Relevant bibliographies by topics / Fuel cladding

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Academic literature on the topic 'Fuel cladding'

Author: Grafiati
Published: 4 June 2021
Last updated: 7 July 2024

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Journal articles on the topic "Fuel cladding"

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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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Dissertations / Theses on the topic "Fuel cladding"

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Paramonova, Ekaterina (Ekaterina D. ). "CRUD resistant fuel cladding materials." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/82447.

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Thesis (S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.
"June 2013." Cataloged from PDF version of thesis.
Includes bibliographical references (pages 27-29).
CRUD is a term commonly used to describe deposited corrosion products that form on the surface of fuel cladding rods during the operation of Pressurized Water Reactors (PWR). CRUD has deleterious effects on reactor operation and currently, there is no effective way to mitigate its formation. The Electric Power Research Institute (EPRI) CRUD Resistant Fuel Cladding project has the objective to study the effect of different surface modifications of Zircaloy cladding on the formation of CRUD, and ultimately minimize its effects. This modification will alter the surface chemistry and therefore the CRUD formation rate. The objective of this study was to construct a pool boiling facility at atmospheric pressure and sub-cooled boiling conditions, and test a series of samples in simulated PWR water with a high concentration of nanoparticulate CRUD precursors. After testing, ZrC was the only material out of six that did not develop dark, circular spots, which are hypothesized to be the beginnings of CRUD boiling chimneys. Further testing will be needed to confirm that it is indeed more CRUD resistant, even under realistic PWR conditions in a parallel testing facility.
by Ekaterina Paramonova.
S.B.
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2

Reece, Warren Daniel. "Theory of cladding breach location and size determination using delayed neutron signals /." Diss., Georgia Institute of Technology, 1988. http://hdl.handle.net/1853/13317.

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Andrews, Michael Robert. "The interaction of deposition promoters with AGR fuel cladding surfaces." Thesis, University of Newcastle Upon Tyne, 1998. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.244466.

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Jarvis, Jennifer Anne. "Hydrogen entry in Zircaloy-4 fuel cladding : an electrochemical study." Thesis, Massachusetts Institute of Technology, 2015. http://hdl.handle.net/1721.1/103730.

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Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2015.
Cataloged from PDF version of thesis.
Includes bibliographical references (pages 291-297).
Corrosion and hydrogen pickup of zirconium alloy fuel cladding in water cooled nuclear reactors are life-limiting phenomena for fuel. This thesis studies the fate of hydrogen liberated by waterside corrosion of Zircaloy-4 fuel cladding in Pressurized Water Reactors (PWRs): are the adsorbed protons incorporated into the oxide and eventually the metal, or are they evolved into molecular hydrogen and released into the coolant? Water chemistry modeling was used to understand effects of radiolysis and CRUD. Density functional theory (DFT) was used to investigate the role of oxidized Zr(Fe,Cr)2 second phase particles. Chemical potentials and the electron chemical potential were used to connect these two modeling efforts. A radiolysis model was developed for the primary loop of a PWR. Dose profiles accounting for fuel burnup, boron addition, axial power profiles, and a CRUD layer were produced. Dose rates to the bulk coolant increased by 21-22% with 12.5-75 pim thick CRUD layers. Radially-averaged core chemistry was compared to single-channel chemistry at individual fuel rods. Calculations showed that local chemistry was more oxidizing at high-power fuel and fuel with CRUD. Local hydrogen peroxide concentrations were up to 2.5 ppb higher than average levels of 5-8 ppb. Radiolysis results were used to compute chemical potentials and the corrosion potential. Marcus theory was applied to compare the band energies of oxides associated with Zircaloy-4 and the energy levels for proton reduction in PWR conditions. Hydrogen interactions with Cr203 and Fe203, both found in oxidized precipitates, were studied with DFT. Atomic adsorption of hydrogen was modeled on the Cr and Feterminated (0001) surfaces. Climbing Image-Nudged Elastic Band calculations were used to model the competing pathways of hydrogen migration into the subsurface and molecular hydrogen formation. A two-step mechanism for hydrogen recombination was identified consisting of: reduction of an adsorbed proton (H+) to a hydride ion (H-) and H2 formation from an adjacent adsorbed proton and hydride ion. Overall, results suggest that neither surface will be an easy entrance point for hydrogen ingress and that Cr203 is more likely to be involved in hydrogen evolution than the Fe203.
by Jennifer Anne Jarvis.
Ph. D.
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Lee, Youho. "Safety of light water reactor fuel with silicon carbide cladding." Thesis, Massachusetts Institute of Technology, 2013. http://hdl.handle.net/1721.1/86866.

