Cladding Materials Research Articles

Corrosion behavior of Nb-doped FeCrAl alloy in 500°C steam

The corrosion behavior of Fe-13Cr-4Al-2.2Mo, a candidate cladding material for accident tolerant fuel, doped with varying concentrations of Nb (0, 0.5, 1.0, and 1.5 wt%), was investigated under conditions of 500°C/10.3 MPa superheated steam. Upon exposure, a triple-layer oxide film comprising outer, inner, and diffusion oxide layers was formed. The results indicate that the addition of Nb significantly enhances the corrosion resistance of this alloy by reducing the thickness and porosity of the oxide film. The enhanced volume diffusion rate of Al and Cr, facilitated by the solid solution Nb, along with the inhibited inward diffusion of oxidant ions along grain boundaries owing to Laves phase precipitates, contributed to the enhanced corrosion resistance of the Nb-doped Fe-13Cr-4Al-2.2Mo alloy.

Simulation of nuclear fuel clad high-temperature ballooning under loss-of-coolant accident conditions considering anisotropic creep

Zirconium (Zr) alloys are standard clad material in commercial light water reactors. However, it is vulnerable to ballooning and burst under high-temperature conditions such as LOCA (Loss-Of-Coolant Accident), due to large internal/external pressure difference. Considering that clad ballooning is detrimental in terms of nuclear reactor safety especially for high-burnup fuel rod, high-fidelity numerical simulation of Zircaloy clad ballooning is desirable. This paper presented a finite-element based implementation for the simulation of Zircaloy clad ballooning. The developed simulation capability predicts progressive clad ballooning up to failure given empirical burst limits. The material models coupled together are phase transformation and high-temperature creep. Especially, anisotropic creep was considered for α phase Zircaloy. The implemented algorithm for stress update is implicit in time integration, which allows relatively large time step size to efficiently simulate the process that can last for hundreds of seconds. An analytic benchmark, the PUZRY ballooning and burst test, and the Hardy stress-rupture test were utilized to verify and validate the simulation capability. The anisotropic creep calculation conforms to the analytical solution. The comparison with PUZRY and Hardy test data shows that considering anisotropic creep in the α phase region is essential. 3D ballooning was also simulated by considering circumferential temperature difference for both conventional and novel annular clad designs. Interestingly, it is the first time to note that the risk of buckling failure is predicted for the double-cooled annular nuclear fuel clad during the accident condition, only after considering the anisotropic creep algorithm.

Effect of Process Parameters on the Microstructure and Wear Resistance of Fe3Al/Cr3C2 Composites

In this paper, a brake cylinder coating comprising a composite material of an Fe3Al and Cr3C2 mixed powder was prepared by adding laser cladding onto carbon structural steel. We studied the influence of process parameters on the microstructure and tribological properties of the cladding materials using scanning electron microscopy (SEM), energy-dispersive spectroscopy (EDS), and 3D white light interferometer and wear tests. The influence of different processes on the morphology of the carbide strengthening phase was found to be relatively small with a Cr3C2 content of 15 wt.%. The carbides mainly exhibited a network structure in each group of cladding layers. The area of the network strengthening phase varied under different processes. Of the cladding layers formed with different processes, the scanning speed of the 0.003 m/s cladding layer had the lowest wear rate. When the laser power was too low or the powder feed rate was too high, unmelted Cr3C2 particles could be found in the cladding layer. During the wear process, the particles peeled off, causing severe abrasive wear. When the powder feeding rate was too low, more materials in the base material entered the cladding layer. This made the composition of the cladding layer similar to that of the grinding material, resulting in severe adhesive wear.

Metal loss and corrosion attack of FeCrAl overlay welds on evaporator tube shields of a waste‐fired power plant

AbstractThree FeCrAl alloys (APMT, EF100 and EF101) from Kanthal® and the reference Ni‐Cr Alloy 625 were used as weld cladding materials on tube shields in the evaporator tube bank of a waste‐fired combined heat and power plant. For each alloy type, the overlay welded tube shields were placed in both roof and floor positions within the evaporator for 6 months. The metal‐loss rate, the microstructure and hardness of the overlay welds before and after exposure and the corrosion products were analysed. The results showed higher metal‐loss rates in the welds placed in the roof position, confirming heterogeneities in the evaporator bank environment. Alloys were ranked from higher to lower erosion–corrosion resistance as follows: APMT ≈ Alloy 625 > EF101 > EF100. The analysis of the corrosion attacks showed a significant variation among the alloys, from a primarily homogeneous corrosion attack on APMT to intergranular corrosion in EF100 and pit formation in EF101.

