and VVERs for shroud pressure pipes. However, the E-series of alloys are generally found to be more sensitive to breakaway corrosion compared to zircaloy-4, leading to increased oxidation and hydrogen uptake rates [17,18]. Thus, improved versions of these alloys (such as E125 opt. E110 opt., E635M and E110M) doped with Fe and O were introduced to over an absolute integrity upon utilization in future VVERs generations, designed to operate for higher fuel burn-ups [19,20].

On the other hand, ZIRLO was designed by Westinghouse as a PWR’s fuel cladding material; it includes dopants of Sn, Fe, Nb and O. An optimized version named OPT ZIRLO, which has a reduced Sn (0.66wt%) ratio and similar Fe content (0.10wt%–0.11wt%) was designed in 2004. In comparison with ZIRLO alloy, OPT ZIRLO is developed to possess higher resistance to corrosion and creep, particularly in media containing Li [21].

The French alloy M5 is being commercially utilized since the 1990s mainly as base cladding material in PWRs in the French nuclear industry. Both E110 alloy and M5 have similar basic constitution of Zr-1%Nb that alloyed with controlled amount of other dopants. An optimized process of heat treatment in M5 alloy fabrication led to a very good resistance to corrosion and improved mechanical properties under the harsh reactor conditions represented by high temperatures and high irradiation doses [22-24].

The High performance Alloy for Nuclear Application (HANA) was developed at the Korea Atomic Energy Research Institute (KAERI) to be used in the fuel assemblies, and it has been shown that these alloys possess better corrosion, High Temperature (HT) oxidation and creep performances [25,26].

2.1 Performance of Zirconium Alloys Under Normal and Accident Conditions

In both PWRs and BWRs, Light water is used as coolant and moderator. Fig. 2 shows the light water phase diagram in which, the two points that represent the conditions of normal operation for the PWRs and BWRs are showed [27]. The PWRs normal operation conditions in the core are ~330°C and 15.5 MPa while for BWR it is normally ~ 285°C and 7.5 MPa [28]. Due to the reactor core high temperature and pressure, the water is in a subcritical state, acting consequently as a highly corrosive medium. Above the critical points (374°C, 22 MPa), the water is in supercritical state [29].

Corrosion of the Waterside of Zr-based alloys in normal operation conditions induces a thin oxide scale growth on cladding’s external surface. The kinetics for the weight gain generally falls into two stages, being referred to as the pre- and the post-transitions. A black, thin and adherent oxide scale composed of microcrystalline grains will form initially during the pre-transition stage with a kinetic that follows cubic growth law, and it is governed by the oxidizing species diffusion through the oxide layer [30,31]. Once exceeded a critical certain thickness (~2−3 μm), the growth kinetics of the oxide scale increases to post-transition linear rate, in which, the oxide scale containing large porosities will be cracked [31]. The thickness of oxide scale eventually reaches tens of microns at the end of the fuel cycle [13]. Another concern associated with Zr based alloys during normal reactors’ operation is the hydrogen pickup. Hydrogen normally produced by the water radiolysis or as a result of cladding corrosion. Fractions of hydrogen ions penetrate the oxide scale and dissolve in the metallic matrix. Finally, it precipitates as hydrides (ZrHx) once its content exceeded the solubility limits. As a result, cladding ductility and toughness reductions occurs [31], and this phenomenon can significantly affect cladding performance during accidental scenarios.

The behavior of Zr-based alloys under Design Basis Accidents (DBAs) and the Beyond Design Basis Accidents (BDBAs) is widely and extensively investigated by many authors [32-41], such events could typically be triggered by a loss of coolant accident (LOCA). The decay heat and stored energy in the fuel could rapidly drive up core temperature, and the Zr-based claddings are consequently oxidized by the high-temperature steam, eventually undergoes ballooning and burst in temperature range of 700°C– 1,200°C [37]. In DB-LOCA, Emergency Core Cooling System (ECCS) will be activated and quenches the fuel by the injection of cooling water into the reactor core. Generally, in accordance with the US regulatory commission (NRC) requirements, cladding’s peak temperature should be limited to 1,204°C and the equivalent cladding reacted (ECR) to 17% of its initial thickness, in order to avoid severe degradations [28]. In BDBA scenarios with having ECCS lost or failed, Zr-based fuel claddings exhibit accelerated oxidation once its temperature exceeded 1,200°C. The oxide layer formed during normal operation conditions will not be/or be weakly beneficial in providing sufficient protection in such situations, when HT steam exposures take place [38]. The overall reaction of Zr with the high temperature steam (exothermic) is given by the equation:

This Zr-based alloys low resistance to HT steam oxidation in accident conditions was addressed in several studies, such as that done by Malgin et al. [39], where HT oxidation behavior of E110opt and E110M alloys in steam (~900−1,250°C) were tested. Their results showed similar oxidation kinetics for both alloys. Yan et al. [41] showed the oxidation kinetics acceleration associated with the increased temperature in the sponge-based E110 alloy, and at 1,200°C, a good agreement with the Cathcart-Pawel (CP) model [40] was observed (Fig. 3) [41].

3. Overview of ATF Cladding Concepts

Many ATF cladding concepts were proposed to improve Zr-based alloys performance by different research groups and active projects worldwide. These concepts mainly fall into two routes: replacing the current Zr-based cladding material with new cladding materials [1,42,43], along with the use of protective coatings on Zr-based fuel claddings surfaces [28,44]. Proposed new cladding materials such as silicon carbide and molybdenum based alloys generally entail large investments and development time, therefore this option is being seen as a long-term strategy. On the other hand, Protective coatings are expected to improve corrosion and HT oxidation resistance, and reduce hydrogen absorption on Zr-based alloys as well [44], with no need to introduce significant design modifications, and hence its considered as a short term strategy, enabling the coated claddings to be put into commercial application within a relatively short time frame [1]. An example of the Potential performance and the anticipated development time is shown by Fig. 4 for AREVA/CEA/EDF investigated cladding concepts.