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Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2013.
Cataloged from PDF version of thesis.
Includes bibliographical references (pages 303-314).
Structural aspects of the performance of light water reactor (LWR) fuel rod with triplex silicon carbide (SiC) cladding - an emerging option to replace the zirconium alloy cladding - are assessed. Its behavior under accident conditions is examined with an integrated approach of experiments, modeling, and simulation. High temperature (1100°C~1500°C) steam oxidation experiments demonstrated that the oxidation of monolithic SiC is about three orders of magnitude slower than that of zirconium alloys, and with a weaker impact on mechanical strength. This, along with the presence of the environmental barrier coating around the load carrying intermediate layer of SiC fiber composite, diminishes the importance of oxidation for cladding failure mechanisms. Thermal shock experiments showed strength retention for both [alpha]-SiC and [beta]-SiC, as well as A1₂O₃ samples quenched from temperatures up to 1260°C in saturated water. The initial heat transfer upon the solid - fluid contact in the quenching transient is found to be a controlling factor in the potential for brittle fracture. This implies that SiC would not fail by thermal shock induced fracture during the reflood phase of a loss of coolant accident, which includes fuel-cladding quenching by emergency coolant at saturation conditions. A thermo-mechanical model for stress distribution and Weibull statistical fracture of laminated SiC cladding during normal and accident conditions is developed. It is coupled to fuel rod performance code FRAPCON-3.4 (modified here for SiC) and RELAP-5 (to determine coolant conditions). It is concluded that a PWR fuel rod with SiC cladding can extend the fuel residence time in the core, while keeping the internal pressure level within the safety assurance limit during steady-state and loss of coolant accidents. Peak burnup of 93 MWD/kgU (10% central void in fuel pellets) at 74 months of in-core residence time is found achievable with conventional PWR fuel rod design, but with an extended plenum length (70 cm). An easier to manufacture, 30% larger SiC cladding thickness requires an improved thermal conductivity of the composite layer to reduce thermal stress levels under steady-state operation to avoid failure at the same burnup. A larger Weibull modulus of the SiC cladding improves chances of avoiding brittle failure.
by Youho Lee.
Ph. D.
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Paul, James. "Joining of silicon carbide for accident tolerant PWR fuel cladding." Thesis, University of Manchester, 2017. https://www.research.manchester.ac.uk/portal/en/theses/joining-of-silicon-carbide-for-accident-tolerant-pwr-fuel-cladding(f9851a0a-ef68-465e-8029-a31ab77fab27).html.

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Following two previous nuclear reactor accidents involving light water reactors, there is a renewed interest in accident tolerant fuels. These accident tolerant fuels should not oxidise in a steam atmosphere during loss of coolant accidents. One such accident tolerant fuel that has been suggested is the use of ceramic composite cladding material as a replacement for the current zircaloy cladding. The high temperature stability of silicon carbide, together with its high resistance to corrosion may make it preferable to zircaloy during accident conditions. Furthermore, if the neutron absorption cross section of the cladding is less than the current zircaloy, extended life might be available when compared with current fuels. One of the main difficulties in using ceramic cladding materials as nuclear fuels is the lack of a reliable joining process to manufacture end caps for the cladding tubes. A manufacturing method would need to be developed to produce ceramic joint that is able to withstand a PWR environment. Two methods of ceramic joining have been proposed. Firstly, silicon carbide deposition process that is used to infill the gap between two ceramic components and secondly a ceramic soldering technique. A silicon carbide deposition process has been developed. The deposit was confirmed to be 3C silicon carbide which has preferable irradiation response to the other polytypes. The deposit was found to be carbon rich which was largely removed through the use of a thermal treatment step. The deposit was used to coat metallic surfaces for increased hardness, reduced sliding wear and corrosion resistance. Silicon carbide joints were produced using an oxide powder frit of silicon dioxide, yttrium oxide and aluminium oxide. Tubular samples were joined, however they contained circumferential cracking resulting in a join that was not hermetically sealed. The thermal conductivity of each joint varied from sample to sample. X-ray computed tomography showed there were large inconsistencies in the volume of joined material present in each sample giving rise to the large variation in thermal conductivity.
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Phuah, Chin Heng. "Corrosion of thermally-aged Advanced Gas-Cooled Reactor fuel cladding." Thesis, Imperial College London, 2012. http://hdl.handle.net/10044/1/10550.