Enhanced side illumination property of end emitting polymer optical fiber (EEPOF)/polyester (PET) woven fabrics by acetone/methanol mixture etching process

The rapid development of fiber optic technology has enriched smart wearable technology. Standard (end emitting) polymer optical fibers (EEPOF) are linear and flexible, which is feasible to integrated them into textiles and applied to curved parts of the human body for achieving individualized functionality (e.g., safety light-emitting garments, wearable photo-therapeutic textiles, fashionable displays, smart sensing apparel for monitoring etc.). However, the luminance of the EEPOF-incorporated fabric should be improved. In this work, the EEPOF/polyester (PET) woven fabrics (POF-PET-Fs) were firstly prepared and then etched by using acetone/methanol mixtures with different acetone concentrations. The morphology, breathability, and side illumination behavior of the etched POF-PET-Fs were investigated. As a result, higher acetone concentration in acetone/methanol mixture causes more destroy of EEPOF cladding layer and even a slight damage for core material. With higher acetone concentration is used, the dissolving cladding material blocks the pores of the warp and weft interweave of the woven fabric, thus affecting its air permeability. Besides, the bell-shaped side illumination behavior of etched POF-PET-Fs is found. The maximum luminance value of POF-PET-Fs reaches 74.8 cd/m2 when 80% acetone concentration in the acetone/methanol mixture is used. This work provides a feasible method for the preparation of POF fabrics for luminous purpose, which is beneficial to the textile industry and POF-related fields.

Neutronic Analysis of the AP1000 Fuel Assembly with Accident Tolerant Cladding Materials

Neutronic Analysis of the AP1000 Fuel Assembly with Accident Tolerant Cladding Materials

Source term analysis of FeCrAl accident tolerant fuel using MELCOR

It has been established that incremental improvements in beyond design basis accident performance can be achieved through accident tolerant fuel (ATF). However, they have the potential to recover margin with respect to conventional fuel and therefore enhance plant economics through uprate or cycle length increase. To realize this potential, it is necessary to quantify the reduction in source term due to use of ATF, and correspondingly how this is affected by increasing the cycle length and/or burnup. This requires development of a risk informed analysis methodology for ATF under high burnup conditions, which is being developed within LWRS. To this end, a MELCOR simulation of FeCrAl ATF in a recovered Large Break LOCA (LBLOCA) scenario has been developed, using a model based on the Zion Pressurized Water Reactor (PWR). The new user defined material capability, along with the inclusion of detailed neutronics- and depletion-derived parameters such as core power distribution, decay heat behavior, and fission product inventories, allows a more detailed simulation of FeCrAl clad material properties and behavior than has previously been possible using MELCOR. These detailed FeCrAl results were compared with a zircaloy clad model to investigate the differences between the two clad types and quantify the benefits of FeCrAl ATF with an 18-month cycle. Next, the fuel cycle with ATF was extended to 24 months, to determine whether any additional safety margin provided by FeCrAl ATF could be leveraged to implement high burnup FeCrAl-clad fuel while retaining the same operating and safety limits as current zircaloy-clad fuel. For the particular scenario analyzed, the delay to fuel failure from using ATF was of the same order as the LOCA recovery time, and hence significant in reducing fission product release, with the 24-month FeCrAl core performing better than the 18-month zircaloy core. It is noted that for other transients, the reduction in release due to using FeCrAl may be less significant. The material model developed here can be used in such further studies in support of determining the overall source term.

Materials Science Analysis of RCCA Structural Materials and Calculations for Lifetime Extension