4. Protective Coatings

The adoption of protective coatings on Zr-based alloys surfaces is an evolutionary approach considered by some ATF programs. Major impacts on claddings thermo mechanical behavior are not expected when thin films will be applied, assuming an engineered adequate creep and limited strain mismatch into the coatings. In fact, Coatings are thought to have an advantage relative to other ATF claddings concepts. Mechanical and Neutronic properties of Zr alloys in particular, are expected to be conserved for the coated rods, given that the coating will have an adequately small thickness, relative to the total clad thickness. Surface coatings can leverage Decades of research and development in the industry offered by academia, national governmental laboratories, and the relevant R&D institutions worldwide, to potentially reduce the regulatory burdens associated with the demonstration of their safe use in LWRs, provided that a large amount of data already existing, and the Zr-based cladding response in LWR environments is well understood. Nevertheless, a critical disadvantage may arise with relation to this concept, the Enhanced accident tolerance will entirely depend on the coating itself; therefore, their protective nature will be compromised by any event or process that would significantly damage or otherwise alter coating’s physical integrity [45].

Therefore, requirements such as Coatings adhesion and chemical stability during normal and off-normal operation conditions are considered of crucial importance, to provide the necessary protection from rapid oxidation during BDBAs. Furthermore, coatings morphology must not drastically change, in a way that could lead to exposures of significant un-reacted Zr amounts during transients. Finally, an industrial scale deposition of these coatings will be required along the entire fuel rods with the necessary quality assurance measures; and hence, a technological challenge is introduced.

It is necessary for the Proposed ATF coating materials to contain one of the elements that exhibit high temperature steam oxidation resistance such as chromia, alumina, and silica formers [44]. Several organizations have initiated studies on protective coatings over the past five years under the concept of enhancing the accident tolerance. A large range of critical factors such as coating adhesion, radiation resistance, absorption cross-section for thermal neutrons, thermal conductivity, and mechanical properties [44,46], which can affect the behavior of coated Zr-based claddings both in normal conditions and accident events, are being studied. Currently, numerous of these studies are working to assess coated Zr-based alloys’ performance, by the deposition of metallic (Cr, Cr-Al, Ni-Cr, Fe-based alloys, etc.), non-metallic (oxides, nitrides, carbides) and MAX-phase [44] coatings. The studied coatings’ materials broadly categorized within two groups [47]:

Nitride coatings can harden and improve wear behavior of materials, especially TiN and TiAlN. The MAX phases, a group of layered ternary carbides and nitrides, have attracted a great attention recently due to its unique properties which combine both metallic and ceramic characteristics. MAX phases are being considered as both cladding material and protective coatings in LWRs.

To increase the HT strength of Zr based substrates, Surface treatments as a complement to coating deposition have also been investigated by KAERI; through the formation of oxide dispersion strengthened surface layer (ODS), incorporating Y₂O₃ Nano-particles. Hence, this concept adopted by KAERI is considered more complex as compared to other concepts featuring only one external layer, due to adoption of the two surface treatments for HT steam oxidation performance and HT strength [45].

4.1 Current Development Status

To date, coatings that from chromia represent the most widely explored coating technology [51]. Metallic Cr, CrAl, and CrN coatings have been studied extensively. In the case of Cr coatings with thickness of few to tens microns, chromia formed under both aqueous or HT steam conditions is found to be able to protect the underlying Zr substrate from HT steam oxidation, as shown by Fig. 5 in which 12–15 μm Cr coating deposited on Zr-based cladding surface, the analysis showed a fully dense and homogeneous coating with a good metallurgical bonding of Zr–Cr interface and no indications of voids or cracks [48,49,51], therefore, adequate corrosion protection for the substrate is provided.

Furthermore, resistance to cladding post-quench ductility loss has been reported for these coatings. Multiple in-pile experiments for further performance evaluation for these concepts are being conducted, with preliminary results indicating adequate behavior under ion irradiation [50]. As a body-centered cubic (BCC) metal, metallic Cr is expected to exhibit dimensional stability during neutron irradiation [52] at LWRs temperature range.

Excellent stability has been demonstrated for thin (< 5 μm) CrN coatings on the surface of Zr-based cladding under prototypical fuel irradiation conditions [53]. Integral LOCA testing for the un-irradiated CrN coated cladding demonstrated excellent adhesion even after burst testing, but showed no improvement in oxidation or burst behavior compared to the uncoated cladding sample. Although the adverse effects of N during Zr-based alloys air oxidation are well understood, a small quantity of this element in the thin coating is not expected to cause large degradation of the cladding during HT steam oxidation [52].

Yttrium oxidation resistance has also attracted the attention of some researchers. At very high temperatures, stable protective yttrium oxides are formed, making it very attractive for study as potential coating material. The investigation of Y-coated Zr-based alloys surface is conducted by Sridharan et al. [54] and Kim et al. [55], in the former study the authors observed an oxidation rate decline in 400°C supercritical water for 168 hrs. While in the later, the surface of a 2 mm Zircaloy-4 substrate is modified through the deposition of 10 μm Y₂O₃ using laser beam scanning method, in order to produce an ODS treatment. The result showed that the Zr-based alloy with the ODS layer possessed a 65% higher yield strength compared to uncoated alloy at 500°C, hence, an enhanced high temperature strength to improve the ballooning behavior of fuel cladding during accidental events is expected [56]. However, this coating type has a limited data in the literature, making it difficult to judge their feasibility of use as ATF coating material and more research is needed.

The same oxide scale that protects Cr-, Si and Al-bearing coatings at high temperatures are also formed in LWRs coolant aqueous environment [11]. Unfortunately, chromia is the only one that showed stability in this environment, while alumina and silica tend to dissolve rapidly as aluminum oxy-hydroxide (AlOOH) and Silicic acid (H₄SiO₄) [57]. Incorporation of titanium (Ti), which has the ability to form a stable oxide into these coatings, can mitigate the dissolution (e.g., TiN/TiAlN); nevertheless, at elevated temperatures, Ti undergoes rapid oxidation and its dominance in the coating will probably compromise the protectiveness of silica/alumina [51,58].

Ti₂AlC [59,60], Cr₂AlC [61], TiAlN [62], and Ti₃SiC₂ [63] as MAX-phase coatings have also been studied; nevertheless, none of the studies performed have produced a complete assessment of their performance under normal operation, DBA, and BDBA conditions. It has been shown in a recent study that some MAX phase coatings are resistant to heavy ions and neutron irradiations, its crystallinity and phase stability were found to be maintained up to high irradiation doses [64]. The control of stoichiometry can be considered as a key requirement for manufacturing MAX phase coatings, yet it is often represents a challenging issue, due to the ternary nature of MAX phase coatings with a large number of other potential compounds possible to exist with the same elements [45].

