Materials In Light Water Reactors Research Articles

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

Enhanced Mechanical Properties and Wear Resistance of FeCrAl Alloys at ~ 300°C and Higher Temperatures

AbstractWorldwide, light water reactors (LWRs) have been using zirconium (Zr)-based alloys for the cladding of the uranium dioxide fuel for more than 6 decades. Zr alloys oxidize rapidly in the presence of water and steam at temperatures > 450°C; therefore, they do not respond well to scenarios of loss of coolant accidents. There is a global effort by nuclear materials technologists to find more robust or stronger cladding materials for LWRs. One option is to use an iron-chromium-aluminum (FeCrAl) alloy since they have excellent resistance to high temperature oxidation and superior mechanical properties at LWR operation temperatures. Results show that (1) FeCrAl alloys have better mechanical properties than Zr alloy and are orders of magnitude more resistant to creep at temperatures higher than LWR normal operation conditions. (2) FeCrAl alloys have better resistance to fretting wear than Zr alloys at the normal operation conditions of LWRs.

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.

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.

Development Status of Control Rod Neutron Absorber Materials for Light Water Reactors with Extended Control Rod Lifetime and Enhanced Safety

The control rod neutron absorbers for light water reactors with extended control rod lifetime and enhanced safety are being developed. It is an innovative material concept that can extend the lifetime by reducing irradiation-induced swelling compared to the existing control rod neutron absorber materials by utilizing the oxide-based materials. Furthermore, it can enhance the safety by improving various characteristics such as oxidation, eutectic reaction, volatilization behaviors of neutron absorber material. First, the oxide-based composition composed of a combination of neutron absorbers and structure stabilizers has been designed as material candidates for neutron absorber materials for control rods. The preliminary manufacturability test of the designed neutron absorber material candidates, and it has been confirmed a possibility that the designed material candidates can be fabricated using the conventional ceramic manufacturing process. In addition, the neutronic calculation/analysis and various out-of-pile tests (neutron absorber phase stability, oxidation and corrosion resistance, interaction and eutectic reaction between neutron absorber material and control rod tube material, melting behavior, etc.) are in progress. It has been also preparing the in-reactor swelling test of the selected material candidates at the HANARO research reactor in KAERI.

A Rapid Sintering Method for Cerium Nitride Pellet: A Uranium Mononitride Surrogate

Uranium mononitride (UN) is a candidate fuel material for light water reactors with higher uranium (U) loading and thermal conductivity than uranium dioxide (UO2). However, the sintering of UN pellets is challenging as the UN powder particles oxidize rapidly at high temperatures unless the oxygen concentration is extremely low. Oxidation during sintering either reduces the relative density of the sintered UN pellet or disintegrates the sintered UN pellet to powder. To address this problem, the present work developed a rapid sintering method for producing highly densified UN surrogate pellets with minimal oxidation. Cerium nitride (CeN) is used as a surrogate for UN to reduce radiation hazards. With the custom-developed fast-heating system, the sintering process was completed within 150 s. The sintering atmosphere was flowing nitrogen (N2). The sintered CeN pellet density was 95% of the theoretical density (TD) or higher. The microstructure was uniform with a 10–25 µm grain size as demonstrated by scanning electron microscopy (SEM) and contained trivial levels of oxides as demonstrated by X-ray diffraction (XRD). The resultant pellets indicate that the rapid sintering method is a promising method to make UN fuel pellets with equivalent or higher density to pellets made by conventional sintering methods, while also being more efficient in time and costs.

