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Journal articles on the topic 'Fuel cladding'
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1
Kobylyansky, G. P., А. О. Mazaev, Е. А. Zvir, S. G. Eremin, Е. V. Chertopyatov, and А. V. Obukhov. "The effect of long-term annealing simulating the parameters of dry storage of VVER-1000 fuel rods on the mechanical properties of E110 alloy shells in the longitudinal direction." Physics and Chemistry of Materials Treatment 4 (2021): 42–49. http://dx.doi.org/10.30791/0015-3214-2021-4-42-49.
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Presented are the results of mechanical tensile tests of longitudinal (segmental) samples cut from the midsection of claddings spent as VVER-1000 FA during one- and six-year campaigns and subject to thermal tests in helium at 480 °С during 468 full days. An average burnup of these fuel rods achieved ~ 20 and ~ 70 (MW·day)/kg U, respectively. The tests followed the examinations for cladding mechanical properties performed using the tests results for ring samples cut from the specified fuel rods. These fuel rods were tested in the experiments to determine impact of long-term thermal tests that model dry storage conditions on mechanical properties of Zr E110 claddings. Based on mechanical tests results at room temperature and at 380 °С there was determined as follows: ultimate strength sв, yield strength s0,2 and total relative elongation d0 of claddings length-wise on the fuel rod segments at the fuel column midsection. The obtained characteristics were compared to corresponding values for initial (unirradiated) cladding tubes and mechanical test results of the ring samples in the transverse direction. Long-term thermal tests have led to partial return to initial (before operation) values sв, s0,2 and d0 of radiation-hardened claddings; this return was more prominent in the longitudinal direction than in the transverse one. A radiation hardening decrease was accompanied with an increase in total relative elongation values in both cladding directions. Anisotropy of yield strength has changed as well. These changes can be explained by partial annealing of radiation defects, which are obstacles to dislocation movements during cladding strain. The morphology of above radiation defects is different in various sliding planes in texturized grains of cladding material.
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Ivanov, Sergey N., Sergey I. Porollo, Sergey V. Shulepin, Yury D. Baranaev, Vladimir F. Timofeev, and Yury V. Kharizomenov. "Examination of fuel elements irradiated in the reactor of the World’s First NPP after long-term storage." Nuclear Energy and Technology 9, no. 1 (March 17, 2023): 51–58. http://dx.doi.org/10.3897/nucet.9.102492.
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Examinations of fuel elements with two different fuel compositions, U-Mo+Mg and UO2+Mg, irradiated in the AM reactor after their long-term storage do not reveal any visible defects on the surface of their outer claddings. However, in the fuel elements with U-Mo fuel, an increase in the diameter of the outer cladding is observed. This is most noticeable in the upper part of the fuel element. Storage of the fuel elements with UO2 fuel for 15–22 years does not lead to a change in their diameter within the measurement accuracy. At the same time, metallographic studies have shown that on the external surface of the outer cladding and the internal surface of the inner cladding of the fuel elements with U-Mo+Mg and UO2+Mg fuel compositions, after long-term storage, defects are observed in the form of intergranular and irregular frontal corrosion, pits and pittings up to 20 µm deep. No interaction is found at the points of contact between the fuel claddings and the fuel composition of the layers. There is no noticeable decrease in the thickness of the outer and inner claddings of the fuel elements after long-term storage, nor does the thickness of the claddings at the locations of defects go beyond its minimum initial value, taking into account the technological tolerance for variations in thickness. It is noteworthy, however, that cracks are found in both types of fuel elements both in the fuel grains and in the magnesium matrix. As a result of long-term storage of the fuel elements with U-Mo fuel for 45–55 years, the mechanical properties of their outer claddings gradually degrade, due to which the plasticity of the cladding is significantly reduced.
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Lys, Stepan, Igor Galyanchuk, and Tetiana Kovalenko. "Prediction of thermophysical characteristics of fuel rods based on calculations." Energy engineering and control systems 7, no. 2 (2021): 79–86. http://dx.doi.org/10.23939/jeecs2021.02.079.
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The paper analyzes operating conditions, thermophysical characteristics were calculated as applied to WWER-1000 fuel rods in a four-year cycle for unified core. The paper gives a summary of models for calculating gas release, pressure of gases within fuel rod cladding, fuel swelling and thermal conductivity, fuel-cladding gap conductance. The thermophysical condition of fuels in a reactor core is one of the main factors that determine their serviceability. The stress-strained condition of fuel claddings under design operating conditions is closely related to fuel rod temperature, swelling, gas release from fuel pellets and the mode in which they change during the cycle and transients. Aside from this, those parameters are an independent goal of studies since their ultimate values are governed by the system of design criteria.
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Alrwashdeh, Mohammad, and Saeed A. Alameri. "SiC and FeCrAl as Potential Cladding Materials for APR-1400 Neutronic Analysis." Energies 15, no. 10 (May 20, 2022): 3772. http://dx.doi.org/10.3390/en15103772.
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The aim of this study is to investigate the potential improvement of accident-tolerant fuels in pressurized water reactors for replacing existing reference zircaloy (Zr) fuel-cladding systems. Three main strategies for improving accident-tolerant fuels are investigated: enhancement of the present state-of-the-art zirconium fuel-cladding system to improve oxidation resistance, replacement of the current referenced fuel-cladding system material with an alternative high-performance oxidation-resistant cladding, and replacement of the current fuel with alternative fuel forms. This study focuses on a preliminary analysis of the neutronic behavior and properties of silicon carbide (SiC)-fuel and FeCrAl cladding systems, which provide a better safety margin as accident-tolerant fuel systems for pressurized water reactors. The typical physical behavior of both cladding systems is investigated to determine their general neutronic performance. The multiplication factor, thermal neutron flux spectrum, 239Pu inventory, pin power distribution, and radial power are analyzed and compared with those of a reference Zr fuel-cladding system. Furthermore, the effects of a burnable poison rod (Gd2O3) in different fuel assemblies are investigated. SiC cladding assemblies present a softer neutron spectrum and a lower linear power distribution compared with the conventional Zr-fuel-cladding system. Additionally, the SiC fuel-cladding system exhibits behaviors that are consistent with the neutronic behavior of conventional Zr fuel-cladding systems, thereby affording greater economic and safety improvements.
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Li, Jing, Sa Jian Wu, Yong Li Wang, Liang Yin Xiong, and Shi Liu. "Performance of 14Cr ODS-FeCrAl Cladding Tube for Accident Tolerant Fuel." Materials Science Forum 1016 (January 2021): 806–12. http://dx.doi.org/10.4028/www.scientific.net/msf.1016.806.
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In the framework of Accident tolerant fuel (ATF) program, several types of claddings and pellets with enhanced accident tolerance have been developed for light water reactors. Oxide dispersion strengthened (ODS) FeCrAl alloys have been considered as a promising candidate for cladding materials due to their good mechanical strength, excellent structural stability and chemical durability at high temperature. The out-of-pile performance of 14Cr ODS-FeCrAl cladding tube fabricated by cold-rolling, such as microstructure, thermophysical property, mechanical property, and corrosion resistance, has been examined and discussed. The results confirm that iron-based ODS alloy is one of the promising candidates to be used as ATF cladding. It could also aid in the supplement of property database of ODS-FeCrAl for future use in nuclear cladding and structural applications in next generation nuclear systems.
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Yakushkin, A. A. "On the problems of creating shells of fuel rods from zirconium alloys for tolerant fuel." Physics and Chemistry of Materials Treatment 3 (2021): 69–78. http://dx.doi.org/10.30791/0015-3214-2021-3-69-78.
