Fuel Cladding Research Articles

Advanced corrosion-resistant materials for enhanced nuclear fuel performance: A conceptual review of innovations in fuel cladding against molten salt degradation

The increasing demand for efficient and sustainable nuclear power has led to the development of advanced corrosion-resistant materials that can significantly enhance nuclear fuel performance. This conceptual review focuses on the innovations in fuel cladding materials designed to resist degradation in molten salt environments, which are crucial for advanced nuclear reactors like molten salt reactors (MSRs). Fuel cladding serves as a critical barrier between the nuclear fuel and the reactor environment, and its ability to withstand extreme temperatures and corrosive conditions is essential for safe and efficient reactor operation. Recent advancements in materials science have introduced novel alloys and coatings that demonstrate superior corrosion resistance in molten salts, which are used as both coolants and fuel solvents in MSRs. These innovations include high-temperature nickel-based alloys, refractory metals, and ceramic coatings, which are engineered to reduce the effects of molten salt corrosion, such as oxidation, pitting, and embrittlement. By enhancing the durability of fuel cladding, these materials contribute to improved fuel performance, longer reactor lifespans, and increased safety. This review explores key developments in corrosion-resistant materials, emphasizing the mechanisms by which these materials mitigate molten salt degradation. Additionally, it highlights the challenges associated with material selection, fabrication, and long-term performance in the highly corrosive environments of advanced nuclear reactors. Ongoing research into these materials offers promising avenues for the future of nuclear energy, particularly in addressing the limitations of current fuel cladding technologies. In conclusion, the use of advanced corrosion-resistant materials represents a significant step toward enhancing the performance and safety of nuclear fuel, ensuring the viability of MSRs and other advanced reactor designs. Continued innovation in this field is essential for the future of nuclear power.

Microstructural analysis of ex-service neutron irradiated stainless steel nuclear fuel cladding by high-speed AFM

Fuel cladding in advanced gas-cooled nuclear reactors is made of an austenitic stainless steel referred to as 20Cr/25NiNb. This material can undergo microstructural changes in these extreme environments through a combination of thermal effects and radiation damage. In this work, the microstructures present in an ex-service irradiated 20Cr/25Ni-Nb sample are studied using correlated high-speed atomic force microscopy (HS-AFM) and electron microscopy techniques for the first time. Correlated topographic, crystallographic, and chemical information from the sample surface enabled identification of secondary phase precipitates (SPPs) including M23C6, sigma phase, NbC, and G phase. These SPPs can have adverse effects on the material properties. HS-AFM analysis showed surface textures unique to each SPP. Voids formed due to irradiation were also identified across the surface by HS-AFM. These voids were found to be larger in size and depth along grain boundaries. Further analysis identified a relationship between void size and frequency and grain boundary misorientation and SPP presence.

Improved understanding of the growth of blocky alpha in welded Zircaloy-4

Zirconium alloys are widely used in nuclear reactors as fuel cladding materials. Fuel cladding is used to contain the nuclear fuel and cladding tubes are typically sealed using welds. Welding of zirconium alloys can result in changes in the local microstructure, with the potential to grow so called ‘blocky α’ grains in the welded region during subsequent thermal processing and these blocky α grains have the potential to be detrimental to the integrity of the component. In this work, complimentary heating experiments with ex situ and in situ electron backscatter diffraction (EBSD) analysis are used to aid understanding of the blocky α grain growth within the weld region. These experiments reveal that blocky α grain growth is related to the prior-β (high temperature) microstructure, as grains grow adjacent to a prior-β grain boundary and the orientation of the growing α grain can be explained using neighbourhood orientations from this prior-β grain. This growth mechanism is explained via a simple mechanism which is related to the α grain orientations, grain boundary structures and local stored energy. Ultimately, our findings indicate that the likely grain growth (size and morphology) across the weld region can now be predicted from the initial as-welded microstructure.

