Accident Tolerant Fuel Cladding Materials Research Articles

Early-stage diffusion and oxidation behavior of Cr-Nb coated Zr alloy accident tolerant fuel cladding materials at 1200°C–1500°C

Although the Cr-coated Zr alloys are considered as the most promising accident tolerant fuel cladding materials in recent years, the Cr-Zr interdiffusion problem is a major obstacle for the widely used of the Cr coating. In this study, Cr-Nb coated Zr alloy samples were prepared by magnetron sputtering. The Nb layer was designed as a barrier layer to retard the Cr-Zr interdiffusion. By using a fast-in and fast-out equipment, short time annealing was conducted in argon environment to study the coating evolution in high temperature environment. Results show that the Nb coating can retard the inward diffusion of Cr because of the low diffusion coefficient of Cr in the Nb layer. However, the complete miscibility between Nb and Zr lead to the fast consumption of the Nb coating. After 10 min annealing at 1200 °C, the 2.5 μm Nb coating disappeared, and a Cr-Zr-Nb mixed phase formed between the residual Cr coating and Zr substrate. The early-stage microstructure evolution of the sample was analyzed, and the barrier effect of the Nb coating was summarized.

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

Diffusion and oxidation behavior of Cr-ZrN coated Zr alloy accident tolerant fuel cladding materials at high temperature

In this study, a Cr-ZrN coating was constructed on Zr alloys as a protective layer. The ZrN layer was employed as an interlayer to restrain the CrZr interdiffusion. The samples were annealed in inert gas and steam environment up to 1400 °C. During high-temperature annealing, the ZrN coating that produced via multi-arc ion plating decomposed into zirconium and nitrogen. Discontinuous pores formed beneath the residual ZrN layer, which were formed by accumulation of nitrogen. With the increase of exposure time, the sizes of these pores decreased and finally disappeared. Compared to Cr-coated Zr alloys, the ZrN interlayer effectively limited the CrZr interdiffusion and eutectic reaction over a limited duration. Both the ZrN layer and the subsequently formed α-Zr(N) layer served as barrier layers, impeding the diffusion of Cr and Zr atoms.

Hydrothermal corrosion of PVD and cold spray Cr-coatings on Zircaloy-4 in hydrogenated and oxygenated LWR coolant environments

Accident tolerant fuel (ATF) are currently being developed as part of an international effort to modify or replace zirconium-based cladding materials in light water reactors. The main objective of ATF cladding materials is to improve fuel reliability and safety during accident scenarios. Thus, these materials should perform comparable to or better than the current zirconium alloy. One of the ATF concepts explores modifying standard Zircaloy cladding by coating their outer surface. In this study, surface-modified Zr-4 is produced by depositing a protective coating of chromium through two different coating techniques, Physical Vapor Deposition (PVD) and Cold Spray (CS). The coated specimens were tested at ∼300 °C in out-of-pile autoclaves simulating Boiling Water Reactor (BWR) conditions. Results show that after one month of immersion, Cr-coated Zr-4 evidenced a small mass gain in hydrogenated environments. However, both types of coated specimens showed significant coating degradation and mass loss in oxygenated environments.

Research Progress of ODS FeCrAl Alloys-A Review of Composition Design.

After the Fukushima nuclear accident, the development of new accident-tolerant fuel cladding materials has become a research hotspot around the world. Due to its outstanding corrosion resistance, radiation resistance, and creep properties at elevated temperatures, the oxide dispersion strengthened (ODS) FeCrAl alloy, as one of the most promising candidate materials for accident-tolerant fuel cladding, has been extensively studied during the past decade. Recent research on chemical composition design as well as its effects on the microstructure and mechanical properties has been reviewed in this paper. In particular, the reasonable/optimized content of Cr is explained from the aspects of oxidation resistance, radiation resistance, and thermal stability. The essential role of the Al element in oxidation resistance, high-temperature stability, and workability was reviewed in detail. The roles of oxide-forming elements, i.e., Y (Y2O3), Ti, and Zr, and the solid solution strengthening element, i.e., W, were discussed. Additionally, their reasonable contents were summarized. Typical types of oxide, i.e., Y-Ti-O, Y-Al-O, and Y-Zr-O, and their formation mechanisms were also discussed in this paper. All aspects mentioned above provide an important reference for understanding the effects of composition design parameters on the properties of nuclear-level ODS FeCrAl alloy.