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The microstructure of Advanced Gas-Cooled Reactor (AGR) fuel cladding that underpins its corrosion behaviour has been established, contributing to an understanding of long-term control, monitoring practice and storage decisions for this fuel. AGR fuel cladding specimens sourced from Sellafield Ltd, cut and individually heat treated at temperatures from 400 to 800°C for 24 to 192h were attempts to approximate irradiated AGR fuel cladding and characterised both in terms of their corrosion behaviour and of microstructure. Niobium carbide (NbC) second phases are the primary local corrosion sites. Bulk austenite-γ cladding metal (50.3±1.7 at% Fe, 21.0±1.1 at% Cr and 21.0±0.4 at% Ni) around NbC precipitates exhibited extensive corrosion even though the precipitate themselves appear unchanged. Corrosion observed from the specimen surface took the form of lacy covers around an NbC precipitate at the cover centre (~10 to 25 μm dia. depending on the site) and in the subsurface were voids (~0.1 μm pin-holes), cavities (~2 to 5 μm), an envelope of dissolved-metal region along NbC peripheries (~1 μm thick with austenite-γ composition decreased by on average 20% Fe, 21% Cr and 17% Ni) or a large, smooth concave pit bottom comparable to the cover dimension. These observations collectively suggest that AGR cladding corrosion is a diffusion-controlled phenomenon where the NbC precipitate may act as the cathode in a local galvanic couple and the adjacent austenite-γ metal is the anode that undergoes preferential oxidation. The primary contributing factors to NbC-induced AGR cladding corrosion are high NaCl concentration of the electrolyte solution, large NbC precipitates, small austenite-γ grains and presence of stress in the microstructure. Specifically, corrosion potential measurements in the 0.001M electrolyte NaCl are ~800mV (v.s. Ag|AgCl reference electrode) more noble than in the 0.1M electrolyte, suggesting that cladding wet storage requires maintenance with lowest chloride concentration practically achievable. Specimens with comparatively large NbC precipitates (~5 μm) and small austenite-γ grains (~10 μm) that result from heat treatment are ~810mV more corrosion susceptible than the as-received specimens with ~0.1 dia. NbC precipitate and ~25 μm austenite-γ grains. Increased dislocation densities were observed adjacent to the grown-NbC precipitate, imparting a stress-corrosion effect on the AGR cladding corrosion.
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Rai, Subash. "Role of sulphur on carbon deposition on AGR fuel cladding steel." Thesis, University of Birmingham, 2019. http://etheses.bham.ac.uk//id/eprint/8882/.

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Filamentary carbon deposition on 20Cr/25Ni/Nb austenitic stainless steel from 1000-5000 ppm C2H4 in 1%CO/ bal. CO2 at 600°C has been investigated. This filamentary carbon deposition is catalysed by metallic nickel-rich particles formed intrinsically from the alloy during the early stages of oxidation. Samples were analysed using electron microscopy techniques. A simple model has been proposed to explain the formation of nickel-rich particles within the subsurface oxide layer. Furthermore, in this project the effect of COS, H2S and CH3SH on carbon deposition were examined. Addition of sulphur species suppressed the deposition process but the effect was dependent on the concentration of C2H4 and the sulphur species. This suppressing effect of sulphur on carbon deposition was explained in term of the role played by the adsorbed sulphur in reducing the rate of the key steps of the growth process of a filament.
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Carr, James. "Surface Modification Techniques for Increased Corrosion Tolerance of Zirconium Fuel Cladding." VCU Scholars Compass, 2016. http://scholarscompass.vcu.edu/etd/4474.