The paper considers the possibility of lifetime extending of the rod control cluster assemblies in operation by determining their residual lifetime using an approach that takes into account the actual position of the rod control cluster assembly in the core. Excessive conservatism was introduced by assuming that the immersion of the 10th working group in the core during all fuel campaigns is 70 %, while the actual position is 85-90 %. Taking into account the actual position of the rod control cluster assemblies indicates much less irradiation of the lower, most loaded part of control rod cladding and much less burnup of the 10B compared to 70 % immersion of the rod control cluster assemblies, since they are in the area of lower neutron flux density.
 The results of a comprehensive analysis of available studies and test results to determine the phenomena that can potentially approach the achievement of the boundary state of the structural components or limit the lifetime of control rods and rod control cluster assemblies in general are presented. Such phenomena include general corrosion of structural materials (42CrNiMo, Cr18Ni10Ti) (decrease in wall thickness, increased transfer of radioactive corrosion products the coolant); radiation resistance (degradation of mechanical characteristics) of control rod cladding materials (42CrNiMo) and other structural components of the rod control cluster assemblies (42CrNiMo, CrKh18Ni10Ti); radiation resistance of neutron-absorbing materials (B4C, Dy2O3∙TiO2); compatibility of absorbing materials with the control rod cladding.
 Based on the results of post-irradiation examinations of control rods tested in research reactors as dummy rods or rod control cluster assembly rods operated in the VVER-1000 core, the following boundary values were established that can be used as criteria for maintaining the integrity of the control rod cladding: maximum fast neutron fluence in the lower, most loaded part of the control rod cladding (in the weld area of the lower end plug and the cladding) should not exceed 34×1021 n/cm2; maximum burnup of the 10B isotope should not exceed 45 %. If these two criteria are met, the possibility of other potentially damaging phenomena is too low.
 According to the calculations that use the schedule of changes in the position of the rod control cluster assemblies of the control group and the power of one of the power units during one fuel cycle, one can preliminary conclude that the proposed approach will extend the lifetime of almost all rod control cluster assemblies from 25500 to 38000 hours in the automatic control group and from 75600 to 113500 hours in the shutdown group. The residual lifetime should be assessed for a particular power unit and each rod control cluster assembly separately, since each rod control cluster assembly has its own operating history.

Hybrid Lithography Fabrication of Single Mode Optics for Signal Redistribution and Coupling

This paper describes advances in hybrid-lithography process, combining UV-lithography for planar, single mode redistribution layer (RDL) and 2-photon- polymerization direct-laserwriting (2PP-DLW) for micro-mirrors inside RDL-opening. Improvements to multi-layer direct patterning of OrmoCore/-Clad material system using UV- ithography and need for broadband UV-LED source are presented. Near square core cross sections and smooth sidewalls are achieved. Openings in full stack with steep sidewalls without residual layer are patterned. To optimize 2PP-DLW-process processing window for both OrmoComp and IP-DIP is thoroughly characterized. Roughness measurements prove feasibility even of coarsely printed structure as reflective μ-mirror for 1550 nm wavelength. Finally these results are applied to periscope probe for wafer-level-testing of edge emitting lasers and proof of concept is shown. Outlook to further research on UV-lithography of multi-layer waveguide stack and alignment with μ-mirror printing is given.

Mechanism of dynamic recrystallization of a FeCrAl alloy during hot compression

FeCrAl alloys have the opportunity to replace current Zr-based cladding materials in light water reactors, and dynamic recrystallization (DRX) plays an important role on microstructural controlling of FeCrAl alloys during hot or warm deformation, such as forging, hot rolling, extrusion. In this research, dynamic recrystallization mechanism of FeCrAl alloy has been systematically investigated by microstructure characterization after isothermal compression. Results show that deformation strain rate and temperature could significantly influence microstructure and DRX mechanism of FeCrAl alloy. When deformed at 0.1s−1, prior grains become long strip with serrated boundaries and are refined by geometric dynamic recrystallization (GDRX). Prior grains are not separated by high angle grain boundaries (HAGB) even deformed at 1200 °C, because strain is too small to impingement of HAGB. Whereas, when deformed at 10s−1 strain rate, there are lots of sub-boundaries with about 2° misorientation within grains, and several strain-free grains are found located at boundaries when deformed at 900 °C or 1000 °C, while the prior grains are divided by HAGB and equiaxed grains are formed at 1100 °C. It was found that mechanism of DRX changes from discontinuous dynamic recrystallization (DDRX) to continuous dynamic recrystallization (CDRX) with increasing temperature from 900 °C to 1100 °C. The phenomenon results from the dislocation movement behaviors changes with increasing temperature. This research will provide meaningful guidance for industrial application of FeCrAl alloys.