Experimental and theoretical investigation on the effect of surface porosity on the nucleate flow boiling heat transfer

The surface porosity of accident tolerant fuel cladding materials is different from the fuel material inservice at nuclear power plant, which leads to differences in the heat transfer characteristics. An experiment was conducted to investigate the effect of surface porosity on the nucleate flow boiling heat transfer. The results showed that bubbles on the SiC surface appear to be more uniform and remain isolated longer after bubble generation, and the average diameter of bubbles on the SiC surface tends to be smaller than that on the FeCrAl alloy surface. The nucleation site density is power-law dependent on porosity and superheat, and negative power-law dependent on critical cavity diameter and coolant subcooling. For the same degree of temperature increase on both surfaces, the number of nucleation sites on the surface with larger porosity increases much more than that on the surface with lower porosity, while the coolant subcooling degree inhibits the bubble nucleation and shrinks the bubble detachment diameter accordingly. A new correlation was proposed based on dimensional analysis, and the predictions were compared with the present experimental data and the other research data, with an 86% confidence interval of ± 50% for the consolidated datasets. Furthermore, the side-by-side comparison of the performance on these correlations showed that the present correlation, which includes surface porosity and roughness factors, significantly reduces the prediction error compared to the previous correlations.

Incipient corrosion of FeCrAl alloys in H3BO3- and LiOH-containing pure water at 360 °C and 18.5 MPa

The demand of accident tolerant fuel cladding materials provides a momentum for the development of FeCrAl alloys. To prove its potential, this work studied the corrosion behavior of FeCrAl alloys in H3BO3- and LiOH- containing water at 360 °C and 18.5 MPa for 500 h. XPS, XRD, SEM, Raman and AFM analyses were employed to identify oxide films, and the early corrosion mechanism of FeCrAl was discussed. Duplex oxide films were found on the surface of FeCrAl. The inner layer was Fe-Cr-Al spinel, and the outer crystals was dominantly composed of Fe2O3, accompanied by Fe3O4, FeCr2O4, NiFe2O4 and Li2Fe3O4. Similar to FeCrAl, Zircaloy-4 showed negligible weight change. However, its surface corrosion was more severe, and the oxides formed were mainly ZrO2 and Zr3O1-X. Li6Zr2O7 and other sub oxides were also observed.

Three-Dimensional Modeling of Thermal-Mechanical Behavior of Accident Tolerant Fuels

Considering the safety issues of the traditional UO2-Zr fuel, a variety of accident-tolerant fuel (ATF) candidates have been proposed in recent years. Among the several ATFs, U3Si2, and UN are the two promising candidates for fuel materials owing to their high thermal conductivity and high uranium density. The FeCrAl alloy and the SiC/SiC composite material are the two promising candidates for cladding owing to their high oxidation resistance and high strength. In order to quantitatively evaluate the performance of ATFs, this study summarizes the physical models of typical ATF cladding materials (FeCrAl and SiC) and pellet materials (UN and U3Si2). Then a three-dimensional non-linear finite element method is applied to simulate the thermal-mechanical behavior of several typical fuel-cladding combinations, including UO2-FeCrAl, UN-FeCrAl, U3Si2-FeCrAl, U3Si2-Zr, and U3Si2-SiC. The important physical quantities, such as the fuel centerline temperature, the deformation of the pellet and the cladding as well as the pellet-cladding mechanical interaction (PCMI) were studied. The fission gas release model was also verified and improved.

High-strength joint of nuclear-grade FeCrAl alloys achieved by friction stir welding and its strengthening mechanism

Nuclear-grade iron-chromium-aluminum (FeCrAl) alloys are promising candidates for accident tolerant fuel (ATF) cladding materials employed in advanced light-water reactors. Apart from the fusion welding processes, it is necessary to explore available solid-state joining processes such as Friction Stir Welding (FSW) for the FeCrAl alloys. In this study, the feasibility of friction stir welding in the achievement of high-strength Fe-13Cr-5Al alloy joint was investigated. The results demonstrated that a nuclear-grade FeCrAl alloy joint with nominal ultimate tensile strength of ∼629 MPa and fraction elongation of ∼19 % was successfully fabricated using friction stir welding (FSW). For the first time, the joint efficiency of FeCrAl alloys was approximately 100 % or better according to microhardness and tensile tests. Texture in the stir zone (SZ) consisted of a mixture of J ¯ 1 ¯ 1 ¯ 0 [ 1 ¯ 1 2 ¯ ] , D 1 1 ¯ 1 ¯ 2 [ 111 ] , D 2 1 ¯ 1 ¯ 2 [ 111 ] , and E ¯ 1 ¯ 1 ¯ 0 [ 1 ¯ 11 ] components due to severe simple-shear strain. Microstructures in the SZ were predominately refined through discontinuous dynamic recrystallization (DDRX) due to the high-density of high-angle grain boundaries (HABs) and the abundancy of triple junctions induced by the increased strain. The excellent properties of the FeCrAl alloy joint were attributed to the formation of high-density low-angle boundaries (LABs) and refined grain structures. The fundamental discoveries in this study reveal a promising application prospect of FSW in the joining of nuclear reactor ATF cladding materials.