Microstructure Investigation of Mechanically Alloyed 316L-6Al-0.5Y2O3 Oxide Dispersion Strengthened Austenitic Steel Powder

Development of accident-tolerant fuel cladding (ATFC) materials for LWRs (Light Water Reactors) is being intensively carried out in many research centers in the world as the lesson-learned from the Fukushima Daiichi nuclear reactor power plant accident. One of the advanced high temperature materials which is intensively developed is Oxide Dispersion Strengthened (ODS) steel. The strengthened mechanism is generated by homogeneous dispersion of nano-meter sized ceramic oxide particles in the matrix of the steel which is done by mechanical alloying technique so called mechanosynthesis. Synthesis of an austenitic ODS steel powder of 316L with additional element of 6 wt.% aluminum and dispersion of nano-particles of 0.5 wt.% yttrium oxide (yttria: Y2O3) has been done. The aim of the research is to achieve a better performance of 316L austenitic steel as the proven material for power plant at elevated temperature by modified to become ODS austenitic steel with additional of aluminum. Mechanosynthesis was done using High Energy ball Milling (HEM) for 30 hours with BPR (Ball to Powder Ratio) of 10:1. Afterward, the sample was characterized using Scanning Electron Microscope-Energy Dispersive Spectroscope (SEM-EDS) and X-ray Diffraction (XRD) to analyze the microstructure characteristics. The results showed that the alloying between 316L and aluminum powders was formed. Furthermore, crystal grains and the particles of 316L-6Al-0.5Y2O3 steel powder was much smaller than as received 316L austenitic steel and aluminium powders. The particles size and element composition was homogeneously distributed. The results showed that the parameters of mechanical alloying for development of 316L-6Al-0.5Y2O3 ODS austenitic steel powder have reached the optimum condition.

Development of Liquid Phase Sintering Silicon Carbide Composites for Light Water Reactor

Silicon carbide composites are expected for light water reactors. The objective is to understand the steam oxidation behavior and the high-temperature water corrosion behavior of the liquid phase sintering silicon carbide and to develop the liquid phase sintering silicon carbide composites, which are stable at the high-temperature water conditions in normal operation and the high-temperature steam conditions in a severe accident. The steam oxidation experiments were carried out at 1200 and 1400 °C. The high-temperature water corrosion experiments were carried out at 320 and 360 °C. The formation of the silicate, which is expected to have excellent resistance to the steam, was confirmed following the steam exposure at 1400 °C. High-temperature water corrosion resistance was improved by the formation of Yttrium Aluminum Garnet at the grain boundary. The particle-dispersion silicon carbide composite tubes with the modified condition were developed, and the thermal shock experiments from 1200 °C to ambient temperature were carried out. The composite tubes showed excellent oxidation and thermal shock resistance. The particle-dispersion liquid phase sintering silicon carbide composites with the modified condition are promising materials for light water reactors.

Development of an Oxide Dispersion Strengthened Austenitic Steel Powder with 4% Aluminum Addition

Development of accident-tolerant fuel (ATF) cladding materials for LWR (Light Water Reactors) is being intensively carried out in many research centers in the world as the lesson-learned from the Fukushima Daiichi nuclear reactor power plant accident. One of the advanced high temperature materials which is intensively developed is Oxide Dispersion strengthened (ODS) steel. The strengthened mechanism is generated by homogeneous dispersion of nanometer sized ceramic oxide particles in the matrix of the steel which is done by mechanical alloying technique so called mechanosynthesis. Synthesis of an austenitic ODS steel powder of SS316L with additional element of 4% aluminum and dispersion of nanoparticles of 0.5% yttrium oxide (yttria: Y2O3) has been done. The aim of the research is to achieve a better performance of SS316L steel as the established material for power plant at elevated temperature. Mechanosynthesis was done using High Energy ball Milling (HEM) for 30 hours with BPR (Ball to Powder Ratio) of 10:1. Afterward, the sample was characterized using Scanning Electron Microscope-Energy Dispersive Spectroscope (SEM-EDS) and X-ray diffraction (XRD) to analyze the microstructure characteristics. The results showed that the alloying between SS316L and aluminum powders was formed. Furthermore, crystal grains and the particles of SS316L-4Al-0.5Y2O3 powder was smaller than as received SS316L and aluminum powders.