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Three directions of the establishment of accident tolerant fuel cladding for light water reactors are actively exploring at present: 1) replacement zirconium alloy E110 for more corrosion-resistant material in accident operation conditions; 2) surface dispersion hardening or doping of the zirconium cladding of fuel element; 3) deposition a corrosion-resistant coating to the fuel cladding. The first direction requires significant and irreversible changes in fuel rod production technology and has long-term prospects. Conversely, the second direction suggest minimal changes in the fuel rod production technology, however, it has no significant effect on the high temperature oxidation kinetics of fuel claddings in steam. Using of a corrosion resistant coating results in a significant change in the high temperature oxidation kinetics of the zirconium alloy, (no transition to linear oxidation) that is related to maintaining the continuity of the oxide layer formed during oxidation. The issue provides a brief overview of the current state of research in the field of fuel, tolerant to the effects of coolant in emergency situations.
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Halabuk, Dávid, and Jiří Martinec. "CALCULATION OF STRESS AND DEFORMATION IN FUEL ROD CLADDING DURING PELLET-CLADDING INTERACTION." Acta Polytechnica 55, no. 6 (December 31, 2015): 384. http://dx.doi.org/10.14311/ap.2015.55.0384.
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The elementary parts of every fuel assembly, and thus of the reactor core, are fuel rods. The main function of cladding is hermetic separation of nuclear fuel from coolant. The fuel rod works in very specific and difficult conditions, so there are high requirements on its reliability and safety. During irradiation of fuel rods, a state may occur when fuel pellet and cladding interact. This state is followed by changes of stress and deformations in the fuel cladding. The article is focused on stress and deformation analysis of fuel cladding, where two fuels are compared: a fresh one and a spent one, which is in contact with cladding. The calculations are done for 4 different shapes of fuel pellets. It is possible to evaluate which shape of fuel pellet is the most appropriate in consideration of stress and deformation forming in fuel cladding, axial dilatation of fuel, and radial temperature distribution in the fuel rod, based on the obtained results.
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Newell, Ryan, Abhishek Mehta, Young Joo Park, Yong Ho Sohn, Jan Fong Jue, and Dennis D. Keiser Jr. "Relating Diffusion Couple Experiment Results to Observed As-Fabricated Microstructures in Low-Enriched U-10wt.% Mo Monolithic Fuel Plates." Defect and Diffusion Forum 375 (May 2017): 18–28. http://dx.doi.org/10.4028/www.scientific.net/ddf.375.18.
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Monolithic fuel system with U – 10 wt.% Mo (U10Mo) fuel alloy has been developed for the Materials Management and Minimization reactor conversion program to replace highly-enriched fuels in research and test reactors with low-enriched fuels. Interdiffusion and phase transformations in the system constituents, i.e., U10Mo fuel, AA6061 cladding, and Zr diffusion barrier, have been investigated using fuel plates fabricated via rolling and hot-isostatic pressing. Diffusion couples, utilizing the constituents of the fuel system were also carried out to help understand the findings from fuel plates based on phase equilibria and diffusion kinetics. Findings from both fuel plates and diffusion couples can provide a comprehensive knowledge to assess, model, and predict the performance of monolithic low-enriched fuel system from fabrication to irradiation. This paper summarizes the experimental results reported from characterization of the fuel plates and diffusion couples with emphasis on interactions at the fuel-cladding, fuel-diffusion barrier, cladding-diffusion barrier, and cladding-cladding interfaces. Constituent phases and relevant diffusion kinetics are compared and contrasted, taking into account differences in thermodynamics and kinetics variables such as pressure, temperature, and cooling rate.
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Gávelová, Petra, Patricie Halodová, Ondřej Libera, Iveta Adéla Prokůpková, Věra Vrtílková, and Jakub Krejčí. "Experimental Verification of Phase Diagram Calculations of Zr-Based Alloys after High-Temperature Oxidation." Defect and Diffusion Forum 405 (November 2020): 351–56. http://dx.doi.org/10.4028/www.scientific.net/ddf.405.351.
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Zirconium-based alloys are commonly used as a material for nuclear fuel claddings in the light water reactors. The cladding material must function to fix a huge number of pellets, while conducting heat into the coolant that flows turbulently around the fuel rods. Cladding tubes can contain gaseous fission products that escape the fuel. Thus, by functioning as a sealed unit, it prevents a contamination of the coolant water with high-radioactive fission products. The integrity of claddings is always a critical issue during reactor operation and wet or dry storage and transport of the spent fuel rods. Moreover, the role gains importance at Loss of Coolant Accidents (LOCA). After Fukushima accident, cladding materials are widely studied with the purpose to reduce the high-temperature oxidation rate and enhance accident tolerance. In our contribution, we introduce the studies on Zr-1Nb (E110) cladding tubes after high-temperature steam oxidation at 1350 °C. During the testing of claddings, microscopy analytical methods play an important role in experimental verification of pseudo-binary phase diagram Zr1Nb-O, i. e. particularly in oxygen content determination at phase transitions. Wave Dispersive Spectroscopy (WDS) with complementary nano-indentation method were used to characterize the Zr1Nb microstructure formed after LOCA. It includes the regions from an oxide and oxygen-stabilized α-Zr(O) to the acicular prior β-Zr phase. The decrease of hardness and Young's modulus corresponds with oxygen content measured in line-profiles by WDS. The oxygen level at transition points was partly determined from Fe, Nb β-stabilizers and significant change in mechanical properties in fine-grained prior β-Zr. The slight fluctuation of oxygen values in adjacent grains can be caused by preferential oxidation through the favorably oriented α-Zr(O) grains studied by WDS+EBSD. As well, the non-uniform oxygen-rich α-Zr(O) phase adjacent to the oxide was characterized by EBSD & WDS. Increasing hydrogen content in specimens, 10, 700 and 1000 ppm H, caused increasing solubility of oxygen in prior β-Zr phase upon high-temperature and the cladding material hardening.
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Tarı, Doğaç, Teodora Retegan Vollmer, and Christine Geers. "High Temperature Corrosion Behavior of 15-15Ti Cladding Tube Material in Contact with Liquid Lead, Outside, and Cs2MoO4, Inside." ECS Meeting Abstracts MA2023-02, no. 12 (December 22, 2023): 1107. http://dx.doi.org/10.1149/ma2023-02121107mtgabs.
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Lead-Cooled Fast Reactors are one of the emerging new technologies connected to the Generation IV reactor designs. Even though the reactor features are extensive and beneficial, some technical and regulatory difficulties remain that hinder the deployment. One of them is the corrosion behavior of fuel claddings in high burn-up and elevated temperatures, as fuel claddings are a part of accident barriers. In this study, 15-15Ti austenitic steel was the investigated cladding material. As this cladding would be used in a tube form, corrosion attack from both sides, inside and outside is studied in contact to their respective environments. From the outside of the tube, the proposed coolant liquid lead is the corrosive substance. At the inside of the tube, the major corrosive species in our test setup is, for simplicity reasons, Cs2MoO4. However, this choice has been made from some considerations regarding the fuel evolution upon burn-up. In a high burn-up situation, the generated fission products would start to accumulate to the gap between the fuel pellet and the cladding tube. These accumulated products are the so-called “Joint-Oxide Gain” (JOG) phases, and they are the corrosive substances in question when discussing the corrosion attack from the inside of a cladding tube. Cs2MoO4 has been identified as one of the main components of JOG phases and thus will be focused on in this study. A new capsule was designed to investigate the dual-atmosphere corrosion of cladding tubes. Unexposed capsule parts (a) and the basic schematic of the assembled capsule (b) can be seen in the Figure below. The cladding tube material itself is used as the container for JOG as it is filled with Cs2MoO4 and sealed on both ends. This assembly is then placed inside a bigger capsule, which is filled with lead powder to submerge the cladding tube assembly. All assemblies were done in Ar-filled glovebox. The capsules were then exposed to 600-1000 oC for 52 to 168 hours, which was followed by cutting and epoxy embedding to investigate the cross-sections. Figure 1
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Dolganov, K. S., A. E. Tarasov, A. V. Kapustin, and D. Yu Tomashchik. "Numerical Investigation of Cladding Ballooning and Burst in VVER and PWR Fuel Rods in Experiments with Various Loading Conditions." Известия Российской академии наук. Энергетика, no. 3 (May 1, 2023): 57–78. http://dx.doi.org/10.31857/s0002331023030044.