Pool boiling critical heat flux studies of accident tolerant fuel cladding materials

Pool boiling experiments have been performed at atmospheric pressure on potential accident tolerant fuel (ATF) cladding materials (Zirlo®, Cr coated Zirlo®, FeCrAl coated Zirlo®, and monolithic SiC) to evaluate the Critical Heat Flux (CHF) points. Corrosion resistant coatings of Cr and FeCrAl alloy were deposited on flat Zirlo® samples using cold spray technology. The as-prepared samples after uniform surface polishing were subjected to autoclave tests at 360 °C water and 18.6 MPa for 360 h to simulate prototypic corrosion of the ATF cladding materials in LWR normal operation. Surface characteristics potentially influencing CHF such as surface morphology, roughness, static contact angle, and surface chemistry were characterized using a suite of characterization methods including scanning electron microscopy, 3D optical profilometry, optical contact angle measurements, and x-ray photoelectron spectroscopy (XPS). Thermo-physical properties of the samples such as density, thermal conductivity, and heat capacity were measured by differential scanning calorimetry and laser flash thermal diffusivity measurement. FeCrAl coatings showed CHF values comparable to bare Zirlo® samples, while slightly lower CHF values were observed for Cr coated samples and bulk SiC flats (produced by chemical vapor deposition, CVD). As-deposited Cr coatings showed 67% higher CHF than the as-polished Cr coatings due to its higher surface roughness level. The autoclave testing for the prepared samples generally increased the CHF except for the surface polished Cr coatings. Similar CHF values of bare Zirlo®, FeCrAl coating, and CVD SiC samples were observed after autoclave test but there were negligible change CHF of Cr coatings. XPS studies indicated that the formation or deposition of a few monolayers of hydrophobic carbonaceous species on surface can affect CHF values. The trends in CHF data are discussed in terms of evolution of the surface characteristics of the ATF materials.

Feasibility assessment of the alumina‐forming duplex stainless steels as accident tolerant fuel cladding materials for light water reactors

Previously, the Fe-based alumina-forming duplex stainless steels (ADSSs) were developed for the application of accident tolerant fuel (ATF) cladding materials for light water reactors (LWRs). The on-going research activities focusing on the feasibility assessment of ADSS alloys for ATF cladding are summarized. A long-term corrosion behavior in simulated LWR environment and short-term corrosion resistance in steam environment at temperature range of 800°C to 1200°C were performed, and the results showed excellent corrosion resistance of ADSS alloys. The tensile properties of ADSS alloys were evaluated at room temperature to 550°C, which showed much higher strength compared with other Fe-based alloys. After accelerated thermal aging at 425°C for 1000 hours, strength of ADSS alloys was increased and the elongation was decreased. However, the elongation of aged ADSS alloys was still greater than 15% because of the presence of ductile austenite phase. Meanwhile, because of the neutron absorption by Ni, Cr, and Fe, the use of ADSS alloys as ATF cladding would have detrimental effects on the fuel cycle length. Nonetheless, it was assessed that the neutronics penalty could be readily overcome by adopting thinner fuel cladding and slightly higher fuel enrichment. Finally, the fabrication of thin-walled tube was introduced. Overall, it has been shown that ADSS alloys could be considered as candidate alloys as ATF cladding materials.

Cold-rolling & annealing process for nuclear grade wrought FeCrAl cladding alloy to enhance the strength and ductility

Abstract FeCrAl alloy is a good candidate as accident-tolerant fuel (ATF) cladding material. While cold-working is a preferred routine for fabricating thin-wall FeCrAl tubes, the effects of cold-rolling treatment on the microstructural features and the tensile properties are still unclear. In this study, a successive process containing hot-rolling and annealing at 800°C, cold-rolling at room temperature and final annealing at 750°C were developed for a wrought Fe-13Cr-4.5Al-2Mo-2.5Nb-0.1Ti- 0.1V-0.05Y alloy. The alloy with 60% hot-rolling and 40% cold-rolling reductions exhibited an ultimate tensile strength of 869MPa and a total elongation of 20.9%. The good mechanical properties were attributed to the recovery microstructure with much refined grain size, the dense dislocation network forming the sub-grain boundaries and the Laves precipitates distributed both along the grain or sub-grain boundaries and in the ferritic matrix. The alloy also exhibited a weak deformation texture. The satisfactory mechanical properties together with the nearly isotropic grain morphology and the weak rolling texture provide it good potential for application as ATF cladding materials.