4.2 Influence of Coating Thickness

The thickness of coating materials plays major role in preserving Zr substrate properties and behavior. When adequately thin coatings are applied (below 20 μm), surface properties of nuclear fuel cladding are primarily expected to be modified by the application of these coatings on Zrbased substrates, rather than the behavior of the cladding itself. Generally, for thicknesses below 20 μm, the impact on Neutronic and fuel cycle costs or nuclear fuel cycle length for the majority of studied types of coatings (Cr, Cr-Al, CrN, FeCrAl, MAX phase, etc.) is expected to be small and may easily be compensated for by slight design modifications [65,66]. Since the proposed coating materials mostly have higher thermal neutron absorption cross-sections compared to that of zirconium, coatings with higher thickness will have a negative impact on fuel cycle economics, therefore considerations must be given to thickness of coatings as one of the key factors in normal operation performance evaluations, and a compromise will likely be needed in this term.

Additionally, thicker coatings from materials characterized with low thermal conductivity (ceramic coatings), would deteriorate the heat transfer from the pellet through cladding into the coolant, which may lead to an increased centerline temperature in the fuel. In the case of multi-layered coatings, thermal conductivity may be affected by the number of interfaces as it often represents potential barriers. Therefore, this will have to be thoroughly investigated for multi-layered coatings [45].

The influence of coating thickness on the oxidation resistance has been studied by some researchers. An autoclave analysis of HT oxidation tests demonstrated that at least 10 μm of Cr coatings must be used to provide long-term protection [44]. Comparative analysis was performed by Brachet et al. [67] on the mass gains for zircaloy- 4 samples coated by 1–12 μm Cr coatings deposited by PVD method and oxidized in HT steam at 1,100°C for 850 s. Their results showed mass gain reductions in samples coated by Cr, this reduction varied from 6.5 mg·cm⁻² to < 1 mg·cm⁻², with the increase in coating thickness. Kashkarov et al. [68] studied the kinetics of oxidation for Cr coatings on E110 alloy in HT steam (1,200°C) for 10 min. they showed that Cr coated samples experienced weight gain reductions from 22.1 to 4.1 mg·cm⁻², corresponding to increase of coating thickness from 4.5 to 9.0 μm. Similar observation was recorded by Kashkarov et al. [69] for Cr coated E110 alloy after HT air oxidation tests at 1,100–1,200°C with magnetron sputtering method used for Cr coatings deposition.

It has been observed by Brachet et al. [70] that Cr coated Zr alloys oxidation kinetics accelerates post a transition period, whereby the state of coating changes from protective to non-protective (Fig. 6(a)). Protective period, and transition time durations are mainly dependent on the thickness of coatings (Fig. 6(b)).

4.3 Coatings Performance in Normal Operation & Anticipated Operational Occurrences Conditions

4.3.1 Mechanical Performance

Using thin protective coatings entails that the substrate is governing cladding mechanical properties rather than the coating material. Therefore, mechanical properties and behavior of the coated cladding are expected to be close to that of uncoated cladding especially when the used coatings are not too thick and the deposition technique does not modify the Zr alloy microstructure. In presence of the phenomenon of material hardening under irradiation, the mechanical behavior after irradiation is particularly significant. One of the aims of lead test fuel rods and assemblies’ irradiation will be to study and verify possible impacts of the coating on creep and growth under irradiation. However, some out-of-pile results concerning the mechanical behavior of various coatings are available and will be discussed here for both categories of metallic and ceramic coatings.

Most of the proposed coating materials are harder than Zr-based alloys (such as Cr, Cr-Al, FeCrAl, CrN and MAX phase), hence sufficiently thick (> 30 μm) coatings will modify mechanical properties with increased strength and reduced ductility. In KAERI concept, the adopted surface treated ODS Zr alloy combination with the relatively thicker coating has led to an increased strength and a reduced ductility [56]. Additionally, thicker coatings under irradiation may potentially lead to even lower ductility, due to irradiation hardening of the coating material. However, this increased hardness has the advantage of potentially protecting the cladding against wear and fretting, and some preliminary studies showed that Cr coatings are quiet protective against cladding wear and it may significantly reduce the risk for cladding damage caused by debris or grid-torod fretting. It is also worth to mention that the existent knowledge on the industrial scale indicates that CrN coatings are able to improve wear behavior.

4.3.1.1 Metallic Coatings

It has been shown by Bischoff et al. [71] that Cr coated M5 cladding possessed similar mechanical behavior as that of uncoated M5 in un-irradiated condition (Fig. 7). AREVA/CEA/EDF studied the mechanical behavior for Cr coatings for typical reactor’s normal operation temperatures. The coated samples mechanical behavior was found to be asymptotic to that of uncoated substrates in all the studied cases. This represents a significant advantage, and it is expected to accelerate the process of licensing, given that coated fuel mechanical behavior will remain the same to that of the current zirconium alloy cladding.

4.3.1.2 Ceramic Coatings

Few data exist concerning the mechanical behavior of ceramic coatings on zirconium alloys substrates. However, ceramic coatings are much more brittle compared to metallic coatings and more likely to undergo cracking and damage. Van Nieuwenhove [53] showed that CrN coating can however be stretched by ~1.5–2% before narrow cracks appears.

Nitride ceramic coatings such as CrN have been developed with the primary purpose of improving wear behavior, with other benefits in terms of fretting are likely to be provided [65]. Jiang et al. [66] conducted a comparative analysis on crack resistance for Cr and CrN coatings deposited by multi-arc ion plating on zircaloy-4. The study revealed that metallic Cr coatings exhibit transition of brittle- to-ductile under thermo-mechanical loading, which is, in comparison to the ceramic CrN, will result in a better Cr coatings cracking resistance.

4.3.2 Corrosion Behavior

Coatings compatibility with the coolant in normal operation conditions is investigated under standardized tests for corrosion in water at 360°C with PWRs chemistry (boron concentrations ~650–1,000 ppm), or in 415°C steam with the purpose of speeding up the process of corrosion, to simulate the irradiation induced increased corrosion [45].