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The paper presents the results of numerical modeling for ballooning and burst of fuel rods claddings made of domestic and foreign alloys. The integral code SOCRAT-V1/V2 is used as a means of modeling. The uncertainty analysis of the calculation results to input uncertainties of the temperature and pressure measurements was performed. The modeling results demonstrate a good qualitative and quantitative compliance with measured times of cladding failure under partial core uncovery conditions. The results of SOCRAT-V1/V2 validation evidence on the importance of performing new experiments for domestic fuel rod cladding ballooning and burst.
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Sotnikov, A. S. "Simulation of Initiation and Development of Corrosion Cracks in Zirconium Fuel Claddings Under Conditions of Stress Corrosion Cracking in Environment of Iodine." Herald of the Bauman Moscow State Technical University. Series Mechanical Engineering, no. 5 (128) (October 2019): 135–46. http://dx.doi.org/10.18698/0236-3941-2019-5-135-146.
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The process models of iodine corrosion cracking of zirconium fuel claddings, used to calculate the durability of the cladding (time for loss of tightness) are considered. A method for determining the corrosion crack propagation rate in claddings made of E110 alloy Ø 9.1 × 0.65 mm and the results of corresponding studies (estimation of corrosion crack propagation rate and stress intensity factor KISCC) are given at a temperature of 380 °C in iodine environment at a concentration of ~ 0.2 mg/cm2. Studies were performed using tubular samples with a fatigue crack. A fatigue crack on the inner surface of cladding made of E110 Ø 9.1 × 0.65 mm alloy is the initiator of a corrosion crack emergence (nucleation). The results of corresponding studies are consistent with data from the literature. The proposed study of the corrosion cracking process of fuel claddings in accordance with the results of fracture mechanics is of practical importance for substantiation of the regulation of reactor operating conditions
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Belash, Nikolay N., Anton V. Kushtym, Vladimir V. Zigunov, Elena A. Slabospytska, Gennadіy А. Holomeyev, Ruslan L. Vasilenko, and Аleksandr I. Tymoshenko. "Research and Development of Fuel Rods Metallurgically Bonded with Fuel Cladding for Nuclear Installations." 3, no. 3 (September 28, 2021): 110–15. http://dx.doi.org/10.26565/2312-4334-2021-3-17.
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The design and scheme for manufacturing fuel rods metallurgically bonded with ribbed aluminum claddings using hot isostatic pressing and contact-reactive brazing are presented. It is shown that the developed scheme can be used both for production of dispersive fuels and high-density fuels based on uranium alloys. The results of investigations of brazed joints of aluminum cladding with a matrix composition based on aluminum and with samples of E110 alloy through copper and silumin coatings are presented. The results of research of brazed joints of an aluminum cladding with an aluminum-based matrix composition and samples of zirconium alloy E110 made through copper and silumin coating are presented. The strength of brazed joints, composition of diffusion layers formed as a result of contact-reactive brazing in a high vacuum have been determined. The modes of hot isostatic pressing that provide crimping of the ribbed cladding of fuel pellets and rods and obtaining a metallurgical bonding between their surfaces have been defined. It is shown that satisfactory bond strength is provided starting from the temperature of 610 °С. The maximum strength values obtained on the compounds Al-(Al+12% Si)-Zr and Al-Cu-Zr are 57.0 MPa and 55.3 MPa respectively. The fracture of the of aluminum samples joints, obtained with the Cu layer at a temperature of 620 °C, occurs on threaded joints at the strength value of 82 MPa. The results of research of the composition of diffusion layers formed by brazing compounds Al-(Al + 12% Si)-Zr and Al-Cu-Zr are presented. It was established that hot pressing provides the best results for manufacturing of fuel rod dummies in the studied range of modes at a temperature of 630 °C, a pressure of 380 MPa and exposure of 20 minutes.
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Paaren, Kyle M., Micah Gale, David Wootan, Pavel Medvedev, and Douglas Porter. "Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data." Energies 16, no. 22 (November 16, 2023): 7600. http://dx.doi.org/10.3390/en16227600.
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Using the BISON fuel-performance code, simulations were conducted of an automated process to read initial and operating conditions from the Pacific Northwest National Laboratory (PNNL) database and reports, which contain metallic-fuel data from the Fast Flux Test Facility (FFTF) MFF Experiments. This work builds on previous modeling efforts involving 1977 EBR-II metallic fuel pins from experiments. Coupling the FFTF PNNL reports to BISON allowed for all 338 pins from MFF-3 and MFF-5 campaigns to be simulated. Each BISON simulation contains unique power and flux histories, axial power and flux profiles, and coolant-channel flow rates. Fission-gas release (FGR), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON and compared to available post-irradiation examination (PIE) data. Cladding profilometry, FGR, and fuel axial swelling simulation results for full-length MFF metallic pins were found to be in agreement with PIE measurements using FFTF physics and models used previously for EBR-II simulations. The main two peaks observed within the cladding profilometry were able to be simulated, with fuel-cladding mechanical interaction (FCMI), fuel-cladding chemical interaction (FCCI), and thermal and irradiation-induced creep being the cause. A U-Pu-Zr hot-pressing model was included in this work to allow pore collapse within the fuel matrix. This allowed better agreement between BISON-simulated cladding profilometry and PIE measurements for the peak caused by FCMI. This work shows that metallic fuel models used to accurately represent fuel performance for smaller EBR-II pins may be used for full-length metallic fuel, such as FFTF MFF assemblies and the Versatile Test Reactor (VTR). As new material models and PIE measurements become available, FFTF MFF assessment cases will be reassessed to further BISON model development.
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Cantini, Federico, Martina Adorni, and Francesco D’Auria. "Nuclear Fuel Modelling During Power Ramp." Journal of Energy - Energija 62, no. 1-4 (July 18, 2022): 68–80. http://dx.doi.org/10.37798/2013621-4219.
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Fuel rods operating for several years in a LWR can experience fuel-cladding gap closure as a result of the phenomena due to temperature and irradiation. Local power increase induces circumferential stresses in the cladding as a result of the different expansion in the cladding and the pellet. In presence of corrosive fission products (i.e. Iodine) and beyond specific stress threshold and level of burnup, cracks may grow-up from the internal to the external cladding surface, causing fuel rod failure. The phenomenon, known as pellet cladding interaction-stress corrosion cracking PCI/SCC, or PCI, has been identified as a problem since the 70's. The PWR Super-Ramp experiment (part of OECD/NEA “International Fuel Performance Experiments (IFPE) database”) twenty eight fuel rods behaviour has been simulated using TRANSURANUS code version “v1m1j11”. Two sets (“Reference” and “Improved”) of suitable input decks modelling the fuel rods, based on the available literature are used to run the simulations. Focus is given to the main phenomena which are involved or may influence the cladding failure. Systematic comparison of the code results with the experimental data are performed for the parameters relevant for the PCI phenomenon. Sensitivity calculations on fission gas release models implemented in TRANSURANUS code are also performed in order to address the impact on the results. The results show the ability of TRANSURANUS version “v1m1j11” in conservatively predicting the rods failure due to PCI in PWR fuel and Zircaloy-4 cladding. Increased availability of experimental data would help to perform a deeper analysis.