Effective Multiplication Factor Analysis of Gas-cooled Reactor using OpenMC code with ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 Nuclear Data

As the demand for nuclear energy keeps rising, Generation IV reactor designs continuously improve. One such potential reactor design is the Gas-cooled Reactor. Neutronic analysis is the foundation for ongoing design modifications and alternatives for the Helium Gas-cooled Reactor. An important aspect of neutronic analysis is the nuclear data library. This research compares how different nuclear data libraries impact a Helium Gas-cooled Reactor design by comparing the neutron effective multiplication factor (k-eff) value. This research was conducted by the Monte Carlo method using OpenMC code. ENDF/B-VII.1, ENDF/B-VIII.0 and JENDL-5.0 were used as nuclear data libraries to simulate the reactor design for ten years of depletion time with 1-year timesteps. The reactor is designed in two different spectrums: the fast neutron spectrum design and the thermal neutron spectrum design. The differences between the two designs are in the cladding and reflector materials where the thermal neutron spectrum uses graphite moderator. The results for thermal spectrum design show that while each nuclear data library’s average k-eff values vary, they all produce the same results at timestep zero, with k-eff values around 1.02500. The average k-eff value from simulation using the ENDF/B-VII.1 library is 1.01477, the average k-eff value from simulation using the ENDF/B-VIII.0 library is 1.01122, and the average k-eff value from simulation using the JENDL-5.0 library is 1.01047. Each simulation result has an average uncertainty of 0.00040. The difference is due to the different amounts of data in each library. JENDL-5.0 has the highest amount of data inside its nuclear libraries, followed by ENDF/B-VIII.0 and ENDF/B-VII.1. Meanwhile, the average difference between the different libraries is insignificant for the thermal gas-cooled reactor simulation results. ENDF/B-VIII.0 produces an effective k-average of 1.04437, and JENDL-5.0 produces an effective k-average of 1.04273. Mass change over time results also shows that the transmutation of Uranium-238 into Plutonium-239 occurred. The fast spectrum reactor design shows a greater increase in plutonium-239 than the thermal spectrum design. The three libraries used did not show large differences in results in calculating changes in the mass of Uranium-238 over time due to Uranium-238‘s large mass, but they still showed differences in Plutonium-239‘s mass change over time. The average value difference between libraries in the fast neutron spectrum reactor design is 50 kg, and in the thermal spectrum it is 30 kg.

A Framework for Studying the Possibility of Using Basalt Cement as a Cladding Material for Indoor Spaces

A Framework for Studying the Possibility of Using Basalt Cement as a Cladding Material for Indoor Spaces

Yttrium effect on 475 °C brittleness of Fe–13Cr–6Al–2Mo-0.5Nb-0.15Zr alloy

The effect of yttrium on the 475 °C brittleness and microstructure stability of Fe–13Cr–6Al–2Mo-0.5Nb-0.15Zr alloy after long-term aging at 475 °C has been investigated. At the early stage of aging, the nanoscale metastable transition phase precipitates in large quantities. After aging for 10,000 h, the nanophase in the Y-free alloy has completely disappeared, and only a great deal of granular Laves phases are left, while a large number of nanoscale phases still exist in the grain of Y-containing alloys. This is because the addition of yttrium combines with vacancies, hinders the diffusion of Laves phase-forming elements, and has the result that the transformation of nanoscale phases Fe2(Mo0.3Nb0.7)Al to Laves phase is inhibited during the aging process of the Y-containing alloys. During aging, precipitation-free zones (PFZ) are formed in the surrounding area where the Laves phase and grain boundaries exist. The elongation of Y-containing alloys decreases with yttrium content, but it is still higher than that of Y-free alloy. The addition of yttrium improves the aging plasticity and weakens the degree of 475 °C brittleness of the alloy, and the effectiveness is diminished with the yttrium. This plays a key role in the further study of Fe–Cr–Al alloys for fuel cladding materials.

Corrosion behavior of magnetron sputtering Fe31Cr20Al17Ti16Nb16 high entropy coating in liquid lead-bismuth eutectic with different oxygen concentration at 500 °C