Revisiting the thermodynamic properties of the ZrCr2 Laves phases by combined approach using experimental and simulation methods

A comprehensive study of ZrCr2 Laves phases is important for both fundamental research and technological applications. The ZrCr2 compound is a typical precipitate observed in Zr-based alloys including the newly developed Cr-coated Zr cladding known as Accident Tolerant Fuel cladding. The brittle behavior of the Laves phases and the low melting point at the interface zone between the precipitates and the Zr matrix are considered as governing the thermomechanical behaviors of these newly developed nuclear fuel claddings for normal operating conditions but more dramatically in case of Loss Of Coolant Accident (LOCA) conditions. The aim of the present study is to investigate the various ZrCr2 Laves polymorphs and their thermodynamic features particularly where conflicts exist in the literature. Based on diffraction experiments of annealed samples, it has been established that the transformation from the C14 high-temperature form to the C15 low-temperature form is of displacive nature without the C36 polymorph forming as an intermediate phase. The stable and metastable domains of C14 and C15 and their thermodynamic properties have been determined. The specific heat of both C14 and C15 types was measured over a wide range of temperatures from 2 to 1063 K by coupling relaxation calorimetry and DSC. The experimental data were fitted using a modified Einstein model. The room temperature entropies of the C14 and C15 phases were evaluated. The enthalpy of formation was measured for the first time using drop solution calorimetry in liquid Al at 1173 K in a Tian–Calvet calorimeter. The experimental value is in good agreement with Density Functional Theory (DFT) calculations. Finally, all these new results are discussed regarding the abundant literature on the ZrCr2 Laves phases.

Microstructural and micromechanical characterization of Cr diffusion barrier in ATR irradiated U-10Zr metallic fuel

Chromium (Cr) has been recognized as a promising diffusion barrier candidate to mitigate the Fuel Cladding Chemical Interaction (FCCI) failure in metallic fuels for sodium-cooled fast reactor. This paper, for the first time, conducted an in-depth post-irradiation examination of the microstructure and composition evolution, and micromechanics of the Cr diffusion barrier in U-10Zr fuel/HT9 cladding irradiated in the Advanced Test Reactor (ATR) to 8.7% burnup at an averaged Peak Inner Cladding Temperature of 540–550 °C. Transmission Electron Microscope (TEM) characterization confirmed the preferential intergranular diffusion of Zr and U in the Cr diffusion barrier, suggesting that grain boundaries served as fast path for the diffusion of Zr and U into the Cr diffusion barrier. The interaction zone is dominated by nano crystalline α-ZrCr2 Laves phase. Despite the interaction, there is no microcracks being observed in the preserved Cr diffusion barrier and HT9 cladding, serving as a good barrier to mitigate FCCI under the studied in-reactor irradiation conditions. High density cavities in uniform distribution are observed inside Cr grains, nano particles contains Cr, Mn, and O are confirmed by Electron Energy Loss Spectroscopy (EELS) analysis. However, it is unclear whether the cavities will become an issue to the barrier integrity at higher burnup. In-situ Scanning Electron Microscopy (SEM) micro-tensile testing uncovered mechanical softening in the HT9 cladding nearing the Cr diffusion barrier, possibly due to the coarsening of lath structure and carbides precipitates.

Transmission electron microscopy characterization of Fuel Cladding Chemical Interaction (FCCI) in ATR-irradiated HT9 clad U-10M (10M = 5Mo-4.3Ti-0.7Zr wt%) metallic fuel

The pseudo-binary metallic fuel alloy, U-10M (wt%, M is the optimal combination of Mo, Ti, and Zr), has the potential to increase fuel solidus temperature, reduce the onset temperature of body-centered cubic phase, and increase the fuel's chemical stability compared with the conventional U-10Zr (wt%) metallic fuel. Post Irradiation Examination (PIE) confirmed excellent fuel performance for the U-10M (10M=5Mo-4.3Ti-0.7Zr wt%) fuel irradiated in the Advanced Test Reactor (ATR) to 2.2 at% burnup at Peak Inner Cladding Temperature (PICT) of 650 °C, an upper bound temperature for metallic fuel. But previous PIE study also observed Fuel Cladding Chemical Interaction (FCCI) on the cladding side, which is known as a fuel performance limiting issue but has not been fully understood yet. As an effort to improve the understanding of FCCI phenomenon, this study performed Scanning Electron Microscopy (SEM) and in-depth Transmission Electron Microscopy (TEM) characterization on a FCCI region. The examined FCCI region is dominated by (U, Zr)(Fe, Cr)2, suggesting that U-Fe interdiffusion reaction played a key role in inducing FCCI. Additionally, the FCCI boundary into the cladding consists of four distinctive phases, (U, Zr)(Fe, Cr)2, fcc-Cr, tetragonal UCr0.1Fe9.9Si2, intermetallic σ-FeCr, and lanthanide fission products at concentration up to ∼5.5 at%. Another goal of this study is to verify the involvement of Ti and Mo in FCCI formation on the cladding side. Observable Ti is found halfway of the thickness in the examined FCCI region, while 0.3–5 at% Mo is detected across the entire thickness of the examined FCCI region. Neither Ti nor Mo reacted with HT9 cladding constituents despite their diffusion footmark. Therefore, alloying Ti and Mo into the U-Zr fuel should not complicate the interdiffusion reaction on the HT9 cladding side.