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Al Mahfudz, Alif, Alexander Agung, and Andang Widi Harto. "Investigation on Neutronic Parameters of the KLT-40S Reactor Core with U3Si2-FeCrAl using SCALE Code." Journal of Engineering and Technological Sciences 55, no. 1 (March 6, 2023): 22–30. http://dx.doi.org/10.5614/j.eng.technol.sci.2023.55.1.3.
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From a safety point of view, the fuel-cladding of the current design of the KLT-40S reactor still carries a potential risk in the event of a loss-of-coolant accident (LOCA) allowing the formation of hydrogen gas. The concept of accident tolerant fuels (ATF) offers a variety of new safer fuel-cladding materials, one of which is U3Si2-FeCrAl, a potential fuel-cladding combination according to various research sources. In this research, a study of neutronic parameters (1) cycle length, (2) reactivity feedback coefficient, and (3) reactor proliferation resistance was performed with ATF material U3Si2-FeCrAl as fuel-cladding in the KLT-40S reactor core. Modeling and simulation of the ATF-fueled KLT-40S reactor core were performed using KENO-VI and TRITON modules from SCALE code. The results showed that replacement of the fuel-cladding material with the ATF material in the KLT-40S reactor resulted in a shorter cycle length, and the enrichment required to reproduce the original cycle length was above the safeguard limit. The fuel temperature, moderator temperature, and void reactivity coefficient were negative, although not as negative as the original ones. The spent fuel produced at the end of the cycle had good proliferation resistance, although not as good as the original one.
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Paaren, Kyle M., Nancy Lybeck, Kun Mo, Pavel Medvedev, and Douglas Porter. "Cladding Profilometry Analysis of Experimental Breeder Reactor-II Metallic Fuel Pins with HT9, D9, and SS316 Cladding." Energies 14, no. 2 (January 19, 2021): 515. http://dx.doi.org/10.3390/en14020515.
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BISON finite element method fuel performance simulations were conducted using an existing automated process that couples the Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System database by writing input files and comparing the BISON output to post-irradiation fuel pin profilometry measurements contained within the databases. The importance of this work is to demonstrate the ability to benchmark fuel performance metallic fuel models within BISON using Experimental Breeder Reactor-II fuel pin data for a number of similar pins, while building off previous modeling efforts. Changes to the generic BISON input file include implementing pin specific axial power and flux profiles, pin specific fluences, frictional contact, and irradiation-induced volumetric swelling models for cladding. A statistical analysis of irradiation-induced volumetric swelling models for HT9, D9, and SS316 was performed for experiments X421/X421A, X441/X441A, and X486. Between these three experiments, there were 174 post-irradiation examination (PIE) profilometries used for validating the swelling models presented using a standard error of the estimate (SEE) method. Implementation of the volumetric swelling models for D9 and SS316 claddings was found to have a significant impact on the BISON profilometry simulated, where HT9 clad pins had an insignificant change due to low fluence values. BISON profilometry simulated for HT9, D9, and SS316 fuel pins agreed with PIE profilometry measurements, with assembly SEE values being 4.4 × 10−3 for X421A, 2.0 × 10−3 for X441A, and 2.8 × 10−3 for X486. D9 clad pins in X421/X421A had the highest SEE values, which is due to the BISON simulated profilometry being shifted axially. While this work accomplished its purpose to demonstrate the modeling of multiple fuel pins from the databases to help validate models, the results suggest that the continued development of metallic fuel models is necessary for qualifying new metallic fuel systems to better capture some physical performance phenomena, such as the hot pressing of U-Pu-Zr and the fuel cladding chemical interaction.
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Budanov, P. F., K. Yu Brovko, Е. А. Khomiak, and О. А. Tymoshenko. "IMPROVEMENT OF FUEL ELEMENT SHELL CONTROL METHODS TO INCREASE NUCLEAR REACTOR SAFETY." Bulletin of the National Technical University "KhPI". Series: Energy: Reliability and Energy Efficiency, no. 1 (1) (December 30, 2020): 26–31. http://dx.doi.org/10.20998/2224-0349.2020.01.04.
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The analysis of the existing methods of control of the surface of the fuel element cladding material was carried out, which showed that their use for detecting surface and internal defects, such as local inhomogeneities, micro- and macropores, various cracks, axial looseness, etc. is characterized by low efficiency, is a laborious process that requires additional surface treatment, material of the fuel elements cladding. In addition, the investigated methods of controlling the surface of the fuel element cladding material make it possible to visually identify only rough external cracks and large slag inclusions, small cracks and non-metallic inclusions invisible under the slag layer. It is proposed to assess the quality of the surface of the shell material in case of its damage and destruction, the use of a computational apparatus based on the method of the theory of fractals. It is proposed to use the fractal properties of the shell material structure and a quantitative fractal value – the fractal dimension, which makes it possible to determine the degree of filling of the volume of the shell material structure during fuel element depressurization. A mathematical model of damage to the structure of the fuel element cladding material is developed depending on the simultaneous effect of high temperature and internal pressure caused by the accumulation of nuclear fuel fission products between the nuclear fuel pellet and the inner surface of the fuel element cladding, taking into account the fractal increases in the geometric parameters of the fuel element cladding. It is shown that damaged structures of the fuel rod cladding material depend on the pressure and temperature inside the fuel rod cladding, as well as the fractal increase in geometric parameters, such as: volume and surface area, outer and inner diameters, height and cross-sectional area, cladding length and height of nuclear pellets, gap between the inner surface of the cladding and nuclear fuel. A criterion for assessing the integrity of the fuel rod cladding is determined, which depends on the change in geometric values in the event of damage and destruction of the structure of the fuel rod cladding material. Practical recommendations are given on the use of the proposed method for monitoring the tightness of the fuel element cladding for processing information obtained from the computational module of the system for monitoring the tightness of the cladding for the automated process control system of the nuclear power plant power unit, which makes it possible to detect the depressurization of fuel elements at an earlier stage in comparison with the standard procedure.
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Alrwashdeh, Mohammad, and Saeed A. Alameri. "Chromium-Coated Zirconium Cladding Neutronics Impact for APR-1400 Reactor Core." Energies 15, no. 21 (October 28, 2022): 8008. http://dx.doi.org/10.3390/en15218008.
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The accident-tolerant fuel concept involves replacing the conventional cladding system (zirconium) with a new material or coating that has specific thermomechanical properties. The aim of this study is to evaluate the neutronics performance of a chromium coating concept and design solutions. A Zircaloy–uranium fuel system (Zr–U) is currently used as a standard fuel system in pressurized water reactors around the world. This investigation presents the benefits of utilizing an alternative cladding material such as chromium coating and the effects on the thermal neutron parameters of the way in which the chromium coating is introduced in the reactor fuel. Among these significant benefits is an increase in the reactor fuel’s thermal conductivity, which improves reactor safety. Two types of fuel-cladding systems were investigated: Zircaloy–uranium (Zr–U) and Zircaloy–chromium (Zr–Cr–U) coating fuel systems. Neutronics analysis evaluations were performed for the selected fuel assemblies and a two-dimensional full core based on an APR-1400 reactor design. Neutronics analyses were performed for the application of the new fuel-cladding material systems using the reactor-physics Monte Carlo code Serpent 2.31.
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Ishibashi, Ryo, Yasunori Hayashi, Huang Bo, Takao Kondo, and Tatsuya Hinoki. "Radiation Effect in Ti-Cr Multilayer-Coated Silicon Carbide under Silicon Ion Irradiation up to 3 dpa." Coatings 12, no. 6 (June 14, 2022): 832. http://dx.doi.org/10.3390/coatings12060832.