The major concern of the structure integrity of lead cooled reactor is the corrosion behavior of fuel cladding and other core internal materials. Magnetron sputtering technology is adopted as an surface coating method to improve the corrosion resistance of nuclear fuel cladding exposed to liquid lead-bismuth eutectic (LBE). Fe31Cr20Al17Ti16Nb16 high entropy alloy (HEA) coatings were prepared on tubular ferritic/martensitic (F/M) steel which is the candidate fuel cladding material, and the corrosion behaviors of Fe31Cr20Al17Ti16Nb16 coatings in LBE are investigated in this work. F/M steel specimens with Fe31Cr20Al17Ti16Nb16 coatings were exposed to 500 °C stagnant liquid LBE with oxygen concentration of Co ≈ 1 × 10−4 wt% and Co ≈ 1 × 10−7 wt% for 1000 h respectively. The primary findings suggest that the oxygen content of LBE mainly affects the morphology of duplex oxide layers generated during corrosion process, which resulting from the different inward diffusion flux of oxygen. In oxygen-saturated (Co ≈ 1 × 10−4 wt%) LBE, the overall oxide layer exhibits bulging and incompact morphology and infiltration of Pb is probed at inner Cr2O3 oxide layer, larger-sized cavities (the observed maximum dimension reaches 233.5 nm) distribute close to the outer Fe3O4/FeCr2O4 layer; In oxygen-depleted (Co ≈ 1 × 10−7 wt%) LBE, duplex oxide layer presents relatively dense and uniform state, tiny cavities (observed maximum dimension is 38.3 nm) dispersing at the internal of oxide areas. Corrosion rates of the Fe31Cr20Al17Ti16Nb16 coatings under two oxygen situations are 2 orders of magnitude lower than uncoated F/M steel under same exposure condition, the oxide layers with better uniformity and density produced in Co ≈ 1 × 10−7 wt% LBE indicating the better application advantage of Fe31Cr20Al17Ti16Nb16 coatings in oxygen-depleted LBE.

Numerical simulation prediction of critical heat flux for 5 × 5 rod bundle with multiple grid spacers and different cladding materials

The critical heat flux (CHF) is one of the safety standards for core thermal design and is crucial for the normal operation of the reactor. The CHF problem of rod bundles with grid spacers is a hot topic in related research. The successful design of grid spacers can improve channel CHF by affecting the flow of downstream coolant, and this detection analysis can be achieved through numerical simulation methods. In addition, most simulation studies have not considered the solid region of the cladding, so there is little research on the effect of cladding materials on CHF. The role of grid spacers in rod bundle CHF is the focus of this paper, and the influence of cladding materials is another discussion direction. In this paper, following meticulous mesh generation and computational model validation, a study on the CHF of the 5 × 5 rod bundle with multiple grid spacers and different cladding materials is carried out by using CFD software. The verification results indicated that the physical model established in this study can reasonably predict the generation location and critical power value in the 5 × 5 rod bundle with grid spacers. Furthermore, the study identified that the grid spacer exerts a significant influence on coolant flow within the channel, while the weakening of the mixing effect of the grid spacer on the coolant flow is the main factor to lead the boiling crisis upstream of the grid spacer. By comparing the effects of thermal properties of different materials on CHF, it was found that when the thermal conductivity ratio of the two materials is 1.08, CHF is almost identical under the condition of a reference heat flux of 0.1 MW/m2 as a step size. In addition to thermal properties, further attention may be paid to the influence of surface characteristics of materials on CHF in future work.

Optimization of weldability and joint strength of Al-Mg-Si with additional Al-Si cladding based on a design of experiments investigation

Some aluminum alloys, such as those alloyed with Si and Mg, are difficult to weld due to their susceptibility to hot cracking. Previous research has investigated this challenge in resistance spot welding (RSW). To improve the weldability of the AW-6111 aluminum alloy, the roll cladding process was employed, combining it with AW-4045 as the cladding material. This resulted in a wider weld lobe, improved electrode wear, and enhanced joint quality. However, it is evident that the cladded material could benefit from improved welding conditions. Despite a larger nugget diameter, similar mechanical properties were observed for both the minimum and maximum boundary conditions of the weld lobe, determined by Inom and Imax.This study employed a design of experiments approach to optimize the joint strength and improve the weld quality of the AW-6111 + AW-4045 cladded sheets. The tensile-shear test results demonstrated an improvement in lap-shear strength of 38 % and 44 % compared with the previous study. Furthermore, the test samples predominantly exhibited a ductile partial thickness failure mode, and a significant improvement in the weld nugget quality of the cladded AW-6111 + AW-4045 was observed. The cladded joints also exhibited a Si alloying content of approximately 5 % in the heat-affected zone and fusion zone, thereby reducing the risk of crack formation and propagation, thus improving significantly the weld quality. The results of this study contribute with the continuous validation of Al alloys for the transport industry and with the use of cladding technology to enhance the aluminum sheet properties and respective weld quality.