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Corrosion is a major issue in applications involving materials in normal and severe environments, especially when it involves corrosive fluids, high temperatures, and radiation. Left unaddressed, corrosion can lead to catastrophic failures, resulting in economic and environmental liabilities. In nuclear applications, where metals and alloys, such as steel and zirconium, are extensively em- ployed inside and outside of the nuclear reactor, corrosion accelerated by high temperatures, neu- tron radiation, and corrosive atmospheres, corrosion becomes even more concerning. The objec- tives of this research are to study and develop surface modification techniques to protect zirconium cladding by the incorporation of a specific barrier coating, and to understand the issues related to the compatibility of the coatings examined in this work. The final goal of this study is to recommend a coating and process that can be scaled-up for the consideration of manufacturing and economic limits. This dissertation study builds on previous accident tolerant fuel cladding research, but is unique in that advanced corrosion methods are tested and considerations for implementation by industry are practiced and discussed. This work will introduce unique studies involving the materials and methods for accident tolerant fuel cladding research by developing, demonstrating, and consid- ering materials and processes for modifying the surface of zircaloy fuel cladding. This innova- tive research suggests that improvements in the technique to modify the surface of zirconium fuel cladding are likely. Three elements selected for the investigation of their compatibility on zircaloy fuel cladding are aluminum, silicon, and chromium. These materials are also currently being investigated at other labs as alternate alloys and coatings for accident tolerant fuel cladding. This dissertation also investigates the compatibility of these three elements as surface modifiers, by comparing their mi- crostructural and mechanical properties. To test their application for use in corrosive atmospheres, the corrosion behaviors are also compared in steam, water, and boric-acid environments. Various methods of surface modification were attempted in this investigation, including dip coating, diffu- sion bonding, casting, sputtering, and evaporation. The benefits and drawbacks of each method are discussed with respect to manufacturing and economic limits. Characterization techniques utilized in this work include optical microscopy, scanning electron microscopy, energy-dispersive spec- troscopy, X-ray diffraction, nanoindentation, adhesion testing, and atomic force microscopy. The composition, microstructure, hardness, modulus, and coating adhesion were studied to provide en- compassing properties to determine suitable comparisons and to choose an ideal method to scale to industrial applications. The experiments, results, and detailed discussions are presented in the following chapters of this dissertation research.
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Jena, Anupam S. M. Massachusetts Institute of Technology. "Wettability of candidate Accident Tolerant Fuel (ATF) cladding materials in LWR conditions." Thesis, Massachusetts Institute of Technology, 2020. https://hdl.handle.net/1721.1/127299.

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Thesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, May, 2020
Cataloged from the official PDF of thesis.
Includes bibliographical references (pages [69]-70).
Since the 2011 Fukushima accident, substantial research has been dedicated to developing accident tolerant fuel (ATF) cladding materials. These analyses have mostly concentrated on the capability of potential ATF materials to withstand runaway steam oxidation and preserve their mechanical strength and structural integrity under thermal shocks. However, knowledge is relatively deficient for the thermal-hydraulic properties of these materials, particularly under light water reactor (LWR) operating conditions. The surface wettability is particularly important, as it affects the dynamics of the boiling heat transfer process, and consequently, the critical heat flux (CHF) and rewetting temperatures, which are important thermal limits for LWRs. Surface wettability determines nucleation site density, bubble departure diameter, and bubble departure frequency.
Therefore, it is essential to quantify the surface wettability of candidate ATF cladding materials to determine their thermal-hydraulic behavior compared to conventional Zircaloy claddings. The surface wettability is usually quantified through the sessile droplet contact angle, which is the angle formed between the liquid-vapor and the liquid-solid interface. The contact angle depends on the fluid, solid, surface finish, and operating conditions, i.e., temperature and pressure. However, most of the measurements available in the literature are performed at low pressure and in an inert atmosphere, which is quite different from the operating conditions of LWRs (i.e., in a steam-saturated atmosphere at a pressure as high as 15.5 MPa or 155 bars).
To close this gap, in this study, we designed and built an autoclave-type facility capable of measuring static, advancing, and receding contact angle in steam-saturated atmospheres, from sub-atmospheric conditions up to the critical point of water, i.e., 22.1 MPa (221 bar or 3200 psi) and 374°C. We measured the static contact angle of conventional Zircaloy-4 and candidate ATF cladding materials (e.g., Cr-coated Zr-4, FeCrAl, and SiC). The contact angle decreases with an increase in temperature for all the materials. Rough surfaces showed higher wettability, i.e., lower contact angle, compared to the smooth surfaces. These trends are expected from theory. All the materials showed different wettability under the same temperature and pressure conditions. Individual correlations for temperature dependence for each of them are proposed.
by Anupam Jena.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
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Books on the topic "Fuel cladding"

1

Agency, International Atomic Energy, ed. Management of cladding hulls and fuel hardware: Report of a technical committee meeting on management of cladding hulls and fuel hardware. Vienna: International Atomic Energy Agency, 1985.

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Flanagan, Michelle. Mechanical behavior of ballooned and ruptured cladding: Prepared by Michelle Flanagan. Washington, DC: U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2012.