Accident Tolerant Fuel Cladding Materials for Light Water Reactors: Analysis of Neutronic Characteristics

Accident Tolerant Fuel Cladding Materials for Light Water Reactors: Analysis of Neutronic Characteristics

Revenue Requirements Method for Assessing the Cost Impact of Fuel Cladding Corrosion in a Supercritical Water-Cooled Small Modular Reactor: A Methodological Review on Life Cycle Costing Corrosion

Abstract Canadian Nuclear Laboratories (CNL) is collaborating in the Joint European Canadian Chinese Development of Small Modular Reactor Technology (ECC-SMART) project to understand the corrosion behavior of the most promising candidate materials for a future supercritical water-cooled – small modular reactor (SCW-SMR). To support this aim and the project's requirements, the present study develops a costing method for assessing the impact of corrosion in a power generation cost model. This cost model builds on a methodological study of various corrosion engineering economics topics in nuclear power generation, such as the expected fuel cladding corrosion phenomena in a supercritical water-cooled reactor (SCWR) concept and estimating the main corrosion costs categories. This understanding is incorporated in a power generation cost model that applies the revenue requirements approach to life cycle costing (LCC). The LCC includes the main corrosion cost categories and a reliability factor used in assessing power generation costs, the costing of chemical species for controlling corrosion, and the present worth of revenue requirements. The method and model, therefore, provide a framework for understanding the kind of information available and needed for taking economical preventative corrosion measures for the current generation of water-cooled reactors and advanced reactors, such as the SCWR.

Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding.

After the Fukushima nuclear disaster, the nuclear materials community has been vastly investing in accident tolerant fuel (ATF) concepts to modify/replace Zircaloy cladding material. Iron-chromium-aluminum (FeCrAl) alloys are one of the leading contenders in this race. In this study, we investigated FA-SMT (or APMT-2), PM-C26M, and Fe17Cr5.5Al over a time period of 6 months in simulated BWR environments and compared their performance with standard Zirc-2 and SS316 materials. Our results implied that water chemistry along with alloy chemistry has a profound effect on the corrosion rate of FeCrAl alloys. Apart from SS316 and Zirc-2 tube specimens, all FeCrAl alloys showed a mass loss in hydrogen water chemistry (HWC). FA-SMT displayed minimal mass loss compared to PM-C26M and Fe17Cr5.5Al because of its higher Cr content. The mass gain of FeCrAl alloys in normal water chemistry (NWC) is significantly less when compared to Zirc-2.

Distinctive Oxide Films Develop on the Surface of FeCrAl as the Environment Changes for Nuclear Fuel Cladding

The corrosion-resistant properties of IronChromium–Aluminum (FeCrAl) alloys have been known for nearly a century. Since the 1950s, they have been explored for application in the generation of nuclear power. In the last decade, the focus has been on the use of FeCrAl as cladding for uranium dioxide fuel in light water reactors (LWRs). The corrosion resistance of this alloy depends on the oxide that it can develop on the surface. In LWRs in the vicinity of 300 °C, the external surface oxide of the FeCrAl cladding could be rich in Fe under oxidizing conditions but rich in Cr under reducing conditions. If there is an accident and the cladding is exposed to superheated steam, the cladding will protect itself by developing an alpha aluminum film on the surface.