Direct Energy Deposited Cladded Material Characterization: Exploring Barkhausen Noise and Thermal Field with Parametric Simulation

Direct Energy Deposited Cladded Material Characterization: Exploring Barkhausen Noise and Thermal Field with Parametric Simulation

Post-quench ductility limits of coated ATF with various zirconium-based alloys and coating designs

Post Quench Ductility (PQD) of various Cr-coated cladding designs with respect to base cladding materials, Cr coating methods, and its thickness were investigated to explore Emergency Core Cooling System (ECCS) limits. Both inner wall and outer wall of coated (Cr and Cr/CrN) claddings were steam oxidized at ∼1204 °C and then water quenched. The post-oxidized specimens were ring compressed to attain offset strains based on which ductility was assessed in compliance with the U.S. Nuclear Regulatory Commission (U.S. NRC)’s protocols. Weight gain-based Equivalent Cladding Reacted (ECR) limits of coated specimens (Cr and Cr/CrN) are consistently lower than those of the base cladding materials due to the early load drop under Ring Compression Test (RCT). Yet, time needed to reach the ECR limit is still increased for coated specimens because it effectively undergoes single side oxidation with protective coating. Cracks imitated from ZrCr2 are primarily responsible for the early major load drop observed for tested coating thicknesses (8.9 µm and 18.9 µm), and hence reduced ECR limits compared to bare Zircaloy. The cracks from ZrCr2 was promoted by an increased oxygen concentration of the interfacing Zr matrix owing to the diffusion of oxygen through the coating. The thicker coating reduces this oxygen level, thereby delaying the load drop from ZrCr2 during RCT and increases the ECR limit. The maximum attainable ECR limit of coated ECR limits is affected by the base Zircaloy as it gives as the ECR ceiling from which ECR reduction for coated cladding occurs. For the tested base cladding materials (Opt. ZIRLO™ and HANA-6), ECR 19 %, which is the limit for Cr-coated (8.9 µm) HANA-6, may serve as the lower envelope limit. Hence, ECR 19 % (single-side ECR) can conservatively serve as the conservative, yet non-design specific, Cr-coated ECR limit for the most of the modern base Zircaloy materials (Opt. ZIRLO™ and HANA-6).

Characteristics of ejecta resulting from hypervelocity impact on Al/Mg explosive welding clad materials

In this study, we propose the use of a clad material, created from aluminum alloy A1100P-H24 and magnesium alloy AZ31B-H24 through explosive welding, as a Whipple shield for space applications. Two kinds clad materials with thickness ratios of Mg: Al = 0.5: 0.5 and Mg: Al = 0.7: 0.3 made from explosive welding, are compared to the current standard, the 1 mm thick A6061-T6 Whipple shield utilized in the international space station. By explosive welding low density, high specific strength, and high impact strength magnesium alloy with aluminum alloy, clad materials achieved 16.8% and 23.7% reduction in area density compared to A6061-T6. In the context of space debris mitigation, we performed hypervelocity impact tests on these clad materials. Detailed comparisons were made between the impact characteristics of the explosive welded clad materials and A6061-T6 by examining the perforation hole size and shape, and by collecting and analyzing the ejecta produced during impact. The experimental results revealed that the clad material with the thickness ratio of Mg: Al = 0.7: 0.3 experienced delamination at the material interface due to the penetration of projectile. On the other hand, the Mg: Al = 0.5: 0.5 clad material exhibited a perforation hole pattern comparable to that of monolithic aluminum and magnesium alloy materials. The cumulative number distribution of ejecta indicated that the clad material with a Mg: Al = 0.5: 0.5 ratio could reduce ejecta production. To further evaluate interface delamination, post-impact ultrasonic analyses were performed on the clad materials.

Fuel-clad interactions: Uranium nitride and uranium dioxide with FeCrAl and Ti-modified stainless steel

This study elucidates on the compatibility of fuel and clad materials for potential use in lead cooled fast reactors. Diffusion couples were performed on representative, unirradiated fuel forms (UN and UO2) with potential cladding materials, a Ti-modified stainless steel (D9) and a FeCrAl alloy (Kanthal APMT). Temperatures of 800–1200°C, with durations up to 100 h, were investigated. Chemical interaction between all combinations of the fuel and cladding candidates was found to be minimal, with no diffusion gradients observed or extensive interface phases formed. When interaction occurred, it was in the production of a minor discrete phase. At the highest temperatures, D9 with UN resulted in the formation of a Ti-C-N phase. D9 with UO2 resulted in the formation of a Ti-C-O phase. FeCrAl with UN resulted in the formation of an Al-N phase. FeCrAl with UO2 results in the formation of an Al-O phase. Despite the observed mechanical bonding between fuel and cladding material, no evidence of any chemical interaction with the uranium in the fuel material was found. At the higher temperatures investigated (>1000°C) there was evidence of clad softening resulting in transfer of material to the fuel surface.