On the feasibility of duplex stainless steel 2205 as an accident tolerant fuel cladding material for light water reactors

A few tests were performed to evaluate the feasibility of 2205 duplex stainless steel (DSS) as an accident tolerant fuel (ATF) cladding material, including corrosion tests in simulated pressure water reactor (PWR) primary water and high-temperature steam, stress corrosion cracking (SCC) initiation test, mechanical properties test and thermal aging effects. In both 360 °C PWR primary water and 1200 °C steam, DSS 2205 showed weight changes comparable to the high-Cr alloys. After 3163 h of exposure in 400 °C steam, U-bend specimens of DSS 2205 showed no obvious cracks, suggesting that SCC is not likely to be initiated during the service life. Compared to FeCrAl alloys, DSS 2205 showed comparable strength but greater ductility. Though 400 °C thermal aging caused an increase in strength for DSS 2205, total elongation was more than 35% even after 11,061 h aging, greater than other Fe-based ATF alloys. Also, concerns on neutron penalty, irradiation, fabrication and spent fuel storage were discussed. The overall results suggest that DSS 2205 could be considered as a candidate ATF cladding material.

Effect of Nb content on the precipitation behaviors of Fe-Cr-Al fuel cladding alloys during long-term aging at 475 °C

The precipitation behavior and tensile properties of Fe-13Cr-6Al-2Mo-xNb (x = 0.15, 0.5, 1.0) alloys for fuel cladding materials were investigated during aging at 475 °C up to 8000 h. During the aging process, the precipitation of the metastable (Nb, Mo)-Al-rich phase was promoted with Nb content. Meanwhile, the dominant precipitation of the μ-phase on the grain boundaries transited to the Laves phase with Nb content. With aging, the yield strength of the alloys increased firstly and then kept almost unchanged while the elongation decreased, owing to the precipitation and growth of different second phases. The grain boundary μ-phase and secondary Laves phase led to the transformation from ductile to brittle fracture. After long-term aging at 475 °C, the 0.5Nb alloy possessed the best combination of yield strength and elongation, making it more suitable for application as fuel cladding materials in light-water reactors.

Molecular dynamic simulations evaluating the effect of the stacking fault energy on defect formations in face-centered cubic metals subjected to high-energy particle irradiation

Austenitic stainless steels, which are used as incore structural materials in light water reactors, are characterized by an extremely low stacking fault energy (SFE) among face-centered cubic (FCC) metals. To evaluate the effects of SFE on defect formation under high-energy particle irradiation, molecular dynamics simulations were performed using the interatomic potential sets for FCC metals with different SFEs and a primary knock-on atom energy (EPKA) of 100 keV at 600 K. The results show that the number of residual defects is independent of the SFE. However, the characteristics of self-interstitial atom (SIA) clusters do depend on the SFE. For clusters smaller than a certain size, the ratio of glissile SIA clusters decreases as the SFE increases, which is similar to the trend observed at the low EPKA. However, for larger clusters, which can be detected only at a high EPKA, the ratio of glissile clusters increases. These results correspond to static energy calculations, in which the difference in the formation energy between a Frank loop and perfect loop (ΔEF-P) for the small clusters decreases as the SFE increases. In contrast, for the larger clusters, the SFE dependence of ΔEF-P changes due to the shape restrictions of stable perfect loops. At a high temperature of 600 K, large vacancy clusters with stacking faults can be detected at EPKA = 100 keV, resulting in the enhanced formation of these clusters at lower SFEs. Furthermore, several of these clusters were similar to perfect loops, with the edges split into two partial dislocations with stacking faults, although the largest clusters detected at low EPKAs were similar to stacking fault tetrahedrons.

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.

Current-driving dissolution of nanoscale brittle precipitates produced by spinodal decomposition in FeCrAl alloys

Due to its excellent resistance to high temperature oxidation and steam corrosion, FeCrAl alloys have received much attention recently as a highly potential candidate cladding material for light water reactors. However, under the effect of thermal aging and neutron irradiation applied in reactor, the nanoscale brittle precipitates will occur in the ferritic FeCrAl alloys after spinodal decomposition due to the presence of Cr, which greatly increase the hardness and reduce the toughness of the material, thus seriously affecting the operation of the nuclear reactor. In this work, the alloy after heat aging was treated by pulsed electric current at room temperature. The micro-hardness and electrical properties of the aged alloy are greatly restored. In addition, compared with the heat treatment at the same temperature, the dissolution time of precipitates in aged alloys into the matrix is reduced by 10 times by electropulsing treatment with higher parameters. In comparison to the ex-situ conventional annealing treatment, this in-situ pulsed treatment can be used as a novel efficient and convenient method to repair the performance degradation of aged materials. Based on the thermodynamic and kinetics, electric pulses reduces the thermodynamic barrier for precipitation dissolution and increases the atomic diffusion rate.

