• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.423-430
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• 2014

An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.

• 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.

• 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.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.3
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• pp.151-164
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• 2004

The changes of microstructural and mechanical properties were evaluated for the high burn-up fuel claddings after the neutron irradiation of $1.8{\sim}3.1{\times}10^{20}n/cm^2$ (E>1.0 MEV) in HANARO research reactor. After the irradiation, the spot-type dislocations (a-type dislocations) were easily observed in most claddings, and the density of the dislocations was different depending on the grains and was higher at grain boundaries than within grains. As the final annealing temperature increased, the density of spot-type dislocations increased and the line-type dislocations (c-type dislocations) which was perpendicular to the <0002> direction, appeared sporadically in some claddings. However, the types of precipitates in the fuel claddings after the irradiation were not changed from that in unirradiated claddings. The mechanical properties including the hardness, strength and elongation after the irradiation were changed due to the formation of spot-type dislocations. That is, the increase in hardness and strength as well as the decrease in elongation after the irradiation was occurred simultaneously with increasing the final annealing temperature. Owing to the Nb contribution to the formation of spot-type dislocation during the irradiation, the increase in hardness and strength in higher Nb-contained Zr alloys after the irradiation was higher than that in lower Nb-contained Zr alloys.

• 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 Metals and Materials
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• v.48 no.8
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• pp.717-723
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• 2010

The degradation behavior of fuel cladding is a very import concern in nuclear power generation, because the operation of nuclear plants can be limited by fuel cladding degradation. In order to evaluate the hydride effect on failure of zirconium fuel claddings, a ring tensile test for the circumferential direction was carried out at room temperature for claddings having different hydride characteristics such as density and orientation; microstructural evaluation was also performed for those claddings. The circumferential failure of the claddings was promoted by increasing the hydride concentration in the matrix; however, the failure of the claddings was affected by the hydride orientation rather than by the hydride concentration in the matrix. From fracture surface observation, the cladding failure during the ring tensile test was matched with the hydride orientation.

• 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.53 no.11
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• pp.3704-3710
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• 2021

The characteristics of the external radiation on the surface of the casks for spent nuclear fuel (SNF) storage by dry method are investigated for the case when the spatial distribution of SNF in the basket changes due to the destruction of the fuel rod claddings. The surface areas are determined, where the changes in fluxes of neutrons, produced by ²⁴⁴Cm actinide, and γ-quanta, produced by long-lived isotopes, are maximum in the result of the decrease in the height of the SNF area. Concrete (VSC-24) and metal (SC-21) casks are considered as examples. The procedure of periodic measurement of the dose rate of neutrons or γ-quanta at the specified points of the cask surface is proposed for identifying the fuel rod cladding destruction. Under normal operation, the decrease in the dose rate produced by neutrons as the function of SNF storage duration is determined by the half-life of ²⁴⁴Cm, and for γ-quanta - by the half-lives of long-lived SNF isotopes. Consequently, a stepwise change in the dose rate of neutrons or γ-quanta, detected by the measurements, as compared to the previous one, would indicate the destruction of the fuel rod claddings.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1525-1531
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• 2019

The indigenous nuclear power program of India is based mainly on a series of Pressurised Heavy Water Reactors (PHWRs). A burst correlation for Indian PHWR fuel claddings has been developed and empirical burst parameters are determined. The burst correlation is developed from data available in literature for single-rod transient burst tests performed on Indian PHWR claddings in inert environment. The heating rate and internal overpressure were in the range of 7 K/s-73 K/s and 3 bar-80 bar, respectively, during the burst tests. A burst criterion for inert environment, which assumes that deformation is controlled by steady state creep, has been developed using the empirical burst parameters. The burst criterion has been validated with experimental data reported in literature and the prediction of burst parameters is in a fairly good agreement with the experimental data. The burst criterion model reveals that increasing the heating rate increases the burst temperature. However, at higher heating rates, burst strain is decreased considerably and an early rupture of the claddings without undergoing considerable ballooning is observed. It is also found that the degree of anisotropy has significant influence on the burst temperature and burst strain. With increasing degree of anisotropy, the burst temperature for claddings increases but there is a decrease in the burst strain. The effect of anisotropy in the ${\alpha}$-phase is carried over to ${\alpha}+{\beta}$-phase and its effect on the burst strain in the ${\alpha}+{\beta}$-phase too can be observed.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.425-426
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• 2005

1. The oxide thickness on the fuelled test rods was within the following range from 7 ${\mu}m$ to 17 ${\mu}m$. In general, the HANA claddings showed better corrosion behavior than the two reference alloys (A-Cladding and Zr-4). 2. The weight gains of corrosion coupons were ranged from 21 to 56 mg/$dm^2$.