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Replacement of conventional Zircaloy fuel cladding with silicon carbide (SiC) fuel cladding is expected to significantly decrease the amount of hydrogen generated from fuel claddings by the reaction with steam during severe accidents. One of their critical issues addressed regarding practical application has been hydrothermal corrosion. Thus, the corrosion resistant coating technology using a Ti-Cr multilayer was developed to suppress silica dissolution from SiC fuel cladding into reactor coolant under normal operation. The effect of radiation on adhesion of the coating to SiC substrate and its microstructure characteristics were investigated following Si ion irradiation at 573 K up to 3 dpa for SiC. Measurement of swelling in pure Ti, pure Cr and SiC revealed that the maximum inner stress attributed to the swelling difference was generated between the coating and SiC substrate by irradiation of 1 dpa. No delamination and cracking were observed in cross-sectional specimens of the coated SiC irradiated up to 3 dpa. According to analyses using transmission electron microscopy, large void formation and cascade mixing due to irradiation were not observed in the coating. The swelling in the coating at 573 K was presumed to be caused by another mechanism during radiation such as point defects rather than void formation.
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Chen, Huan, Xiaoming Wang, and Ruiqian Zhang. "Application and Development Progress of Cr-Based Surface Coatings in Nuclear Fuel Element: I. Selection, Preparation, and Characteristics of Coating Materials." Coatings 10, no. 9 (August 21, 2020): 808. http://dx.doi.org/10.3390/coatings10090808.
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To cope with the shortcomings of nuclear fuel design exposed during the Fukushima Nuclear Accident, researchers around the world have been directing their studies towards accident-tolerant fuel (ATF), which can improve the safety of fuel elements. Among the several ATF cladding concepts, surface coatings comprise the most promising strategy to be specifically applied in engineering applications in a short period. This review presents a comprehensive introduction to the latest progress in the development of Cr-based surface coatings based on zirconium alloys. Part I of the review is a retrospective look at the application status of zirconium alloy cladding, as well as the development of ATF cladding. Following this, the review focuses on the selection process of ATF coating materials, along with the advantages and disadvantages of the current mainstream preparation methods of Cr-based coatings worldwide. Finally, the characteristics of the coatings obtained through each method are summarized according to some conventional performance evaluations or investigations of the claddings. Overall, this review can help assist readers in getting a thorough understanding of the selection principle of ATF coating materials and their preparation processes.
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Moshev, A. A., S. P. Martynenko, S. S. Martynenko, E. M. Kudryavtsev, A. N. Ableyev, A. N. Tokarev, R. A. Panasenko, et al. "Determination of temperature dependences of Young's modulus and internal friction of fuel cladding by resonance method." MATEC Web of Conferences 277 (2019): 03011. http://dx.doi.org/10.1051/matecconf/201927703011.
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We study elastic characteristics and internal friction of fuel claddings to improve computer codes for VVER-1000 fuel rods. We analytically described elastic characteristics of cladding material and obtained coefficient of the form of the first longitudinal frequency numerically. We described new measuring module for automatic acquisition data. We’ve established temperature dependences of Young’s modulus and internal friction via high-temperature facility and developed electronic module and noted maximum of these characteristics at the temperature 1160 K. It can be explained by the destruction of the texture in the material of claddings.
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Hallstadius, Lars, Steven Johnson, and Ed Lahoda. "Cladding for high performance fuel." Progress in Nuclear Energy 57 (May 2012): 71–76. http://dx.doi.org/10.1016/j.pnucene.2011.10.008.
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Konovalov, Igor I., Boris A. Tarasov, and Eduard M. Glagovskiy. "Irradiation Creep of Uranium-Plutonium Nitride Fuel and Serviceability of Fuel Element." Defect and Diffusion Forum 375 (May 2017): 91–100. http://dx.doi.org/10.4028/www.scientific.net/ddf.375.91.
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Article discusses experimental data on creep of (U,Pu)N and other uranium compounds, and possible mechanism of mass-transfer. Proposed equation describes the following creep features: weak temperature dependence at T < 1000°C, creep acceleration in a fuel with micron-sized grains, and acceleration with the content of second phases formed by impurities and fission products. The difference in creep behavior in reactors with thermal and fast neutrons environmentsis discussed. Comparison of irradiation creep of nitride fuel and properties of cladding materials shows that under parameters of fast reactors and typical design of fuel element it is impossible to implement restraining of external nitride swelling. As initial porosity in the fuel will not compensate the nitride swelling, the cladding of fuel element will work in a mode of following the changing of fuel size. Some suggestions on the cladding material properties are done.
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Dreganov, Oleg I., Vitalij N. Shulimov, Irina V. Kiselyova, and Aleksandr V. Alekseev. "Measurement of the spent fuel rod cladding temperature during the in-pile testing at 500 – 900°C." Nuclear Energy and Technology 4, no. 1 (October 17, 2018): 21–26. http://dx.doi.org/10.3897/nucet.4.29838.
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This paper deals with the problem of measuring the VVER-1000 burnup fuel cladding temperature in a 500–900°C range in the process of experiments in a channel of the MIR research reactor to obtain data on the fuel element behavior under the influence of the parameters typical of the maximum design-basis loss-of-coolant accident (LOCA). Studying the burnup fuel cladding deformation pattern requires measurements of the cladding temperature with no (thermal, mechanical and other) impacts on the cladding in the maximum deformation region. For dynamic experiments in the MIR reactor channel with fuel testing in a vapor-gas environment, a cladding temperature measuring unit has been developed, in which the cladding is not subjected to external impacts in the maximum deformation region. In the process of being installed into the spacer grid, the thermoelectric transducer (TET) has its hot junction forced against the cladding making it possible to prevent the external impact on the cladding. The thermometric characteristic of the TET attachment, which is associated with the impact of the grid as such on its thermal condition, was studied using a laboratory facility. This technique was used in an in-pile experiment to study the fuel cladding deformation pattern.
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Sihotang, Juan Carlos, Maman Kartaman Ajiriyanto, Ely Nurlaily, Junaedi Junaedi, and Aslina Br Ginting. "OXIDE LAYER CHARACTERIZATION OF AlMg2 CLADDING OF IRRADIATED U3Si2/Al FUEL WITH 4,8 gU/cm3 DENSITY." Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir 28, no. 3 (October 31, 2022): 125. http://dx.doi.org/10.17146/urania.2022.28.3.6755.
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OXIDE LAYER CHARACTERIZATION OF AlMg2 CLADDING OF IRRADIATED U3Si2/Al FUEL WITH 4,8 gU/cm3 DENSITY. To investigate the performance of AlMg2 cladding in the U3Si2/Al dispersion fuel, oxide layer characterization of AlMg2 cladding of the irradiated U3Si2/Al fuel with 4.8 gU/cm3 density was conducted. The oxide layer on the surface of AlMg2 cladding is one of the changes that occur on the cladding after the U3Si2/Al fuel plate has been irradiated in the RSG-GAS reactor to a burn-up of ∼40%. The characterization and observation of the oxide layer was conducted using SEM (Scanning Electron Microscope) and Energy-dispersive X-ray spectroscopy (EDS). Samples with a size of 3x3 mm were taken from the middle of the fuel plate (middle position). After cutting, metallographic preparation includes mounting, grinding, polishing, and ultrasonic cleaning. SEM preparation was carried out by sputter coating using Au layer. The oxide layer on the AlMg2 cladding has a thickness of 10.3 µm with a uniformly distributed cracks along the oxide layer.Keyword: LEU, uranium-silicide, post-irradiation examination, AlMg2 cladding, oxide layer.
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Alaleeli, Maithah, Saeed Alameri, and Mohammad Alrwashdeh. "Neutronic Analysis of SiC/SiC Sandwich Cladding Design in APR-1400 under Normal Operation Conditions." Energies 15, no. 14 (July 18, 2022): 5204. http://dx.doi.org/10.3390/en15145204.
