• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.741-751
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• 2021

There has been a deep interest in trying to find better-performing fuel clad motivated by the desire to decrease the likelihood of the reactor barrier failure like what happened in Fukushima in recent years. In this study, the effect of move towards accident tolerant fuel (ATF) cladding as the most attracting concept for improving reactor safety is investigated for SMART modular reactor. These reactors have less production cost, short construction time, better safety and higher power density. The SiC and FeCrAl materials are considered as the most potential candidate for ATF cladding, and the results are compared with Zircaloy cladding material from reactor physics point of view. In this paper, the calculations are performed by generating PMAX library by DRAGON lattice physics code to be used for further reactor core analysis by PARCS code. The differential and integral worth of control and safety rods, reactivity coefficient, power and temperature distributions, and boric acid concentration during the cycle are analyzed and compared from the conventional fuel cladding. The rod ejection accident (REA) is also performed to study how the power changed in response to presence of the ATF cladding in the reactor core. The key quantitative finding can be summarized as: 20 ℃ (3%) decrease in average fuel temperature, 33 pcm (3%) increase in integral rod worth and cycle length, 1.26 pcm/℃ (50%) and 1.05 pcm/℃ (16%) increase in reactivity coefficient of fuel and moderator, respectively.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3741-3758
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• 2021

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.2
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• pp.231-238
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• 2003

The fuel of light water reactor is used for several years under high temperature and pressure, so it needs to be clad with high corrosion resistance material. The cladding materials must have the characteristics of low absorption of a neutron and high corrosion resistance. Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor have been used as cladding materials and Zirlo has been developed as the material for preventing the corrosion. If the fracture of the cladding tube occurs during operation, it will cause the economic loss to shut down and replace the system. So it is needed to evaluate the integrity of the cladding materials. In this paper, the tensile characteristics of the cladding materials were investigated for the basic research of fracture characteristics. Also the residual stress was analyzed to compare the tube type(original type) specimen and the flattened type specimen.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.489-494
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• 2001

The fuel of light water reactor used far several years at high temperature and pressure, so it needs to clad with high corrosion resistance material. The cladding materials need low absorption of a neutron and high corrosion resistance. Cladding materials used Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor and Zirlo has good for long term corrosion. If fracture of cladding tube occured during operation, it caused disaster. So it is needed to estimate of integrity fur cladding materials. In this paper, tension characteristics of cladding materials are investigate which is basic research far fracture characteristic. Also analysis of residual stress effect between tube type(original type) specimen and flattened type specimen.

• Journal of the Korean Society for Heat Treatment
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• v.18 no.4
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• pp.235-241
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• 2005

Thermal creep properties of the advanced zirconium fuel claddings named by HANA alloys which were developed for high burn-up application were evaluated. The creep test of HANA cladding tubes was carried out by the internal pressurization method in temperature range from 350 to $400^{\circ}C$ and in the hoop stress range from 100 to 150 MPa. Creep tests were lasted up to 800 days, which showed the steady-state secondary creep rate. The creep resistance of HANA fuel claddings was affected by final annealing temperature and various factors, such as alloying element, applied stress and testing temperature. From the results the microstructure observation of the samples before and after creep test by using TEM, the dislocation density was increased in the sample of after creep test. The Sn as an alloying element was more effective in the creep resistance than other elements such as Nb, Fe, Cr and Cu due to solute hardening effect of Sn. In case of HANA fuel claddings, the improved creep resistance was obtained by the control of final heat treatment temperature as well as alloying element.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.178-187
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• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

• Korean Journal of Metals and Materials
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• v.48 no.5
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• pp.410-416
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• 2010