First principles study of point defect effects on iodine diffusion in zirconium

The formation energies, diffusion barriers, and vibration frequencies of complex iodine defects in zirconium are calculated with first principles calculations and nudged elastic band method. The effective diffusion rates of these complex defects are evaluated in the temperature range from 300 K to 2000 K. The iodine interstitial diffusion and the iodine diffusion via IZr−VZr are the dominant diffusion mechanisms. The vacancies may be traps for iodine diffusion and they can slow down the diffusion of iodine atoms. For the zirconium cladding materials in light water reactors, various point defects may be generated. Our calculated effective diffusion coefficients agree with the experimental data well and it indicates that bulk diffusion may be a main diffusion mechanism for the iodine diffusion in zirconium alloy.

Microstructure change of duplex stainless steels after thermal aging and electron beam irradiation

Duplex stainless steel has been used as a structural material for light water reactors. It is well known that the thermal aging during operation causes spinodal decomposition to Cr-rich (α') phase and Fe-rich (α) phase in the ferrite phase, resulting in brittlement. In order to understand the mechanism of this phenomena and the effect of irradiation, a duplex model alloy (Fe-25Cr-10Ni-2.5Mo-1Mn) prepared by arc melting and thermally aged in the appropriate condition was subjected to accelerated irradiation by a multi-beam ultra-high voltage electron microscope and partially an ion accelerator in this study. Spinodal decomposition of Fe and Cr was confirmed in the model alloy aged at 450 °C. In addition, the electron beam irradiation to aged model alloy resulted in the decrease in the amplitude of spinodal decomposition. From this experiment, it was suggested that spinodal decomposition could be suppressed by accelerated irradiation.

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.

Vanadium, rhodium, silver and cobalt self-powered neutron detector calculations by RAST-K v2.0

Vanadium, rhodium, silver and cobalt Self-Powered Neutron Detectors (SPNDs) are evaluated and compared in this study, including their characteristic response time. The module for response calculations of different SPNDs has been programmed and added to RAST-K v2.0 (Reactor Analysis code for Steady-state and Transient). For simplification purposes and running time minimization, previously studies have often adopted the simplified depletion chain of the SPNDs. The isomers are often neglected in the detectors studies because of their low levels of contribution to the final signal. Besides, few published documents can be found for comparisons of the behavior of different emitters in normal power changes, especially for silver and cobalt emitters. In this paper, the full depletion chains for different emitters are adopted, and the isomers’ contributions are calculated. A new method for detector response calculation based on Chebyshev Rational Approximation Method (CRAM) is used as the depletion method. It is shown that the response of silver SPND is very similar to that of rhodium SPND and can be considered as an alternate emitter material for light water reactors. This study can contribute to the selection of proper SPNDs in specific applications.

Initial experimental evaluation of crud-resistant materials for light water reactors

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

Effect of He implantation on the microstructure of zircaloy-4 studied using in situ TEM

Zirconium alloys are of great importance to the nuclear industry as they have been widely used as cladding materials in light-water reactors since the 1960s. This work examines the behaviour of these alloys under He ion implantation for the purposes of developing understanding of the fundamental processes behind their response to irradiation. Characterization of zircaloy-4 samples using TEM with in situ 6 keV He irradiation up to a fluence of 2.7×1017ions·cm−2 in the temperature range of 298 to 1148 K has been performed. Ordered arrays of He bubbles were observed at 473 and 1148 K at a fluence of 1.7×1017ions·cm−2 in αZr, the hexagonal compact (HCP) and in βZr, the body centred cubic (BCC) phases, respectively. In addition, the dissolution behaviour of cubic Zr hydrides under He irradiation has been investigated.