4.3.2.1 Metallic Coatings

Corrosion tests developed by KAERI and AREVA/ CEA/EDF [56,62,72-74] showed that Cr coated substrates exhibited significantly reduced corrosion rates, especially in the case of AREVA/CEA/EDF coatings, the weight gain remained stable for long exposure periods, suggesting a limited evolution and growth of Cr₂O₃ layer formed on the Zr based substrate surface. Once this layer formed, the Cr coated Zircaloy corrosion rate is lowered to values closer to zero, thus, the cladding hydrogen uptake also decreased, therefore, it will not exhibit hydrogen embrittlement, and eventually results in an increased operational margins and potentially allows for longer fuel irradiations (increased burn-up). However, this behavior will have to be verified under representative LWRs irradiation conditions.

Corrosion tests were also performed in BWRs water chemistry environment at 288°C for FeCrAl coatings. The result was not as obvious as for chromium coatings, since in FeCrAl coated samples, Ni₂FeO₃ deposits were observed, and an increased weight gain was registered relative to uncoated samples of Zircaloy-2. There is no reliable data exists with regard to the corrosion behavior of FeCrAl coatings in representative LWRs environments (normal operation conditions). Bulk FeCrAl has shown to exhibit very good corrosion resistance, but occasionally weight loss, therefore, if the coating thickness is too thin, it may lose its protective nature because of material loss [45]. Consequently, more data are needed to establish a clear behavioral frame for FeCrAl coatings in LWRs environment.

4.3.2.2 Ceramic Coatings

Few data exists in the literature with relation to the investigations of behavior of MAX phases during normal operation conditions, since most of organizations working on developing this type of coatings started the work on them for ATF purposes, and therefore the orientation and focus were primarily on the HT steam behavior [28]. Consequently, there is no much of data available with regard to behavior of MAX phase’s corrosion in normal LWRs environment. Among the few studies available, Joseph Ward et.al [75] studied the corrosion resistance for four MAX phases, namely Ti₃AlC₂, Ti₃SiC₂, Cr₂AlC and Ti₂AlC in autoclave using simulated primary water conditions. Among these studied MAX phases, Cr₂AlC demonstrated the best corrosion resistance with minimal change from the original condition and no phase change is observed under XRD investigations. The authors’ attributed the superior performance of Cr₂AlC to the high Cr content that allowed the formation of passive chromia and thus inhibited the further oxidation. A nano-scale void formation was observed however, in this MAX phase coating.

An autoclave test at 360°C and 18.6 MPa was also performed by Roberts [61] to study the oxidation resistance of magnetron sputtered TiAlC and CrAlC coatings deposited on ZIRLO [61], results showed that TiAlC coating possesses poor protective nature due to hydroxide phases multiple oxide scale formation. Cr₂AlC coating showed improved oxidation performance; nevertheless, the formation of volatilized AlOOH and partial spallation was also observed.

For Nitrides coatings, the behavior is found to be dependent on the aluminum (Al) content into a large extent, due to instability nature of Al₂O₃ and its dissolution in water in LWRs environment. Thus, CrAlN and TiAlN coatings both showed poor corrosion behavior. TiN and CrN have shown good corrosion resistance on the other hand [62,74]. Researchers from the PSU [62,76], have reported that TiAlN dissolved in autoclave; therefore, multi-layered TiAlN/TiN coatings had to be fabricated with the use of final surface layer of TiN. This fabricated multi-layered coating showed good corrosion behavior after 30 days at 360°C PWRs water environment, this indicates that TiN is compatible with the coolant environment, however, additional data is needed to confirm that the final TiN layer thickness is enough to maintain the long-term behavior over the whole nuclear fuel life.

In pile and out of pile autoclave tests to evaluate the behavior of CrN, TiAlN and AlCrN in PWR and BWR environments [53] were performed by IFE/Halden project in collaboration with Canadian Nuclear Laboratories (CNL). Coatings with thicknesses of 1–4 μm were tested, and in all cases, and both PWR and BWR conditions, only CrN remained intact and a stable uniform Cr₂O₃ layer formed. TiAlN and AlCrN are both completely dissolved. Consequently, it seems that the only ceramic coatings tested to date and found to be compatible with the coolant with a significant improvement in corrosion resistance behavior, relative to uncoated Zr based alloys are the CrN and TiN coatings.

4.3.3 Irradiation Behavior

4.3.3.1 Metallic Coatings

Researchers from CEA have performed a preliminary ion irradiation tests for Cr coated samples at 400°C with an induced damage higher than 10 dpa [77]. The purpose was to assess Cr/Zr interface stability under the irradiation effect. The results of these irradiation tests showed that the Cr/Zr interface is preserved and no degradation was observed, and no diffusion was observed from Cr into the Zr based substrate. These results confirm that Cr coatings are stable under irradiation and their adherence to the Zircaloy claddings will be preserved. However, further confirmation is needed under more representative conditions, such as neutron irradiation.

Several studies have also attempted to evaluate the irradiation resistance of chromium coatings using ion irradiation tests [78-80]. C14 stabilization and the disappearance of C15 poly types were observed in Cr coated Zircaloy-4 samples due to the continuous incoming Fe flux in a 20 MeV Kr⁺ irradiation test at 400°C, nevertheless, the samples retained its adhesion and microstructure stability after the ion irradiation [78]. At temperature of 500°C and 1.4 MeV heavy ion irradiations of up to 25 dpa, acceptable swelling of ~0.66% was reported [79], which is lower by an order of magnitude than the accepted swelling value for nuclear reactors materials. FeCrAl alloy has also showed good radiation resistance under heavy ion irradiation [81]. No void formation was observed, and no Cr-enriched phases were detected in ¹⁰Cr and ¹³Cr FeCrAl alloys, the irradiation induced defects contributed only to hardening effect.

4.3.3.2 Ceramic Coatings

For the nitrides coatings TiAlN, CrAlN, and CrN with thicknesses ~2–3 μm, irradiation data in LWRs representative conditions is available from the irradiation campaigns performed in Halden reactor. In the first campaign in 2012, TiAlN, ZrO₂, and CrN coated small samples made of Inconel 600 and zircaloy-4 were irradiated for 126 days in PWR environment [82]. During the second irradiation campaign (2011 to 2012), Inconel 600 small samples, coated by CrN and TiAlN have been irradiated in BWRs conditions for 287 days [82]. The first in pile experiments on coated zircaloy-4 claddings took place in 2014 within the third campaign; the tests were performed for ~150 days in PWRs conditions. The fourth irradiation campaign from 2015 to 2016 incorporated coated Zr-2.5Nb and zircaloy-2 samples irradiated in CANDU conditions for 120 days. The common observation in all of these tests was that CrN coatings came out as a superior, the thickness of this coating remained the same with no cracks or delamination occurred under normal conditions. During in-pile tests of the coated zircaloy, fuel rods were subjected to cooling deficiency, consequently an overheating occurred, and hence it was the only situation where some cracking and CrN coating delamination occurred, nevertheless, approximately 80% of this coating remained intact still.