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Novák, J. Mechanical properties of zirconium alloy cladding tubes and critical fuel element power ramps. Řež, Czechoslovakia: Nuclear Research Institute, Information Centre, 1988.

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Carrotte, Jerome. Thermomechanical aspects of pellet-cladding interaction in a pressurised water reactor fuel rod. Birmingham: University of Birmingham, 1995.

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Yegorova, L. A. Data base on the behavior of high burnup fuel rods with Zr-1% Nb cladding and UO2 fuel (VVER Type) under reactivity accident conditions. Washington, D.C: U.S. Nuclear Regulatory Commission, 1999.

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P, Moeller M., U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Division of Safety Review and Oversight., Pacific Northwest Laboratory, and U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Reactor Accident Analysis., eds. The Impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study. Washington, D.C: Division of Safety Review and Oversight, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1986.

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A, Yegorova L., U.S. Nuclear Regulatory Commission., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research., Joint Stock Company "TVEL", Institut problem bezopasnogo ispolʹzovanii︠a︡ i︠a︡dernoĭ ėnergii (Rossiĭskiĭ nauchnyĭ t︠s︡entr "Kurchatovskiĭ institut"), Institut de radioprotection et de sûreté nucléaire., and Gosudarstvennyĭ nauchnyĭ t︠s︡entr Rossiĭskoĭ Federat︠s︡ii "Nauchno-issledovatelʹskiĭ institut atomnykh reaktorov", eds. Experimental study of embrittlement of Zr-1%Nb VVER cladding under LOCA-relevant conditions. Washington, D.C: U.S. Nuclear Regulatory Commission, 2005.

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P, Kaplar E., ed. Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding: Procedures and results of low temperature biaxial burst tests and axial tensile tests. Washington, D.C: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 2001.

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P, Kaplar E., ed. Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding: Procedures and results of low temperature biaxial burst tests and axial tensile tests. Washington, D.C: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 2001.

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T, Berta V., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research., Idaho National Engineering Laboratory, and EG & G Idaho., eds. Determination of the bias in LOFT fuel peak cladding temperature data from the blowdown phase of large-break LOCA experiments. Washington, DC: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

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Book chapters on the topic "Fuel cladding"

1

Khan, Mansoor A., Nels H. Madsen, and Bryan A. Chin. "Fracture Predictions in Zircaloy Fuel Cladding." In Effects of Radiation on Materials: 12th International Symposium Volume II, 642–55. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1985. http://dx.doi.org/10.1520/stp87019850007.

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Yang, Guangliang, Weixiang Wang, Wenpei Feng, Tao Ding, and Hongli Chen. "Study on the Steady-State Performance of the Fuel Rod in M2LFR-1000 Using KMC-Fueltra." In Springer Proceedings in Physics, 703–16. Singapore: Springer Nature Singapore, 2023. http://dx.doi.org/10.1007/978-981-99-1023-6_60.

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AbstractOperating conditions in the liquid metal fast reactor, like higher power density, higher temperature gradient, and higher burnup, are more severe to the fuel rod comparing to the light water reactor. The integrity and safety of the fuel rod is very essential to the reactor safety. In this study, the fuel rod designed for M2LFR-1000, which is a typical pool type lead cooled fast reactor, is evaluated using a fuel performance analysis code named KMC-Fueltra. Irradiation behaviors and material properties for the MOX fuel and T91 cladding are established and introduced into the code. The steady-state performance of the fuel rod is analyzed. Results concerning both fuel and cladding performance are discussed based on indicative design limits collected from the open literatures. This study is useful to improve the conceptual design of the M2LFR-1000.
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Maksymov, Maksym, Svitlana Alyokhina, and Oleksandr Brunetkin. "Analysis of WWER 1000 Fuel Cladding Failure." In Thermal and Reliability Criteria for Nuclear Fuel Safety, 151–96. New York: River Publishers, 2022. http://dx.doi.org/10.1201/9781003339816-5.

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Shevyakov, Alexander Yu, Vladimir V. Novikov, Vladimir A. Markelov, Kalin Lafchiev, Kyle D. Johnson, Daniel Jädernäs, Alexander V. Ugryumov, Alexander F. Radostin, and Rasmus Waginder. "E110opt Fuel Cladding Corrosion under PWR Conditions." In Zirconium in the Nuclear Industry: 20th International Symposium, 313–30. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2023. http://dx.doi.org/10.1520/stp164520220021.