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Our aim is to study the neutronic behaviour of potential accident-tolerant fuel (ATF) claddings in a pressurised water reactor under normal operations. This work compares ATF silicon carbide composite (SiC/SiC) cladding to conventional ZIRLOTM cladding in APR-1400. Additionally, a “sandwich” cladding design developed by the CEA is used for SiC/SiC. The design structure includes a liner in between two layers of the composite to ensure leak tightness. The two proposed liners are Niobium (Nb) and Tantalum (Ta). Serpent 2, a Monte Carlo reactor physics lattice code, is employed to model both cladding materials in APR-1400 at three different levels: pin cell, fuel assembly, and core. The criticality, neutron spectrum, actinide inventory, and power distribution as a function of burnup are investigated. The simulations show that SiC/SiC with the Nb liner displays a far superior performance than the Ta liner across all examined characteristics. Ta leads to a harder neutron spectrum and increased Pu-239 content throughout the cycle, while Nb presents negligible effects. In fact, SiC/SiC with the Nb liner performs very similarly to ZIRLOTM at all model levels. The results indicate that, in terms of neutronics, the adoption of the SiC/SiC composite would entail little to no changes to current APR-1400 operations.
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Orlova, E. "THE SYNERGETIC OF FUEL ELEMENT." PROBLEMS OF ATOMIC SCIENCE AND TECHNOLOGY. SERIES: NUCLEAR AND REACTOR CONSTANTS 2019, no. 3 (September 26, 2019): 261–72. http://dx.doi.org/10.55176/2414-1038-2019-3-261-272.
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The fuel element is the complex system consisting of fuel, a sublayer, a cladding, protective coatings on a surface of a cladding and contacting to the cooler and material of a contour in general. The synergetic (self-coherence) of interaction of elements of this system has to be provided both concerning high heat conductivity of fuel element, and concerning corrosion compatibility. Use of a liquid metal sublayer (LMS) instead of gas, allows significantly (by hundreds of degrees) to reduce temperature in the center of fuel that increases safety at UTOP accidents (uncontrollable increase in power) at ULOF (loss of an expense of the cooler). Gap thickness when using LMS (unlike helium) can be significantly increased that will practically not affect thermal characteristics of fuel element, but will allow to distance considerably time of approach of direct contact of fuel with a cladding, having increased thereby depth of burning out of heavy atoms and having increased cost efficiency and competitiveness of fast reactors extension of the fuel elements resource. These principles when using heat-conducting nuclear fuel are especially effective (metal, nitride, carbide) and the lead heat carrier. Synergetic reasonable corrosion compatibility of a cladding of fuel element with LMS is confirmed with numerous settlement pilot studies by means of formation and self-curing of accidental damages of a protective coating of nitride of zirconium on the internal surface of steel in LMS of eutectic structure on the basis of lead with magnesium and zirconium. When using nitride fuel heat-conducting LMS with anticorrosive properties the resource of fuel element is limited not by swelling of fuel any more, and the dose damaging a cladding.
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Caha, Vojtěch, and Jakub Krejčí. "POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION." Acta Polytechnica CTU Proceedings 4 (December 16, 2016): 8. http://dx.doi.org/10.14311/ap.2016.4.0008.
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The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF) is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature). The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.
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Khumsa-Ang, Kittima, Alberto Mendoza, Armando Nava-Dominguez, Chukwudi Azih, and Hussam Zahlan. "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." Coatings 13, no. 9 (September 20, 2023): 1648. http://dx.doi.org/10.3390/coatings13091648.
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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.
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Yakushkin, A. A., V. M. Borisov, and V. N. Trofimov. "Properties of chrome coatings applied by various methods to zirconium alloy E110." Physics and Chemistry of Materials Treatment 2 (2021): 42–50. http://dx.doi.org/10.30791/0015-3214-2021-2-42-50.
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The issue presents the results of studies of methods for increasing the corrosion resistance of fuel claddings made from zirconium alloy E110, performed at the SRC RF TRINITI, using coatings (Al, Al2O3, Cr) by pulsed laser deposition, as well as magnetron sputtering and galvanic deposition of chromium. Also presented some results of studies of adhesion, microstructure and corrosion resistance of chromium coatings deposited by various methods. It was revealed that the corrosion resistance of the cladding of fuel rods with chrome coatings at temperatures of 1100 – 1200 °С practically does not depend on the method of their application. However, the coating method has a significant effect on their adhesion, with coatings obtained by atomic deposition methods and characterized by a continuous uniform structure have the greatest adhesion. The depth of oxidation of the outer surface of the cladding of a fuel rod when applying a chromium coating decreases by an average of 30 times for sponge-based alloy E110.
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Nishat, Sadek Hossain, Farhana Islam Farha, and Md Hossain Sahadath. "Study of the Perturbation in Temperature Profile of an AGR Fuel Pin for Surface Roughness of Cladding by CFD Simulation in Ansys Fluent." Dhaka University Journal of Applied Science and Engineering 7, no. 1 (February 1, 2023): 9–15. http://dx.doi.org/10.3329/dujase.v7i1.62881.
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The surface roughness of nuclear fuel cladding plays a crucial role in the thermal-hydraulic response of the Advanced Gas Cooled reactor (AGR). In the present work, the change in the temperature distribution from an isolated AGR fuel rod to primary coolant due to cladding roughness was studied by computational fluid dynamics (CFD) simulation in Ansys Fluent software. Square transverse ribs of the various pitch to height ratios _p / k_ were considered as the surface roughness. Radial temperature profiles from fuel to coolant were generated. Lower fuel temperature was found for the fuel rod with a rough cladding surface as compared to the smooth cladding surface. The peak fuel temperature was determined and found to decrease with decreasing values of _p / k_ . Temperature drop across the fuel and from fuel to coolant was also studied. DUJASE Vol. 7(1) 9-15, 2022 (January)
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Kulakov, G. V., Y. V. Konovalov, A. A. Kosaurov, M. M. Peregud, V. Y. Shishin, and A. A. Sheldyakov. "Post-irradiation examinations of dispersion fuel rods with modified zirconium alloys claddings." Voprosy Materialovedeniya, no. 3(95) (January 10, 2019): 206–12. http://dx.doi.org/10.22349/1994-6716-2018-95-3-206-212.
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Modified zirconium alloys E635Mand E635opt based on E635 alloy (E635 was selected as master alloy) have been developed at Bochvar Institute. Fuel rods with such claddings were manufactured at Bochvar Institute and were irradiated at MIR reactor (SC RIAR, Dimitrovgrad). The results from the PIE performed at RIAR are presented. Such features of claddings as microstructures, corrosion resistance (width and structure of oxide), hydrogen contents, distribution of hydrides, mechanical properties were examined and discussed. Modifications of the alloy E635opt and E635M showed higher resistance to corrosion and hydrogen pick-up compared to the E635 alloy, while maintaining high strength and ductility. They have confirmed their prospects for use as cladding for fuel rods with enhanced characteristics.
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Sagiroun, Mamoun I. A., Xin Rong Cao, Wasim M. K. Helal, and John N. Njoroge. "A Review of Development of Zirconium Alloys as a Fuel Cladding Material and its Oxidation Behavior at High-Temperature Steam." International Journal of Engineering Research in Africa 46 (January 2020): 7–14. http://dx.doi.org/10.4028/www.scientific.net/jera.46.7.
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Currently, Zr-alloys are widely used in nuclear power reactors for fuel cladding and structural components. Many types of zr-based alloys were developed to overcome the challenges encountered in the progress of nuclear reactors (high-burnup and high-duty). Oxygen diffused into the cladding, hydrogen absorbed in the cladding (breakaway oxidation and ruptured balloons) and rapid oxidation rate are results of chemical interaction of cladding material with steam at high temperature. Zirconium alloys seem to be the most suitable for use in fuel cladding, if they can overcome the rapid oxidation at temperature higher than 1200 °C. Previous studies on the oxidation behavior for some Zr-alloys nuclear fuel cladding tubes in steam and steam–air atmospheres at high temperatures are reviewed. The oxidation behavior of zirconium-alloys is strongly affected by the chemical composition of alloys and its surface conditions.