For MAX phases, the irradiation resistance of magnetron sputtered Cr₂AlC coatings was analyzed by Imtyazuddin et al. [83] using a 320 keV Xenon ions at 300 and 623 K. The results showed that Cr₂AlC films and even at low doses are amorphized at room temperature; but, no amorphous phase has been found for up to 90 dpa at 623 K, indicating the good irradiation resistance of this coating material at high temperatures (Fig. 8). Additionally, both Cr₂AlC and Ti₂AlC coatings were reported to reduce hydrogen uptake effectively [84].

4.3.4 Neutronic Performance

Reactor operation is also expected to be affected by the proposed ATF coatings and cladding materials through their possible impacts on neutron balance of the system, neutron energy spectrum changes, and the impact on fuel manufacturing prices [49,65,68]. Neutronic penalties are expected, since most of the coating materials have higher probabilities of neutron absorption, in comparison with the reference Zr-based alloys. The coating materials thus decrease the reactivity of the whole system which has direct economic consequences on reactor operation and profit of utilities in general [37,51]. Additionally, safety parameters and system responses in different reactor conditions could be affected.

Several studies attempted to assess and evaluate the Neutronic aspects for different ATF concepts [37,65,85-87]. Fejt et al. [65] analyzed the Neutronic performance for some proposed protective coatings that show promising potentials of application in LWRs, based on research at the Czech Technical University (CTU) in Prague. A VVER- 1200 fuel assembly is used as an evaluation model in this study. The coatings considered were Cr, CrN, ZrSi₂, Cr₂AlC, and FeCrAl. The lowest Neutronic penalty was found to be induced by ZrSi₂ coatings, due to its small neutron absorption cross-section and low density. Both Cr and CrN coatings showed close behaviors, an approximately 10% higher absorption for CrN was indentified in comparison with Cr coatings for fresh fuel. Neutron flux showed three main regions for all of studied coatings materials:

Reactivity Feedbacks in terms of reactivity coefficients were also studied; the results of all feedback coefficients for both cases showed insignificant differences. Results were comparable for Cr, CrN, and FeCrAl coatings. Calculations of Boric acid concentration coefficient showed less negative feedback for boric acid concentration changes compared to reference uncoated case. The authors attributed this to the decreased volume of water together with harder neutron flux spectrum.

The authors attempted to highlight the effect of reduced water to fuel ratio in reactivity feedbacks responses, and concluded that coating composition acts as a secondary attribute as opposed to its thickness which primarily affects the water to uranium ratio. An unusual aspect is noticed in the analyses of Cr and Cr₂AlC coatings with 50 μm thickness, these coatings showed an observable decrease in the relative coating effect at this specific thickness. The authors attributed this behavior to their assumption that the decreasing amount of water, and thus decreasing thermalization effect, is partly compensated by a balance of absorption and slowing down capabilities of the specific material with the given thickness.

Depletion calculations for the same coating materials indicated that reactivity penalties will go closer to 0 for an increasing burn-up, and it can even reach positive values, i.e. (multiplication factor of the coated fuel assembly is larger than reference fuel assembly), this effect is specifically observed for ZrSi₂ coatings, at very high burn-up. The decrease in reactivity for coated fuel assembly in higher burnup is thought to be caused by harder neutron spectrum, and increased ²³⁹Pu production at the end of fuel assembly life [88]. Effective Full Power Days (EFPDs) penalty was found to be higher for CrN and less for Cr₂AlC and FeCrAl. All materials, with exception of ZrSi₂ exhibited shorter fuel cycles compared with uncoated fuel assembly, CrN was found to have the largest effect on fuel assembly operating life by showing 3.7% less fuel cycle.

Neutronic analysis for Ti₃AlC₂, Ti₂AlC, Nb2AlC, and TiAlN coatings was also performed [66] in PWRs and BWRs environments, to evaluate the potential impacts for these coatings during normal reactor operation. Criticality and depletion calculations were performed by Serpent, which is a Monte Carlo code for reactor physics calculations. It was found that when adding thin (i.e. less than 100 μm) ceramic coatings, the corresponding reactivity penalties will be small and proportional to the coating thickness, and it could be alleviated if fuel enrichment is increased by 0.5% or less. Due to larger quantities of Zr-based alloy and ceramic coatings, BWRs reactivity and cycle length losses are a factor of two greater than in PWRs. Introducing coated spacer grids is found to increase cycle length penalty by an additional 25%. The authors recommended minimizing the coating thickness (10-30 μm) to limit Neutronic penalties for materials considered in this study.

4.4 Design Basis Conditions

The study of proposed ATF coatings concepts under HT steam in LOCA conditions represented the main focus for most of the evaluations performed on these coatings. This is due to the fact that ATFs development was promoted by the Fukushima accident and the rapid zirconium alloys oxidation at elevated temperatures. Hence, the primary purpose for these coatings’ development and application on zirconium alloys (claddings) was to provide a barrier against HT steam oxidation, and therefore significantly reduce heat and hydrogen production, comparing to the uncoated zirconium alloys [4].

4.4.1 Metallic Coatings

For the concepts being developed by AREVA/CEA/ EDF and KAERI, Cr and other metallic coatings showed significant HT oxidation rates reductions [51] (Figs. (9) and (10)). The coated samples showed orders of magnitude lower weight gains and the coating layer efficiently served as a barrier for oxygen diffusion. Results of post steam oxidation for duration of 600 s at 1,200°C showed no oxygen diffusion within the coated Zr-based samples, compared to the uncoated samples where ~150 μm deep oxygen ingress penetrations was observed. This lack of diffusion of oxygen has a major impact in terms of post quench ductility, since Zr-α(O) phase formed by the solid solution of oxygen in zirconium is very fragile and the associated diffusion of oxygen within prior-β_Zr inner layer is generally known to cause a post quenching ductility decrease in fuel claddings. Consequently, residual ductility and post quench strength for Cr coated claddings seem to considerably be increased in LOCA conditions. Ma et al. [89] demonstrated that Cr coatings could be able to protect zirconium based (Zr-1Nb) alloys in HT steam oxidation environment at 1,200 and 1,300°C. The study revealed that the formed Cr₂O₃ layer thickness and residual Cr layers depend on the oxidation time.