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Filburn, Thomas, and Stephan Bullard. "Nuclear Fuel, Cladding, and the “Discovery” of Zirconium." In Three Mile Island, Chernobyl and Fukushima, 105–14. Cham: Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-34055-5_10.

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Motta, Arthur T. "Mechanistic Understanding of Zirconium Alloy Fuel Cladding Performance." In Zirconium in the Nuclear Industry: 18th International Symposium, 19–51. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2018. http://dx.doi.org/10.1520/stp159720160095.

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Kai, He, Song Zifan, Zheng Yuntao, Jiang Xiaochuan, Yang Changjiang, and Yang Wei. "Spent Fuel Cladding Performance Analysis Under Spent Fuel Pool Boiling-off Accident." In Proceedings of The 20th Pacific Basin Nuclear Conference, 595–99. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2311-8_55.

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Nakazato, Naofumi, Hirotatsu Kishimoto, Yutaka Kohno, and Akira Kohyama. "Sic/Sic Fuel Cladding by Nite Process for Innovative LWR-Cladding Forming Process Development." In Ceramic Transactions Series, 109–15. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2014. http://dx.doi.org/10.1002/9781118771327.ch12.

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Field, Kevin G., Yukinori Yamamoto, Bruce A. Pint, Maxim N. Gussev, and Kurt A. Terrani. "Accident Tolerant FeCrAl Fuel Cladding: Current Status Towards Commercialization." In The Minerals, Metals & Materials Series, 1381–89. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-030-04639-2_91.

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Daub, K., S. Y. Persaud, R. B. Rebak, R. Van Nieuwenhove, S. Ramamurthy, and H. Nordin. "Investigating Potential Accident Tolerant Fuel Cladding Materials and Coatings." In The Minerals, Metals & Materials Series, 1431–50. Cham: Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-030-04639-2_95.

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Conference papers on the topic "Fuel cladding"

1

Deng, Yangbin, Bowen Qiu, Yingwei Wu, Dalin Zhang, Wenxi Tian, Suizheng Qiu, and G. H. Su. "Simulation on Pellet-Cladding Mechanical Interaction (PCMI) of Accident Tolerant Fuel (ATF) With Coated Cladding." In 2017 25th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2017. http://dx.doi.org/10.1115/icone25-66774.

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In this study, based on the code FROBA (Fuel ROd Behavior Analysis), a thermal-mechanical analysis code initially developed for traditional UO2-Zr fuel elements by our research group, a modified version named FROBA-ATF was developed to perform the fuel performance simulation of ATFs with different claddings, including Zr-4, SiC and Zr-4 coated with SiC. Compared with initial version, the cladding could be divided into arbitrary number control volumes with different materials in the new code, so it can be used to perform the calculation for multilayer coatings. In addition, a new non-rigid PCMI calculation model was established in the new code. Neither of the cladding and the pellet was regarded as the rigid body in this study, which means it can provide more accurate prediction compared with the rigid-fuel model in the initial code when Pellet-cladding Mechanical Interaction (PCMI) happened. The FROBA-ATF code was used to predict PCMI performance of two kind fuels with coated claddings, including the internal-surface coating and external-surface coating. The calculation result indicates that because the coating surface was close to the inner surface of the clad where also was the PCMI surface, the absolute value of the combine pressure of internal-surface coated cladding was substantial larger than that of the external-surface coated cladding, which might be harmful the coating behavior. However, the internal-surface coated mode can provide a protection for alloy due to the isolation from direct contact with fuel pellets, which can result in a much lower equivalent stress of zirconium body during the PCMI.
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Pooja, Nupur Aggarwal, and Naveen Kumar. "Accident tolerant fuel cladding material challenges." In 2ND INTERNATIONAL CONFERENCE ON RECENT ADVANCES IN COMPUTATIONAL TECHNIQUES. AIP Publishing, 2023. http://dx.doi.org/10.1063/5.0141087.

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Dabney, Tyler, Hwasung Yeom, Kyle Quillin, Nick Pocquette, and Kumar Sridharan. "Cold Spray Technology for Oxidation-Resistant Nuclear Fuel Cladding." In ITSC2021, edited by F. Azarmi, X. Chen, J. Cizek, C. Cojocaru, B. Jodoin, H. Koivuluoto, Y. C. Lau, et al. ASM International, 2021. http://dx.doi.org/10.31399/asm.cp.itsc2021p0167.