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Nagase, Fumihisa, Kan Sakamoto, and Shinichiro Yamashita. "Performance degradation of candidate accident-tolerant cladding under corrosive environment." Corrosion Reviews 35, no. 3 (August 28, 2017): 129–40. http://dx.doi.org/10.1515/corrrev-2017-0014.
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AbstractLight-water reactor (LWR) fuel cladding shall retain the performance as the barrier for nuclear fuel materials and fission products in high-pressure and high-temperature coolant under irradiation conditions for long periods. The cladding also has to withstand temperature increase and severe loading under accidental conditions. As lessons learned from the accident at the Fukushima Daiichi nuclear power station, advanced cladding materials are being developed to enhance accident tolerance compared to conventional zirconium alloys. The present paper reviews the progress of the development and summarizes the subjects to be solved for enhanced accident-tolerant fuel cladding, focusing on performance degradation under various corrosive environmental conditions that should be considered in designing the LWR fuel.
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Krejčí, Jakub, Jitka Kabátová, Jan Kočí, Zuzana Weishauptová, and Věra Vrtílková. "HYDROGEN CHARGING OF FUEL CLADDING METHODOLOGY." Acta Polytechnica CTU Proceedings 4 (December 16, 2016): 38. http://dx.doi.org/10.14311/ap.2016.4.0038.
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Hydrogen content is a very important parameter for mechanical properties of fuel cladding, especially after LOCA transients. Therefore, it is necessary to take into account the amount of hydrogen absorbed in the fuel cladding during normal operation (before a hypothetical LOCA). The required value of hydrogen content is possible to reach by a long-term pre-oxidation test or a much shorter hydrogen charging experiment. The methodology of hydrogen charging developed in UJP is described in this contribution. Results of experiments aiming to prepare samples with uniform hydrides and samples with a rim-layer and other hydrides are shown.
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Červenka, Petr, Jakub Krejčí, Ladislav Cvrček, Vojtěch Rozkošný, František Manoch, David Rada, and Jitka Kabátová. "EXPERIMENTAL STUDY OF DAMAGED CR-COATED FUEL CLADDING IN POST-ACCIDENT CONDITIONS." Acta Polytechnica CTU Proceedings 28 (December 1, 2020): 1–7. http://dx.doi.org/10.14311/app.2020.28.0001.
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To enhance the safety of nuclear power, the focus of researchers all around the world has recently mainly objected on the development of Accident Tolerant Fuels. Especially the Chromium coating of current Zirconium based cladding has been widely suggested and discussed for its immense positive effect on overall cladding properties. Nevertheless, it was observed that during the first stage of the Loss of Coolant Accident, cracks appear in the Cr coating due to its inability to tolerate higher plastic strain. Therefore, experimental methodology used in this article focuses on testing fuel cladding with damaged Cr coating after the high-temperature transient. The impact of cracks on degradation of cladding mechanical properties was observed using optical microscopy, ring compression test, microhardness, and evaluating hydrogen content and weight gain.
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König, Tobias, Ron Dagan, Kathy Dardenne, Michel Herm, Volker Metz, Tim Pruessmann, Jörg Rothe, Dieter Schild, Arndt Walschburger, and Horst Geckeis. "Spectroscopic and chemical investigations on volatile fission and activation products within the fuel-cladding interface of irradiated pressurised water reactor fuel rod segments." Safety of Nuclear Waste Disposal 1 (November 10, 2021): 5–6. http://dx.doi.org/10.5194/sand-1-5-2021.
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Abstract. In Germany, the present waste management concept foresees the direct disposal of spent nuclear fuel (SNF) in deep geological repositories for high-level waste available by 2050, at best. Until then, SNF is encapsulated in dual-purpose casks and stored in dry interim storage facilities. Licenses for both casks and facilities will expire after 40 years following loading of the cask and emplacement of the first cask in the storage location. Yet, due to considerable delays in the site selection process and the estimated duration for construction and commissioning of a final repository of at least 2 decades, a prolonged dry interim storage of SNF is inevitable (ESK, 2015). Concerning these considerable timespans, integrity of the cladding is of utmost importance regarding the ultimately conditioning of the fuel assemblies for final disposal. Various processes strain the structural integrity of Zircaloy cladding during reactor operation and beyond such as delayed hydride cracking, fuel-cladding chemical interactions or irradiation damage induced by α-emitters present in the fuel pellet's rim zone (Ewing, 2015). Especially with higher burn-up, the gap between fuel and cladding closes and results in the formation of an interaction layer, in which precipitates of fission and activation products are present, displaying an interface for degradation processes. For chemical analysis and speciation of these agglomerates, Zircaloy-4 and SNF specimens were sampled from fuel rod segments irradiated in commercial pressurised water reactors during the 1980s. Zircaloy-4 specimens were taken from an UOX (50.4 GWdtHM-1) and mixed oxide fuel (MOX) (38.0 GWdtHM-1). In addition, SNF fragments were sampled from the closed gap of both fuel types to examine volatile activation and fission products, which had been segregated from the centre to the pellet periphery during irradiation and thus contribute to the possible chemically assisted cladding degradation effect of the precipitates within the fuel-cladding interface. Spectroscopic analysis of precipitates within the interface layer between fuel and cladding were performed by optical microscopy, X-ray absorption and X-ray photoelectron spectroscopy, as well as by energy-dispersive scanning electron microscopy. Moreover, the radionuclide inventory of the respective Zircaloy-4, fuel and interaction layers was determined using liquid scintillation counting, γ-spectroscopy, gas mass spectrometry, ion chromatography and inductive-coupled plasma mass spectrometry and compared to results received by MCNP/CINDER and webKORIGEN calculations. In this study, we provide results regarding the speciation and chemical composition of previously identified Cs-U-O-Zr-Cl-I bearing compounds found in the interaction layer of irradiated nuclear fuel and inventory analyses of radionuclides present therein, with particular emphasis on Cl-36 and I-129. Furthermore, the agglomerates within the fuel-cladding interface were characterised for the first time utilising synchrotron radiation-based Cl K-edge and I K-edge measurements, resulting in compounds with structural similarities to CsCl and CsI. The outcomes obtained from this study provide further insights into the complex chemistry within the fuel-cladding interface with respect to the aging management and integrity of SNF under the conditions of interim storage. In future studies we will examine whether the different compounds at the fuel-cladding interface have the potential to affect the mechanical properties of Zircaloy cladding.
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Pham, Hai V., Masaki Kurata, and Martin Steinbrueck. "Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview." Thermo 1, no. 2 (July 27, 2021): 151–67. http://dx.doi.org/10.3390/thermo1020011.
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Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.
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Isayev, Rafael Sh, Pavel S. Dzhumaev, Irina A. Naumenko, and Maria V. Leontieva-Smirnova. "Corrosion resistance of chromium coating on the inner surface of EP823-Sh steel cladding." Nuclear Energy and Technology 10, no. 2 (May 7, 2024): 81–88. http://dx.doi.org/10.3897/nucet.10.119642.
