• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.372-378
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• 2000

The effect of alloying elements on the creep resistance of Zr alloys was investigated using thermal creep tests that were performed as a part of advanced fuel cladding development. The creep tests were conducted at 40$0^{\circ}C$ and 150 MPa for 240 hr. A statistical model was derived from the relationship between the steady-state creep rate and the content of individual alloying elements. The creep strengthening effect decreased in the following sequence : Nb, Sn, Mn, Cr, Mo, Fe and Cu. The high creep resistance of Sn and the opposite effect of Fe on zirconium alloys seem to be associated with their lowering and enhancing, respectively, the self-diffusivity of the zirconium matrix.

• Corrosion Science and Technology
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• v.2 no.6
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• pp.253-259
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• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.423-424
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• 2005

The performance of HANA claddings was evaluated in out-of-pile conditions. All the performance test results revealed that HANA claddings were superior to the reference claddings such as Zircaloy-4 and A-cladding. Corrosion resistance was improved by 60 to 70% compared to the commercial claddings. Creep, burst, tensile, LOCA, wear and microstructural properties were shown to be as good as the commercial claddings.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2003.11a
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• pp.64-71
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• 2003

In order to evaluate the effect of alloying elements (Nb and Sn) on the wear resistance of advanced Zr fuel claddings, sliding wear tests have been performed in room temperature air and water and these results were compared with those of commercial alloys such as Zircaloy-4, A and B alloys. As a result, the advanced Zr fuel claddings have a similar wear resistance compared with the commercial alloys. The wear resistance of the advanced Zr fuel claddings is closely releted to the content of Nb and Sn even though the effects of transition elements are involved in deforming wear properties. In the tested specimens with similar Sn content, wear volume became down to a minimum at $0.4\;wt\;\%$ Nb, then rapidly increased at 1.0 wt Nb. This behavior results in the variation of grain size with alloying contents. But Sn did not have a significant effect on the wear volume of advanced Zr fuel claddings below $1.1\;wt\%$. The relationship between alloying elements and wear behaviour was evaluated and discussed using material compatibility factor.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.163-170
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• 2009

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were $M5^{(R)}$ and $ZIRLO^{TM}$, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though $M5^{(R)}$ shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup $M5^{(R)}$ and $ZIRLO^{TM}$ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.1-8
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• 2008

One of the major current challenges to nuclear energy lies in its competitiveness. To stay competitive the industry needs to reduce maintenance and fuel cycle costs, while enhancing safety features. Extended burnup is one of the methods applied to meet these objectives However, there are a number of potential fuel failure causes related to increased burnup, as follows: l) Corrosion of zirconium alloy cladding and the water chemistry parameters that enhance corrosion; 2) Dimensional changes of zirconium alloy components, 3) Stresses that challenge zirconium alloy ductility and the effect of hydrogen (H) pickup and redistribution as it affects ductility, 4) Fuel rod internal pressure, 5) Pellet-cladding interactions (PCI) and 6) pellet-cladding mechanical interactions (PCMI). This paper discusses current and potential failure mechanisms of these failure mechanisms.

• 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.

• 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.

• Korean Journal of Materials Research
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• v.11 no.5
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• pp.405-412
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• 2001

Corrosion tests were carried out in $360^{\circ}C$ water and $360^{\circ}C$ 70ppm LiOH solution to investigate the corrosion behavior of new zirconium alloys (Zr-0.4Nb-0.8Sn-xFeCrMn, Zr-0.2Nb-1.1Sn-xFeCrMn, Zr-1.0Nb-xFeCu). Microstructures of tested alloys were analyzed by optical microscope and TEM. The cross-sectional surface and crystalline structure of the oxide layer were analyzed by SEM and XRD. From the results of corrosion test, all the alloys showed higher corrosion rates in $360^{\circ}C$ 70ppm LiOH aqueous solution thats in $360^{\circ}C$ water. Especially, high Nb-containing alloy exhibited the acceleration of corrosion rate in LiOH solution. The low Nb- and Sn-added alloys showed better corrosion resistance than the Sn- free high Nb alloy. from the effect of final annealing on the corrosion, it was observed that the partially recrystallized alloys showed better corrosion resistance than fully recrystallized alloys. This would be related to the size and distribution of the second phase particles.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.