For tests performed at UIUC on Cr-Al and FeCrAl coatings, there were no observations indicating any HT steam oxidation rate reductions for these coatings, a slight delay in the oxidation kinetic was observed however. This may be attributed to the use of insufficient coating thickness (1 μm) in UIUC performed tests, thus, allowing the diffusion of oxygen straightforwardly via the coatings. No characterization is performed post to the oxidation tests to confirm the oxygen diffusion and formation of zirconium oxide. Therefore, it could be concluded that a minimum coating thickness might be necessary to provide adequately sufficient protection during HT oxidation [45].

Dabney et al. [90] showed that thick layers of FeCrAl coatings obtained by cold spraying exhibited high oxidation resistance in HT oxidation test in air and autoclave. Nevertheless, prompt Fe diffusion from the coating into the zircaloy substrate caused the formation of FeZr₂ laves phases and a thick interlayer of (Fe,Cr)₂Zr, FeZr₃ (Figs. 11(a) and 11(b)) [90,91].

Park et al. [92] studied the HT oxidation behavior of FeCrAl coatings deposited on Zr substrates by cold spraying technique, with a deposited Mo layer between the Fe- CrAl and the Zr based substrate to hinder inter-diffusion at high temperatures. Results showed the formation of complex multilayer structure resulted from the FeCrAl/Mo/Zr system mutual diffusion. An improved oxidation resistance of Zr based substrate exposed to 1,200°C steam environment resulted from the use FeCrAl coatings. Mechanical properties and ballooning behavior of coated samples studied under LOCA conditions revealed higher burst temperatures, and less ruptures rates compared to the uncoated zircaloy samples. Furthermore, 4-point bending tests showed that coated Zr based alloy samples possessed an increased residual ductility [92].

Four-point bending and ring compression tests after HT steam oxidation were performed by KAERI, to ensure the retention of mechanical behavior of coated claddings, and the increased post-quench ductility by ring compression tests were verified by CEA (Fig. 12) [48,56].

Both KAERI and AREVA/CEA/EDF performed creep and ballooning tests for Cr coatings. Results introduced by both groups showed similar observations. Good adhesion of Cr coatings was observed even after substantial ballooning (no observed delamination), a strengthening impact at increased temperature (creep rate reductions, increase time to rupture) were reported, especially within the 600–800°C α-Zr temperature range; and reduced balloon size and burst opening size was observed, thus, fuel fragment relocation risk and the dispersal in the reactor coolant will be limited. The strengthening impact of Cr coatings observed at HT is beneficial, as it delays the rupture time and better maintain the coolable geometry for nuclear fuel sub-assembly by mitigating flow blockage [48,56].

4.4.2 Ceramic Coatings

HT steam oxidation tests for different MAX phases at various temperatures were performed at KIT. Initially Ti₂AlC coating was used, but the weight gain was found to be equivalent or slightly worse than the uncoated samples [28,93]. The same results had also been obtained by AREVA for the same MAX phase, which was the reason for AREVA to abandon MAX phases as potential ATF coatings concepts [94]. This behavior is attributed to that TiO₂ is not stable beyond 800°C, which resulted in the overall coating degradation. Attempts to improve this poor behavior were done by changing to another MAX phase type which is the Cr₂AlC coatings [93], and it has shown reductions in HT steam oxidation kinetics (Fig. 13).

Thin Cr₂AlC coatings self healing effect and the improved oxidation resistance in 1,000°C steam were demonstrated by Tang et al. [93], their results are shown in Figs. 13 and 14. However, partial spallation and volatilized AlOOH formation were also revealed. Wang et al. [95] have shown that a 10 μm thick Cr₂AlC coating exhibited an improved oxidation resistance in air at 1,100°C.

In some other studies, like that conducted by Li et al. [96], a 12 μm Ti₂AlC coating deposited on ZIRLO is tested for 5 min in 1,000–1,200°C pure steam to assess its HT oxidation resistance. Ti₂AlC coating remained intact and demonstrated good protective properties up to 1,200°C within the oxidation test time. Ti₂AlC coatings deposited by cold spray on zircaloy-4 were also tested by Maier et al. [60] in Ar/steam mixture at 1,005°C. Their results showed that Ti₂AlC coatings offered adequate oxidation protection for the zircaloy substrate and possessed increased hardness and wear resistance. Magnetron sputtered thinner Ti₂AlC/TiC coatings (5/0.5 μm) showed improved oxidation resistance in 800°C steam, by forming a triple-layer of scale (θ-Al₂O₃ + TiO₂/θ-Al₂O₃/TiO₂) [97]. The barrier TiC layer proved to mitigate inward diffusion of Al into the zirconium alloy substrate. However, the accelerated oxidation at 1,000°C has led to cracks formation and spallation of Ti₂AlC coatings.

In general, the MAX phases “A” element rapid diffusion into the Zr based substrate with an inter-diffusion layer formation has been reported in the literature, thereby affecting the coatings protectiveness nature and lead to decreased coating protectiveness. Thus, barrier layers between the substrate and coatings need to be considered to prevent diffusion when developing MAX-phases coatings.

For other types of ceramic coatings, few data exists in the literature regarding their HT steam oxidation behavior. In a collaboration between CNL and IFE/Halden, a number of tests on CrN, TiAlN, and CrAlN coatings were performed in steam at ~1,000–1,100°C [53,98]. Significant cracking and degradation were observed in both TiAlN and CrAlN coatings, indicating a reduced oxidation resistance performance due to zirconium oxide formation beneath the coating. Some cracking were observed in CrN coating, nevertheless, the overall performance shows improved protectiveness and a decreased HT steam oxidation kinetics, but weight gains were not quantified.

LOCA tests on CrN coatings showed that this ceramic is not able to maintain the circumferential extension on the ballooning area of the Zr-based cladding, even though the coating adhered to the matrix without any apparent delamination. Many surface cracks were formed during the test, which could be attributed to inherent brittleness of the ceramic that limits oxidation protection capability of the coating in HT steam environment [49].