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The processes of corrosion damage of the inner surface of the cladding are determined by corrosive reagents aggressive with respect to the cladding and the type of fuel used. Reactor irradiation of cladding made of EP823-Sh steel with mixed nitride fuel planned for use in the BREST-OD-300 reactor revealed non-uniform corrosion of the inner surface of the cladding. In this paper, the use of the chromium coating is proposed to prevent the corrosion of the inner surface of the steel fuel cladding. The results of corrosion tests of chromium coating applied to the inner surface of cladding made of EP823-Sh steel by electrolytic deposition are presented. Electron-microscopic studies of the chromium coating on EP823-Sh steel showed no significant signs of corrosion damage when tested in the environment of simulant fission products (CsI+Te) and in liquid lead at 650 °C.
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Karpyuk, L. A., V. I. Kuznetsov, A. A. Maslov, V. V. Novikov, V. K. Orlov, D. V. Rykunov, and A. O. Titov. "Accident Tolerant Fuel with Chromium-Coated Fuel-Rod Cladding." Atomic Energy 130, no. 3 (July 2021): 149–55. http://dx.doi.org/10.1007/s10512-021-00786-9.
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Okui, Tsutomu, and Akifumi Yamaji. "PRELIMINARY POWER TRANSIENT ANALYSIS OF THE SUPER FR WITH AXIALLY HETEROGENEOUS CORE." EPJ Web of Conferences 247 (2021): 07002. http://dx.doi.org/10.1051/epjconf/202124707002.
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The Super FR is one of the SuperCritical Water cooled Reactor (SCWR) concepts with once-through direct cycle plant system. Recently, new design concept of axially heterogeneous core has been proposed, which consists of multiple layers of MOX and blanket fuels. To clarify the safety performance during power transient, safety analyses have been conducted for uncontrolled control rod (CR) withdrawal and CR ejection at full power. RELAP/SCDAPSIM code was used for the safety analysis. The results show that the peak cladding surface temperature (PCST) is high in the upper MOX fuel layer. It is also shown that axial temperature gradient of cladding greatly increases in a short period. Suppressing such large temperature gradient may be a design issue for the axially heterogeneous core from the viewpoint of ensuring fuel integrity.
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Winter, Thomas C., Richard W. Neu, Preet M. Singh, Lynne E. Kolaya, and Chaitanya S. Deo. "Fretting wear comparison of cladding materials for reactor fuel cladding application." Journal of Nuclear Materials 508 (September 2018): 505–15. http://dx.doi.org/10.1016/j.jnucmat.2018.05.069.
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Konashi, Kenji, Katsuichiro Kamimura, and Yoji Yokouchi. "Model for Fuel/Cladding Chemical Interaction." Nuclear Technology 72, no. 3 (March 1986): 328–37. http://dx.doi.org/10.13182/nt86-a33771.
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Anufriev, B. F., V. M. Baranov, Yu K. Bibilashvili, I. S. Golovnin, M. G. Isaenkova, A. Yu Kalyadin, Yu A. Perlovich, and V. Yu Tonkov. "Homogeneity in zirconium fuel-pin cladding." Soviet Atomic Energy 64, no. 3 (March 1988): 245–48. http://dx.doi.org/10.1007/bf01123133.
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Dyk, Štěpán, and Vladimír Zeman. "Bifurcations in Mathematical Model of Nonlinear Vibration of the Nuclear Fuel Rod." Applied Mechanics and Materials 821 (January 2016): 207–12. http://dx.doi.org/10.4028/www.scientific.net/amm.821.207.
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The paper deals with nonlinear phenomena that occurs during vibration of nuclear fuel rod (FR). The FR is considered as a system consisting of two impact-interacting subsystems FR cladding (zircalloy tube) and fuel pellets stack placed inside FR cladding. Between both subsystems, there is a small radial clearance. The FR is bottom-end-fixed, and at eight equidistant levels, the FR cladding is supported by spacer grids (SG). Both subsystems are modelled by means of finite element method for one-dimensional Euler-Bernoulli continua. During fuel assembly (FA) motion caused by pressure pulsations of the coolant, the FR vibrates and impacts can possibly occur between FR cladding and fuel pellets stack. The paper focuses on qualitative change of vibration with change of bifurcation parameters clearance between FR cladding and fuel pellets stack and stiffness of spacer grids cells. The change of vibration quality is shown by extremes of relative radial displacements of both continua in discretization nodes and by phase trajectories. Dependence of impact motion on modal properties of both subsystems is shown.
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Vorobyov, Yu, O. Zhabin, and M. Frankova. "Application of RELAP5/MOD3.2 Cladding Deformation Model for VVER-1000 Fuel in Design-Basis Accident Analysis." Nuclear and Radiation Safety, no. 3(71) (August 15, 2016): 19–22. http://dx.doi.org/10.32918/nrs.2016.3(71).04.
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The paper presents applicability of built-in RELAP5/MOD3.2 cladding deformation model for VVER-1000 fuel with cladding of Zr+1 % Nb alloy. Experimental data and simplified model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the hot channel blockage after cladding swelling and rupture in the interval of temperatures from 600 to 1200°С and interval of pressures from 1 to 12 MPa. It is demonstrated that RELAP5/MOD3.2 builtin model of cladding deformation can be applied to VVER-1000 cladding of Zr+1%Nb alloy rupture estimation only in the certain limited range of parameters. The analysis of RELAP5/MOD3.2 cladding deformation model parameters influence on the peak cladding temperature in double-ended cold leg break was performed. The paper presents recommendations on the use of RELAP5/MOD3.2 built-in cladding deformation model in the design basis accident analysis of VVER-1000 reactors.
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Lys, Stepan. "Analysis of Calculation Model for Primary Coolant Fission Products." Energy engineering and control systems 9, no. 2 (2023): 69–74. http://dx.doi.org/10.23939/jeecs2023.02.069.
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The sources of radioactive contamination of the primary coolant by fission products when the unit is operating at the rated power are as follows: defect fuel elements with gas leakiness and substantial damages, surface contamination of the outer surfaces of fuel claddings, superficial contamination of structural materials of fuel assemblies. Initially in the reactor operation (if there are no manufacturing defects in fuel elements), the contamination of the coolant by fission products is determined by the release into the reactor coolant circuit of fission fragments of Uranium-235 (due to their kinetic energy) that is present on the outer surfaces of fuel elements as contamination in their manufacturing. During normal operation of the reactor, the integrity of cladding may fail due to various processes of corrosion fatigue type. These processes result in, first of all, micro-fissures and then in large defects in the claddings, which is accompanied by an increase in the release of fission products from fuel elements into the primary coolant.
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Smolík, Vojtěch, Alžběta Endrychová, and Jakub Krejčí. "Simulation of the compression test of the Zr1Nb fuel cladding ring." Acta Polytechnica CTU Proceedings 44 (December 1, 2023): 28–32. http://dx.doi.org/10.14311/app.2023.44.0028.
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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.
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Dinh, Van Chien. "Verification of TVS-2006 fuel rod design of VVER-AES2006 reactor under steady-state operating condition Using FRAPCON-3.5 code." Nuclear Science and Technology 5, no. 2 (June 30, 2015): 47–58. http://dx.doi.org/10.53747/jnst.v5i2.193.
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The purpose of this paper is to discuss the independent verification of TVS-2006 fuel rod design used in VVER-AES2006 reactor (Novovoronezh NPP-2 Power, Unit 1), based on the acceptance criteria and the reference data given in the Preliminary Safety Analysis Report of the State Research, Design, Construction and Survey Institute “Atomenergoproekt” (PSAR) and the operation of VVER-1000 reactor. The calculations were performed using FRAPCON-3.5 code, including fuel temperature, cladding temperature, fission gas release, internal gas pressure, cladding stress and strain, fuel extension, fuel rod elongation, cladding creep rate, fuel swelling rate, cladding oxide thickness and hydrogen concentration. The results are compared with the calculated data using START-3 code in PSAR and the acceptance criteria required by Russian nuclear regulatory body. Despite some discrepancies, the results showed conformance with the calculated data given in the PSAR and meet the acceptance criteria.