• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3535-3542
• /
• 2023

Eutectic reactions of five kinds of Cr-coated Zr alloy cladding with different base materials (Zr-Nb-Sn alloy or Zr-Nb alloy), different coating thicknesses (6~22.5 mm), and different coating materials (Cr single layer or Cr/CrN bilayer) were studied using Differential Scanning Calorimetry (DSC). The DSC experiments demonstrated that the onset temperatures of the Cr single layer coated specimens were almost identical to ~1308 ℃, regardless of base materials or coating thicknesses. This study demonstrated that the Cr/CrN bilayer coated Zr-Nb-Sn alloy has a slightly (~10 ℃) higher eutectic onset temperature compared to the single Cr-coated specimen. The eutectic region characterized by post-eutectic microstructure proportionally increases with coating thickness. The post-eutectic characterization with different holding times at high temperature (1310-1330 ℃) reveals that progression of Zr-Cr eutectic requires time, and it dramatically changed with exposure time and temperature. The practical value of the time gain in non-instantaneous eutectic formation in terms of safety margin, however, seems to be limited.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.279-279
• /
• 1999

To improve the corrosion model of the fuel performance analysis code COSMOS, a model was developed considering thermohydraulic phenomena and the effect of water chemistry and low Sn in the alloy composition on the corrosion behavior. It is assumed that the lithium enhancement factor influences the corrosion behavior only if the subcooled void is present in the coolant. The developed model was verified with the database obtained from Grohnde and Ringhals 3 reactors. Comparison of predicted oxide thickness with measured data showed the applicability of COSMOS code to analyze the cladding oxidation. In the future, the effect of the hydride in the cladding and the precipitate changes due to irradiation should be included.cluded.

• Korean Journal of Materials Research
• /
• v.14 no.7
• /
• pp.451-456
• /
• 2004

Thermal creep properties of the zirconium tube which was developed for high burnup application were evaluated. The creep test of cladding tubes after various final heat treatment was carried out by the internal pressurization method in the temperature range from $350^{\circ}C to 400^{\circ}C$ and from 100 to 150 MPa in the hoop stress. Creep tests were lasted up to 900days, which showed the steady-state secondary creep rate. The creep resistance of zirconium claddings was higher than that of Zircaloy-4. Factors that affect creep resistance, such as final annealing temperature, applied stress and alloying element were discussed. Tin as an alloying element was more effective than niobium due to solute hardening effect of tin. In case of advanced claddings, the optimization of final heat treatment temperature as well as alloying element causes a great influence on the improvement of creep resistance.

• Korean Journal of Materials Research
• /
• v.12 no.2
• /
• pp.140-145
• /
• 2002

This study was conducted to investigate the brazing characteristics between Zircaloy-4 nuclear fuel cladding tubes and bearing pads with filler metals of amorphous $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.5) binary alloy, in which they were produced in the ribbon form by the melt-spinning metod. The crystallization behavior, stability, hardness and micro-structure of brazed zone were examined by X-ray diffraction, differential scanning calorimetry, micro-Vickers hardness test, optical microscopy, and transmission electron microscopy. $Zr_{1-x}Be_x$(0.3$\leq$x$\leq$0.4) amorphous alloys were crystallized to $\alpha$-Zr with increasing the temperature, and the rest were transformed to ZrBe$_2$at higher temperatures. On the other hand, $Zr_{1-x}Be_x$(0.4$\leq$x$\leq$0.5) amorphous alloys were crystallized to $\alpha$-Zr and ZrBe$_2$, simultaneously. The thickness of the layer brazed with amorphous alloy was increased with increasing the beryllium content due to the higher diffusion of Be. The morphology of brazed layer with PVD Be filler metal showed dendrite while that brazed with amorphous alloys appeared globular. Micro-Vickers hardness of brazed zone increased as the beryllium content of filler metal was decreased.