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Abstract Light water reactors (LWR) use zirconium-alloy fuel claddings; the tubes that hold the uranium-dioxide fuel pellets. Zr-alloys have very good neutron transparency; but during a loss of coolant accident or beyond design basis accident (BDBA) they can undergo excessive oxidation in reaction with the surrounding steam environment. Relatively thin oxidationresistant coatings on Zr-alloy fuel cladding tubes can potentially buy coping time in these off-normal scenarios. In this study; cold spraying; solid-state powder-based materials deposition technology has been developed for deposition of oxidation-resistant Cr coatings on Zr-alloy cladding tubes; and the ensuing microstructure and properties of the coatings have been investigated. The coatings when deposited under optimum conditions have very good hydrothermal corrosion resistance as well as oxidation resistance in air and steam environments at temperatures in excess of 1100 °C; while maintaining excellent adhesion to the substrate. These and other results of this study; including mechanical property evaluations; will be presented.
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Crede, Timothy, Julianna Schoenwald, and Brian Mount. "Hydrogen-Based Transient Cladding Strain Limit." In TopFuel 2022 Light Water Reactor Fuel Performance Conference. Illinois: American Nuclear Society, 2022. http://dx.doi.org/10.13182/topfuel22-38994.

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Liu, Rong, Liwen Yang, and Shengyu Liu. "Multiphysics Analysis of Fuel Performance and Tritium Migration in FeCrAl and Cr-Coated Zircalloy Cladding Under PWR Normal Operating and Transient Conditions." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-91981.

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Abstract Due to the fact that the Cr-coated Zircalloy and the FeCrAl alloy have excellent high-temperature oxidation resistance, corrosion resistance, they are considered to be potential replacement materials for PWR cladding. However, it is also reported that the poor tritium resistance of FeCrAl alloy will accelerate the penetration of tritium throughout the cladding and increase the treatment cost. In this work, the physical models and tritium migration models of Cr-coated Zircalloy cladding and FeCrAl claddingare reviewed firstly. Then, based on the developed fuel performance analysis code CAMPUS, the fuel performance models of Cr-coated Zircalloy cladding and FeCrAl cladding are implemented, and a new model of the tritium migration in different claddings is implemented too. Finally, the performances of three fuel cladding combinations (UO2 pellets and Zircalloy cladding, Cr-coated Zircalloy cladding and FeCrAl cladding) under normal and LOCA conditions are simulated and discussed. And the tritium resistance performance of different fuel cladding combinations is further calculated and analyzed with the new tritium migration model. The calculation results show that under normal and LOCA conditions, compared with Zircalloy cladding, the application of FeCrAl cladding can reduce the overall temperature of fuel and improve fuel safety margin. And the application of Cr-coated Zircalloy cladding and FeCrAl cladding can effectively delay the failure time of cladding under LOCA condition; In terms of tritium penetration, the Cr-coated Zircalloy is found to have better tritium resistance effect than FeCrAl alloy, which greatly reduces the pollution treatment cost of reactor operation.
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Estevez, Samanta, Johan Stjärnsäter, Mi Wang, Carolina Losin, Daniel Jädernäs, Hun Jang, Okjoo Kim, Yoonho Kim, and Jaeik Kim. "Post Irradiation Examinations of HANA-6 Cladding." In TopFuel 2022 Light Water Reactor Fuel Performance Conference. Illinois: American Nuclear Society, 2022. http://dx.doi.org/10.13182/topfuel22-39134.

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Altahhan, Muhammad, Noah McFerran, Jonathan Morrell, and Maria Avramova. "Multiphysics Analysis of CMC Silicon Carbide and Zircaloy Cladding." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81464.