Reaction Capsule Design for Interaction of Heavy Liquid Metal Coolant, Fuel Cladding, and Simulated JOG Phase at Accident Conditions

High temperature corrosion of fuel cladding material (15-15Ti) in high burn-up situations has been an important topic for molten metal-cooled Gen-IV reactors. The present study aims to investigate the simultaneous impact of liquid lead (coolant side) and cesium molybdate (fuel side) on the cladding tube material. A capsule was designed and built for experiments between 600 °C and 1000 °C. In order to simulate a cladding breach scenario, a notch design on the cladding tube was investigated pre- and postexposure. Material thinning by corrosion and leaching at temperatures ≥ 900 °C caused breaches at the notches after 168 h exposure. The temperature dependent cladding thinning phenomenon was used for kinetic interpretation. As the first of a two-part study, this paper will focus on the exposure capsule performance, including metallographic cross-section preparation and preliminary results on the interface chemistry.

Simulation of the compression test of the Zr1Nb fuel cladding ring

Fuel cladding is a first protective barrier against the loss of fission products that must withstand extreme conditions, from normal operation to final and interim dry storage. This hostile environment results in mechanical and microstructural damage of cladding caused by different stress levels, temperature, corrosion, hydrogen pick up and other degradation processes further enhanced by radiation. For this reason, the integrity of the cladding is a critical issue. The aim of this work is to simulate a ring compression test to evaluate the stress-strain behavior and hoop fracture properties of a zirconium-based alloy with niobium, which was chosen because it is widely used as fuel cladding in light water nuclear reactors.

Initial Multidisciplinary Study of Oxidized Chromium-Coated Zirconium Alloy for Fuel Cladding of SCW-SMR Concept: Weight-Gain and Thermal Conductivity Measurements and Coating Cost Evaluation

One of the challenges of small modular reactors (SMRs) in comparison with large reactors is the greater difficulty in achieving high burnups in smaller cores. With greater neutron leakage through the periphery, a key factor is the neutron economy of the fuel cladding. However, all large supercritical water-cooled reactor (SCWR) concepts have employed neutron-absorbing stainless steels and nickel-based alloys in order to meet all the requirements in terms of corrosion and thermalhydraulics. In order to achieve higher burnups and extend the time between refueling in a SCW-SMR, the use of chromium-coated zirconium alloy as a potential fuel cladding candidate has been explored. Chromium coatings up to a few micrometers thick have shown improved oxidation resistance of zirconium-based claddings under operating conditions relevant to SCWR concepts. In this study, Zr-2.5Nb alloy (UNS R60904) from pressure tube samples was coated using a physical vapor-deposition (PVD) method. Oxidation tests were performed on coated samples at 500 °C and approximately 25 MPa in a refreshed autoclave. The effects of the oxide on heat transfer and hydraulic resistance are also discussed in this study. Last, but not least, this study evaluates the coating cost of the fuel cladding with chromium in a vacuum plasma spray process.

A Representative Volume Element Model for Investigating the Hydride Inclusion Effect on Ductility of Zry-Based Nuclear Fuel Cladding

A Representative Volume Element Model for Investigating the Hydride Inclusion Effect on Ductility of Zry-Based Nuclear Fuel Cladding

Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water

In pressurized water reactor (PWR), fretting wear is one of the main causes of fuel assembly failure. Moreover, the operation condition of cladding is complex and harsh. A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment. It can simulate the fretting wear experiment of PWR under different temperatures (maximum temperature is 350 ℃), displacement amplitude, vibration frequency, and normal force. The fretting wear behavior of Zr-4 alloy under different temperature environments was tested. In addition, the evolution of wear scar morphology, profile, and wear volume was studied using an optical microscope (OM), scanning electron microscopy (SEM), and a 3D white light interferometer. Results show that higher water temperature evidently decreased the cladding wear volume, the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.