Local failures and cracking of CrN coatings post HT steam oxidation test at 1,200°C for 30 min were showed by the results of Krejčí et al. [99]. These cracks are thought to be caused by partial decomposition of CrN coating material to Cr₂N at high temperatures (below 850°C) [100].

Meng et al. [101] also tested the high oxidation resistance in air up to 1,160°C for a 13 μm CrN coating, deposited on Zr-702 alloy. IFE/Halden and CNL performed an integral LOCA test up to 1,200°C at a ramp rate of 5°C/s for CrN-coated claddings at the Severe Accident ORNL Test Station. In this study, cladding burst behavior was not influenced by the coating; the coating reserved its adhesion but cracked significantly at the balloon/burst vicinity [1].

4.5 Beyond Design Basis Conditions

Coatings behavior during design extension conditions could be assessed by extending the exposure time at the same LOCA tests conditions or by rising the temperature of the test. The HT steam oxidation tests could already be used in the first case to understand coatings behavior at longer times.

One critical issue that arises when evaluating coated claddings behavior at beyond 1,200°C temperatures is the evolution of eutectic point between the zircaloy substrate and the coating material [71,100-103]. This process accelerates with the increasing temperatures; therefore, the evaluation of protective coatings behavior under BDBA conditions is crucial for the further development of ATF coated claddings.

The oxidation performance of Cr coated M5 alloy under HT steam (1,400−1,450°C) for 100 s is investigated by Brachet et al. [70]. They observed a mutual swift Cr-Zr diffusion midst the heating accompanied by Cr-Zr eutectics evolution at 1,305−1,325°C, enriched progressively with Zr along with the growth of an external oxide ZrO₂/α- Zr(O) layers at the isothermal period of the oxidation. After quenching by water, a Cr depleted prior-β_Zr dendritic substructure and inter-dendritic zones enriched with Cr was composed inside the alloy, and a morphology shape of a crocodile skin-like was observed in the coated alloy. Similar microstructure in E110 alloy, coated with Cr and oxidized for 2 min in HT steam at 1,400°C was observed by Krejci et al. [100].

The negative effect of this Inter-diffusion was also observed on FeCrAl coated zircaloy by Wang et al. [102]. In this case, the fast Fe and Cr diffusion led to formulation of brittle inter-metallic compounds like FeZr₂, Fe₂Zr, FeZr₃, and Zr(Cr,Fe)₂, and a decreased melting point of Zr-Fe-Cr or Zr-Fe eutectic phases [91,104,105]. Thus, a non-protective scale formation on the oxides and the creation of micro-cracks can be caused by these eutectic phases and by the CTE mismatch between the Zr based substrate the inter- metallic. A Cr-Al coating has been developed by KAERI with the ability to increase the eutectic temperature between Zr based substrate and the coating [75]. The incorporation of a barrier thin layer between the substrate and the coating, such as molybdenum or refractory metals, could be seen as another solution to this issue of eutectic formation [106].

Wang et al. [107] studied the FeCrAl coating oxidation behavior with plasma electrolytic oxidation formed ZrO₂ barrier layer on Zr-Nb alloy surface. They concluded that a 10–15 μm ZrO₂ layer will be able to eliminate Fe and Cr diffusion to the Zr alloy substrate in steam at 1,000°C. However, the oxidation behavior at higher temperatures needs to be studied. Han et al. [106] showed that a 10.6 μm Mo inter-layer between the FeCrAl coating and the Zircaloy-4 has good properties as a barrier under steam oxidation up to 1,200°C. Despite the enhanced oxidation resistance of FeCrAl/Mo coatings, a strong inter-diffusion between Mo and FeCrAl at 1,200°C led to formation of a FeCrMo interlayer, accompanied by the generation of Kirkendall voids in FeCrAl coating.

The performance of magnetron sputtering deposited CrN interlayer as a barrier against Cr-Zr inter-diffusion at HT steam oxidation was investigated by Krejci et al. [108]. Their study showed that a ZrN thin layer was composed beneath the coating as a result to nitrogen inward diffusion produced from the dissociation of CrN to Cr₂N phase. The 13 μm CrN layer succeeded to inhibit the diffusion of Cr into Zr based substrate, and showed higher resistance to the oxidation at temperatures higher than the Cr-Zr eutectic formation (1,350°C) for only a short duration (Fig. 15). It has been shown by Tang et al. [107] that a 500 nm of TiC interlayer mitigated the Al fast diffusion from Ti₂AlC into the zircaloy-4 substrate, the presence of this TiC layer has led to prolonged oxidation resistance of Ti₂AlC coatings under steam at 800°C. Nevertheless, both Ti₂AlC/TiC and Ti₂AlC coatings showed lower oxidation resistance performance at higher temperatures (> 1,000°C).

The studies show that the approach of barrier layers has promising performance in suppressing the inter-diffusion problem between the protective coatings and the Zr-based alloys in BDBA conditions. Nevertheless, many factors need to be considered when selecting this intermediate layer, such as the possible formation of eutectic phases, the high diffusion coefficient in Zr, the low melting point, the possible phase transitions, etc. Zirconium diffusion coefficients are strongly dependent on the presence of impurities, which has significantly different reported values in the literature. Coatings behavior beyond the eutectic points needs to be investigated. Few data are available in relation to eutectic behavior for the verity of coating materials proposed. If the selected coating is quite thick, coating-substrate interactions would be much greater and may therefore significantly degrade the fuel cladding due to the eutectic formation. Hence, a compromise between the minimum thickness necessary to provide significant benefits in HT steam oxidation and a maximum thickness allowing a reduction of the potential detrimental consequences of the eutectic has to be established. Thermal neutron cross-sections for interlayer materials need to be considered as well, in view of their possible effects on the neutron economy [109].

4.6 Challenges

5. Thin Films Technologies

Although the concept of thin films have been widely applied both as decorative and in functional materials, in aerospace and energy industries, the development of protective coatings to be used with the Zr-based alloys in nuclear industry is fairly recent. Generally, the most common deposition techniques are the Physical Vapor Deposition (PVD) and the Chemical Vapor Deposition (CVD). These two methods are differing in the nature of vapor used during the process. In PVD methods, the vapor is usually made up of atoms and molecules which condense on a substrate, while in CVD methods, the vapor usually undergoes a chemical reaction on the substrate resulting in the formation of the required thin films. Much of the near-term technologies that have been proposed and used in the literature are belonging to one of these two methods. The proposed methods included: magnetron sputtering, arc ion plating, Filtered Cathodic Vacuum Arc Deposition (FCVAD), Pulsed laser Deposition (PLD), spray, 3D laser coating, and electroplating [120].