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Nuclear fuel cladding is an integral part of nuclear reactors and choosing the proper material is imperative to the design of a reactor. In this paper, the neutronic properties and the fuel performance of a 17 × 17 Westinghouse Pressurized Water Reactor (PWR) assembly using ceramic matrix composites (CMC) Silicone Carbide (SiC) as a cladding material is investigated. The material analysis is compared against traditional Zircaloy-4 cladding used in a 17 × 17 Westinghouse PWR assembly. The codes used in the analysis are the Michigan Parallel Characteristics based Transport (MPACT) code coupled with CTF, the North Carolina State University version of the Coolant Boiling in Rod Arrays Two Fluids (COBRA-TF) code, and the fuel performance code BISON as well as the uncertainty analysis code DAKOTA. Additionally, annular geometry for the fuel pellet is modeled to assess its merit compared to ordinary CMC SiC or traditional Zircaloy-4 claddings. It is found that on the neutronics side, the CMC SiC shows lower achievable U-235 enrichments required to reach the same burnup and effective neutron multiplication factor as Zircalloy-4 claddings. These results are an advantage that can be seen in the economic cost analysis done and additionally from the reactor operation point of view. Also, it is found that the different criteria of safe operation of Westinghouse PWR assemblies like the plenum pressure, the fuel-cladding contact pressure, the peak fuel temperature, and the fission gas release criteria are all achieved with CMC SiC with some criteria having larger design margins than of the Zircaloy-4 cladding. Furthermore, a critical heat flux (CHF) study shows that CMC SiC has even larger thermal margins than the ordinary Zircaloy-4 cladding, resulting in a more profitable fuel cycle due to the greater amount of power that the fuel pins can be operated at. An uncertainty quantification for the CHF Ratio (CHFR) is also done to assess the largest magnitudes of importance that affect the CHFR calculated.
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Erickson, A., M. Short, and N. Fassino. "Fuel-Cladding Mechanical Interactions in a Small Pin-Type FHR with SiC Cladding." In 2020 ANS Virtual Winter Meeting. AMNS, 2020. http://dx.doi.org/10.13182/t123-33137.

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Schrire, David, Cecilia Janzon, and Daniel Jädernäs. "Axial Dependency of BWR Zircaloy-2 Cladding Oxidation." In TopFuel 2022 Light Water Reactor Fuel Performance Conference. Illinois: American Nuclear Society, 2022. http://dx.doi.org/10.13182/topfuel22-38914.

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Jädernäs, Daniel, Peter Gillén, Joakim Karlsson, Fredrik Gustavsson, David Schrire, and John Beale. "Detailed Examinations of Debris Fretting Wear on Cladding." In TopFuel 2022 Light Water Reactor Fuel Performance Conference. Illinois: American Nuclear Society, 2022. http://dx.doi.org/10.13182/topfuel22-38925.

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Reports on the topic "Fuel cladding"

1

Wood, E. L., and D. L. Porter. Fuel/cladding compatibility of U-10Zr and U-5Fs fuels with advanced alloy cladding materials. Office of Scientific and Technical Information (OSTI), May 1985. http://dx.doi.org/10.2172/711868.

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Keiser, D., and M. Dayananda. Interdiffusion Studies for Fuel-Cladding Compatibility in IFR Fuels. Office of Scientific and Technical Information (OSTI), December 1993. http://dx.doi.org/10.2172/2328556.

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Leibowitz, L. Phase relations for fuel-cladding interactions. Office of Scientific and Technical Information (OSTI), October 1986. http://dx.doi.org/10.2172/712838.

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Wood, Elizabeth Sooby. Experimental assessment of fuel-cladding interactions. Office of Scientific and Technical Information (OSTI), June 2017. http://dx.doi.org/10.2172/1367820.

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Galloway, Jack, and Cetin Unal. Accident Tolerant Fuel and Cladding Assessment. Office of Scientific and Technical Information (OSTI), August 2014. http://dx.doi.org/10.2172/1150663.

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Rudisill, T., and J. John Mickalonis. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS. Office of Scientific and Technical Information (OSTI), September 2006. http://dx.doi.org/10.2172/939424.

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Sridharan, Kumar, Todd Allen, Jesse Gudmundson, and Benjamin Maier. Surface Modification of Fuel Cladding Materials with Integral Fuel BUrnable Absorber Boron. Office of Scientific and Technical Information (OSTI), November 2008. http://dx.doi.org/10.2172/940909.

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Fraker, Anna C. Corrosion of zircaloy spent fuel cladding in a repository. Gaithersburg, MD: National Institute of Standards and Technology, 1989. http://dx.doi.org/10.6028/nist.ir.89-4114.

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Dryepondt, Sebastien N., Kinga A. Unocic, David T. Hoelzer, and Bruce A. Pint. Advanced ODS FeCrAl alloys for accident-tolerant fuel cladding. Office of Scientific and Technical Information (OSTI), September 2014. http://dx.doi.org/10.2172/1150908.

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Perez, Emmanuel, Dennis D. Keiser, Jr., Bryan Forsmann, Dawn E. Janney, Jody Henley, and Eric C. Woolstenhulme. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel. Office of Scientific and Technical Information (OSTI), February 2016. http://dx.doi.org/10.2172/1259949.

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