Studies were carried out to establish the technology for sodium-clad compatibility and to analyze the compatibility behavior of the Sodium-cooled Fast Reactor (SFR) cladding material under a flowing sodium environment. The natural circulation facility caused by the thermal convection of the liquid sodium was constructed and the 316-series stainless steels were exposed at $650{^{\circ}C}$ liquid sodium for 1458 hours. The weight change and related microstructural change were analyzed. The results showed that the quasi-dynamic facility represented by the natural convection exhibited similar results compared to the conventional dynamic facility. Selective leaching and local depletion of the chromium, re-distribution of the carbide, and the decarburization process took place in the 316-series stainless steel under a flowing sodium environment. This process decreased as the sodium flowed along the channel, which was caused by the change in the dissolved oxygen and carbon activity in the liquid sodium.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.432-442
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• 2023

Characterizing the behavior of nuclear reactor plate fuels is vital to the progression of advanced fuel systems. The states of pre- and post-irradiation plates need to be determined effectively and efficiently prior to and following irradiation. Due to the hostile post-irradiation environment, characterization must be completed remotely. Laser-based characterization techniques enable the ability to make robust measurements inside a hot-cell environment. The Laser Shock (LS) technique generates high energy shockwaves that propagate through the plate and mechanically characterizes cladding-cladding interfaces. During an irradiation campaign, two Idaho National Laboratory (INL) fabricated MP-1 plates had a fuel breach in the cladding-cladding interface and trace amounts of fission products were released. The objective of this report is to characterize the cladding-cladding interface strengths in three plates fabricated using different fabrication processes. The goal is to assess the risk in irradiating future developmental and production fuel plates. Prior LS testing has shown weaker and more variability in bond strengths within INL MP-1 reference plates than in commercially produced vendor plates. Three fuel plates fabricated with different fabrication processes will be used to bound the bond strength threshold for plate irradiation insertion and assess the confidence of this threshold value.

• Tribology and Lubricants
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• v.38 no.4
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• pp.136-142
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• 2022

This paper investigates the enhanced wear performance of a CrAl coated accident tolerant fuel (ATF) cladding. In the wake of the Fukushima accident, extensive research on ATF with respect to improving the oxidation resistance of cladding materials is ongoing. Since coated Zr claddings can be applied without major changes to the criteria for reactor core design, many researchers are studying coatings for claddings. To improve the quality of the CrAl coating layer, optimization of the manufacturing process is imperative. This study employs arc ion plating to obtain improved CrAl coated claddings using CrAl binary alloy targets through an improved coating method. Surface roughness and adhesion are improved, and droplets are reduced. Furthermore, the coated layer has a dense and fine microstructure. In scratch tests, all the tested CrAl coated claddings exhibit a superior resistance compared to the Zr cladding. In a fretting wear test, the wear volume of the CrAl coated claddings is smaller compared to the Zr cladding. Furthermore, the coated cladding manufactured through the improved process exhibits better wear resistance than other CrAl coated claddings. Based on these results, we suggest that fine microstructure is attributed to a mechanically and microstructurally robust CrAl coating layer, which enhances wear resistance.

• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.591-602
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• 2007

The results of Reactivity-Initiated Accidents (RIA) experiments have been analysed and the main variables affecting the fuel failure propensity identified. Fuel burn-up aggravates the mechanical loading of the cladding, while corrosion, or better the hydrogen absorbed in the cladding as a consequence of corrosion, may under some conditions make the cladding brittle and more susceptible to failure. Experiments point out that corrosion impairs the fuel resistance for RIA transient occurring at cold conditions, whereas there is no evidence of important embrittlement effects at hot conditions, unless the cladding was degraded by oxide spalling. A fuel failure threshold correlation has been derived and compared with experimental data relevant for BWR and PWR fuel. The correlation can be applied to both cold and hot RIA transients, account taken for the lower ductility at cold conditions and for the different initial enthalpy. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. The proposed threshold is easy to use and reproduces the results obtained in the CABRI and NSRR tests in a rather satisfactory manner. The behaviour of advanced PWR alloys and of MOX fuel is discussed in light of the correlation predictions. Finally, a probabilistic approach has been developed in order to account for the small scatter of the failure predictions. This approach completes the RIA failure assessment in that after determining a best estimate failure threshold, a failure probability is inferred based on the spreading of data around the calculated best estimate value.