Sputtering is a method of forming thin films by accelerating a gas ionized into plasma at a relatively low vacuum level, causing it to collide with the substrate, and ejecting its atoms, and the surface cleaning is very important in this process. One of the drawbacks of sputtering method is the formation of columnar crystals of the same orientation, which forms diffusion paths for the oxygen and weaken coating’s tensile strength. In addition, magnetron sputtering is characterized by its low deposition rates, when used as a coating method, for instance, a Cr thin film in a columnar structure can be deposited in a rate of 43.3 nm·min⁻¹ [121,122], indicating that ~6 hours will be required to deposit a 15 μm thin film. Hence, the application of this method for coating the ATF claddings would likely be demanded, and some efforts to improve the deposition rate are definitely needed.

Arc ion plating method is a low-temperature process which does not necessitate any post-treatment, it uses plasma that ionizes the evaporated material and applies negative bias voltage to base material. Dense films with no voids or impurities could be obtained due to the cleaning effect via the high ion energy and ion collision on the base material. Arc ion plating is characterized by the fast deposition rates, and very dense thin films with excellent physical properties could be obtained. This method has been used for the deposition of CrN [123], CrAl [124,125] and TiAl- CrN [126] coatings on Zr-based alloys.

FCVAD method makes it possible to synthesize controlled nano layers by using arc discharge to evaporate the cathode material and deposit it on the target material surface, this method is being used by some researchers in the investigation of multilayered coatings concepts. TiN and TiAlN multilayer coatings were deposited on Zr-based alloys using Cathodic Arc Physical Vapor Deposition (CAPVD) method, the HT oxidation resistance of Cr-Zr/ Cr/Cr-N multilayer coatings have also been investigated by Kuprin et al. [127], their evaluation showed the formation CrO and Cr₂O₃ oxides that reduced the penetration of oxygen through the coatings. In 3D laser method, films are created by the reaction of a source material on a substrate using laser energy. 3D laser method can be used to deposit uniform coating layers that are easily controlled to tubeshaped surfaces. The 3D laser method for ATF has been optimized for depositing Cr/CrAl powder on a tubular substrate [75,128]. However, the process temperature to form coatings should be maintained under the Zr-based alloys final annealing temperature in order to avoid any changes in microstructure comparing to the initial condition. 3D laser technology was used by Kim et al. to form FeCrAl [124] and CrAl [75] coatings.

Magnetron sputtering [71,108,129-131], arc ion plating [132-134], laser cladding [73], plasma spraying [135] and cold spraying [136-139] were reported to be used for depositing Cr coatings on the outer side of Zr-based alloys cladding surfaces. Furthermore, Direct Liquid Injection -Metalorganic Chemical Vapor Deposition (DLIMOCVD), a CVD technology, was reported to be successfully used to create Cr coatings [140] and CrxCy coating [141] on inner-side of Zr-based alloy claddings.

PVD and spray methods are the most frequent used technologies for the deposition of Cr coatings, owing to their minimal raw materials oxidation and the relatively low processing temperature. In general, PVD methods are widely used in industry; its performance has been sufficiently verified. However, to be practically used for ATF claddings, it is essential to determine whether it is the optimal technology in terms of both performance and costs [120]. To manufacture ATF claddings with protective coatings, a uniform layer must be deposited on the surfaces of tubes of several meters, conventional PVD equipments and their sizes would not probably easily cope with this length. Hence, the initial investment for equipments costs or maintenance costs would be inevitably large relative to the other simpler manufacturing methods. Therefore, it is crucial to examine the practical use feasibility of the deposition method and the performance of protective coatings manufactured by these methods [120].

6. Conclusions

Enhancing the tolerability of nuclear fuel system to withstand severe accidents will definitely lead to increased nuclear power plants safety, since the fuel and the cladding represent the first barrier against the release of radioactive materials, which in its turn will increase the reliability of using nuclear power and reduce the dependency on fossil fuel. Many organizations and research groups all around the world are currently actively involved in developing ATF systems for near and long term deployments. Coatings technologies are considered as the most promising for the near term deployment. Many materials and compounds are proposed as coatings. The Protection capabilities of these proposed materials were being studied under different conditions simulating LWRs environments, to verify their ability to protect nuclear fuel cladding and enhance the performance in accidental scenarios. The initial evaluations have revealed the strengths and weaknesses of these coatings and indentified the points for further studies. Metallic chromium is one of the most investigated coating materials; the oxide (Cr₂O₃) formed by this coating was found to be the most stable oxide phase in HT steam and water, which made Cr the most preferable option for ATF coatings applications. A technology for full length deposition of Cr coatings on Zr alloys was developed by CEA. In-pile testing with chromium based coatings is currently ongoing (e.g. USA and Russia).

Many ceramic coatings show promising preliminary results in HT oxidation environment, nevertheless, their brittle nature and differences in thermal expansion coefficients comparing to Zr-based alloys could largely affect their use as a single layer protective coatings. МАХ-phases show good protective properties up to 1,000°С; however, poor oxidation resistance is revealed at higher temperatures which entail the need for relatively thicker coatings.

A key issue revealed in the behavior of protective coatings under BDBA conditions is the strong mutual diffusion between the zircaloy substrates and the coatings materials. This phenomenon is evident in most metallic coatings and in the MAX phases as well, even at below DBA temperatures. Hence, the development and testing of different barrier layers with the ability to limit this inter diffusion represents a key future task.

More in depth studies are needed with regards to potential impacts of these coatings on different aspects related to reactor physics, there is indeed many attempts in the literature, but the general trend is generic, therefore, more in depth studies is needed including full core analysis studies. Thickness optimization studies are need as well, in addition to the protection capability, these studies should consider neutrons economy, cost of coatings materials and manufacturing technologies.

The analysis of the literature showed large and scattered studies conducted by different groups worldwide, an international collaboration accompanied with setting a classification for these proposed coatings that prioritize the most and top promising materials, and set a collaborative development system would help in save time and development efforts, evaluate the feasibility, and advance the realization of the commercial application of this proposed ATF concept.