Evaluation of critical heat flux of ATF candidate coating materials in pool boiling

The critical heat flux (CHF) of potential accident tolerant fuel (ATF) cladding materials and concepts, Cr and FeCrAl coatings on zirconium-alloy substrates, was investigated experimentally. A cold spray process, which involves the propulsion of high velocity powder particles on to the substrate surface to form a coating upon impact, was used to create the coatings. Surface parameters such as contact angle, surface roughness, and morphology that are associated with CHF in pool boiling were characterized. Non-hydrodynamic effects of newly developed ATF candidate coating materials on CHF were also evaluated. The effects of thermal properties and coating thickness and expected challenges for evaluation of CHF for new ATF cladding materials development are discussed.

Effects of neutron irradiation on 12Cr–6Al-ODS steel with electron-beam weld line

Oxide dispersion strengthened (ODS) steels have been developed as accident tolerant fuel cladding materials because of their excellent high temperature oxidation resistance as well as high temperature mechanical properties. 12Cr–6Al-ODS steel specimens with electron-beam (EB) weld line were irradiated with neutron in high flux isotope reactor (HFIR) at 300 °C to 2.6 dpa. Irradiation hardening was measured by Vickers hardness (HV) method. The irradiation resulted in a hardening of ∼95 HV in the base metal and ∼140 HV in EB weld line. The post-irradiation microstructures including the stability of the oxide nanoparticles was examined by transmission electron microscopy (TEM). The dispersion morphology of oxide particles in base metal was rather stable under present irradiation condition as no significant change was observed with respect to their average size and number density. As for the oxides in EB weld line, a similar trend with base metal was recognized. The 1/2 < 111 > and < 100> type dislocation loops formed in both base metal and EB weld line after neutron irradiation. The average size of dislocation loops was measured to be 7.6 ± 3.8 and 10.0 ± 3.8 nm in base metal and EB weld line, respectively. Meanwhile, the total number density was estimated to be (1.6 ± 1.0) × 1022 and (2.2 ± 0.5) × 1022 m−3 in base metal and EB weld line, respectively. Oxide/matrix interface may act as a trapping site for irradiation induced defects, leading to the smaller loop size and number density in base metal than in EB weld line. Dispersed barrier hardening (DBH) model indicated that the irradiation hardening could be attributed to the formation of dislocation loops.

Theory-guided bottom-up design of the FeCrAl alloys as accident tolerant fuel cladding materials

As interest in new generations of nuclear reactors is increasing worldwide, development of accident tolerant fuel (ATF) cladding materials is crucial but suffers from large amounts of time and experimental efforts. Here, we present a strategy of the theory-guided bottom-up design of ATF cladding materials using ab initio calculations and the CALPHAD (CALculation of PHAse Diagrams) approach in combination with key experiments, which has been utilized to investigate the FeCrAl system. The thermochemical, physical and elastic mechanical properties, as well as the phase diagrams of the FeCrAl alloys, are well described using both modeling and experiment. Apart from the reported high-temperature steam oxidation resistance as the initial criterion, the present selection and optimization of the FeCrAl alloys as accident tolerant fuel cladding materials is based on three constraints: (i) single phase at 320 °C, a service temperature in a light water reactor; (ii) high melting point; (iii) good elastic mechanical property. Further optimization takes into account of alloy performance in literature. Consequently, a compositional range has been proposed as a guidance for the fabrication of new FeCrAl alloys, i.e., 0 at.% < Cr content ≤15 at.% and 11 at.% ≤ Al content ≤21 at.%. The present scheme may enhance the efficiency in designing new ATF cladding materials by promoting the discovery of the composition-structure-property-performance relationships.

Development of alumina-forming duplex stainless steels as accident-tolerant fuel cladding materials for light water reactors

High aluminum content (>5 wt.%) duplex stainless steels were developed as accident-tolerant fuel (ATF) cladding materials for light water reactors. A set of model alumina-forming duplex stainless steel (ADSS) alloys with a nominal composition range of Fe-(18–21)Ni-(16–21)Cr-(5–6)Al were prepared and tested for corrosion resistance in both 1200 °C steam and simulated pressurized water reactor (PWR) operating conditions. The performance and properties of the ADSS alloys were compared to those of the austenitic (310 S) and ferritic (FeCrAl-APM) stainless steels. The results showed that the ADSS alloys possessed good corrosion resistances in both conditions, and better tensile properties. The applicability of the ADSS alloys as ATF cladding materials was discussed in view of their corrosion resistance, irradiation resistance, and embrittlement.