Thermomechanical evaluation of ATF fuel rods regarding irradiation effects

Fuel clad is the main barrier against the release of radioactive materials. For decades zirconium-based alloys have been used as nuclear fuel clads. Fukushima accident proposed the idea of designing and developing accident tolerant fuels (ATF). This research tries to evaluate the effect of irradiation damage on ATF clads under operating condition of a pressurized water reactor. In this work, the SiC and FeCrAl clads are investigated as two candidates for replacement with conventional cladding, Zircaloy-4. The neutronics of a WWER1000 reactor is modeled based on the Monte Carlo method. The neutronic evaluation of the core with desired clads is performed over three years in different spatial points. Using calculated neutron spectrum, the SPECTER and SPECOMP codes are applied to evaluate the damage caused by the irradiation in the term of displacement per atom (DPA). The thermomechanical evaluation of the fuel rod considering cladding irradiation effects and burnup are performed by a developed thermomechanical code. Also, stress, strain, creep, fuel-clad contact, thickness of oxide layer and … are calculated and the results are compared considering irradiation effects.

The elemental weakening & aggregation effect on the hydrogen adsorption on FeCrAl (110) surface

FeCrAl alloys are promising candidates for Accident Tolerant Fuel cladding in Light Water Reactors. Understanding the adsorption mechanism of hydrogen is important for understanding the hydrogen embrittlement of FeCrAl alloys. In our previous work, we find that the adsorption site can significantly influence the H adsorption energy (Eads.). However, the widely adopted Special Quasi-random Structure (SQS) method for constructing FeCrAl structure hinders comprehensive investigations on all the possible H adsorption sites. In this work, a manually created surface structure is used to systematically study all the possible H adsorption sites. It is found that when Cr/Al atoms are dispersed on the surface, the HCr/Al bonding can be weakened. The weaker HCr/Al interactions are attributed to the charge deficit of Cr/Al caused by their relatively higher electronegativity than Fe. In contrast, when Cr/Al atoms aggregate on the surface, the HCr/Al bonding can be reinforced. The reinforcement comes from the increased charge density of Cr/Al due to aggregation.

Neutronic and Thermal hydraulic evaluation of accident tolerant cladding materials in a WWER1000

The concept of Accident Tolerant Fuel (ATF) claddings is developed to improve the safety of light water reactors (LWRs). The main objective of this work is neutronic and Thermal Hydraulic (TH) analysis of potential cladding materials in a WWER1000. In this analysis, the ZrC, TiC, SiC and FeCrAl are evaluated as enhanced alternatives for traditional Zircaloy-4. For neutronic analysis, Monte Carlo particle transport method is used. For thermal hydraulic analysis, the performance of the materials in unmitigated small break LOCA is evaluated by MELCOR. Assessments show that ATFs can delay the failure of reactor pressure vessel (RPV) and core melting by reducing the heat generated from oxidation reaction. The TiC and FeCrAl claddings are more reliable in the accident, but they are not desirable from the perspective of the Neutronics. The calculations demonstrate that ZrC and SiC can be excellent alternatives to the Zircaloy-4. Also, simulations show using TiC and FeCrAl as a 0.1 mm thick covering of the Zircaloy-4 improves the multiplication factor, but it is not sufficient.

Effects of Nb and Mo on 1200 °C Steam Oxidation and Mechanical Properties of FeCrAl Alloys for Fuel Cladding Materials

Effects of Nb and Mo on 1200 °C Steam Oxidation and Mechanical Properties of FeCrAl Alloys for Fuel Cladding Materials

Zirconium nanoparticle coating development for FCCI diffusion barrier in nuclear cladding

Fuel clad chemical interaction (FCCI) is a pervasive issue for metallic nuclear fuels. FCCI can damage cladding, reduce thermal efficiency, and compromise the structural integrity of the fuel rod. In this study, we evaluated the FCCI mitigation capabilities of electrophoretically-deposited Zr nanoparticle coatings on ferritic-martensitic stainless-steel cladding (HT9). Zr was deposited on the internal surface of 6 mm outer diameter cladding using electrophoretic deposition (EPD). Thickness varied among samples with similar EPD conditions. A uniformly coated sample was selected and sent to the Argonne Tandem Linac Accelerator System (ATLAS) to simulate the coating's behavior under irradiation conditions. The coating prevented FCCI in diffusion couple studies. Results of this study confirm that Zr could potentially make an effective FCCI barrier in metallic fuel rods but the deposition process needs improvement.