• Nuclear Engineering and Technology
• /
• v.43 no.2
• /
• pp.149-158
• /
• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• Nuclear Engineering and Technology
• /
• v.51 no.1
• /
• pp.238-248
• /
• 2019

In the high-temperature and high-pressure irradiation environments, the multi-field coupling processes of hydrogen diffusion, hydride precipitation and mechanical deformation in Zircaloy cladding tubes occur. To simulate this hydrogen-induced complex behavior, a multi-field coupling method is developed, with the irradiation hardening effects and hydride-precipitation-induced expansion and hardening effects involved in the mechanical constitutive relation. The out-pile tests for a cracked cladding tube after irradiation are simulated, and the numerical results of the multi-fields at different temperatures are obtained and analyzed. The results indicate that: (1) the hydrostatic stress gradient is the fundamental factor to activate the hydrogen-induced multi-field coupling behavior excluding the temperature gradient; (2) in the local crack-tip region, hydrides will precipitate faster at the considered higher temperatures, which can be fundamentally attributed to the sensitivity of TSSP and hydrogen diffusion coefficient to temperature. The mechanism is partly explained for the enlarged velocity values of delayed hydride cracking (DHC) at high temperatures before crack arrest. This work lays a foundation for the future research on DHC.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.7 no.2
• /
• pp.113-119
• /
• 2006

Brazing is an important manufacturing process in the fabrication of Heavy Water Reactor fuel rods, in which bearing and spacer pads are joined to Zircaloy-4 cladding tubes. The physical vapor deposition(PVD) technique is currently used to deposit metallic Be on the surfaces of pads as a filler metal. Amorphous Zr-Be binary alloys which are manufactured by rapid solidification process are under developing to substitute the conventional PVD-Be coating. In the present study, brazed joint with PVD and amorphous alloys of $Zr_{1-x}Be_{x}(0.3{\le}x{\le}0.5)$ as filler metals are compared by mechanism, microstructure and hardness. The thickness of brazed joint with amorphous alloys became much smaller than that of PVD-Be. The erosion of base metal did not occur in the brazed joint with amorphous alloys. The brazing mechanism for PVD-Be seems to be Be diffusion into Zr-4 with capillary action resulting from eutectic reaction while that for amorphous alloys are associated with the liquid phase formation in the brazed joint. The brazed joint microstructure with PVD-Be consists of dendrite while that with amorphous alloys is globular. The $Zr_{0.7}Be_{0.3}$ alloy shows the smooth interface with little erosion in the base metal and is recommended a most suitable brazing filler metal for Zircaloy-4.

• Nuclear Engineering and Technology
• /
• v.42 no.3
• /
• pp.249-258
• /
• 2010

Recently, many utilities have considered interim dry storage of spent nuclear fuel as an option for increasing spent fuel storage capacity. Foreign nuclear regulatory committees have provided some regulatory and licensing requirements for relatively low- and medium-burned spent fuel with respect to the prevention of spent fuel degradation during transportation and interim dry storage. In the present study, the effect of cladding creep and hydride distribution on spent fuel degradation is reviewed and performance tests with high-burned Zircaloy-4 and advanced Zr alloy spent fuel are proposed to investigate the effect of burnup and cladding materials on the current regulatory and licensing requirements. Creep tests were also performed to investigate the effect of temperature and tensile hoop stress on hydride reorientation and subsequently to examine the temperature and stress limits against cladding material failure. It is found that the spent fuel failure is mainly caused by cladding creep rupture combined with mechanical strength degradation and hydride reorientation. Hydride reorientation from the circumferential to radial direction may reduce the critical stress intensity that accelerates radial crack propagation. The results of cladding creep tests at $400^{\circ}C$ and 130MPa hoop stress performed in this study indicate that hydride reorientation may occur between 2.6% to 7.0% strain in tube diameter with a hydrogen content range of 40-120ppm. Therefore, it is concluded that hydride re-orientation behaviour is strongly correlated with the cladding creep-induced strain, which varies as functions of temperature and stress acting on the cladding.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3379-3397
• /
• 2021

In this study, a multi-physics modeling method was developed to analyze a nuclear fuel rod's thermo-mechanical behavior especially for high temperature anisotropic creep deformation during ballooning and burst occurring in Loss of Coolant Accident (LOCA). Based on transient heat transfer and nonlinear mechanical analysis, the present work newly incorporated the nuclear fuel rod's special characteristics which include gap heat transfer, temperature and burnup dependent material properties, and especially for high temperature creep with material anisotropy. The proposed method was tested through various benchmark analyses and showed good agreements with analytical solutions. From the validation study with a cladding burst experiment which postulates the LOCA scenario, it was shown that the present development could predict the ballooning and burst behaviors accurately and showed the capability to predict anisotropic creep behavior during the LOCA. Moreover, in order to verify the anisotropic creep methodology proposed in this study, the comparison between modeling and experiment was made with isotropic material assumption. It was found that the present methodology with anisotropic creep could predict ballooning and burst more accurately and showed more realistic behavior of the cladding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.20 no.2
• /
• pp.161-170
• /
• 2022

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm^-1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm^-1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.