• Korean Journal of Materials Research
• /
• v.12 no.4
• /
• pp.259-263
• /
• 2002

Three different ribbons of amorphous $Til_{1-x}Be_x$ alloys such as $Ti_{0.59}Be_{0.41},\;Ti_{0.61}Be_{0.39}\;and\;Ti_{0.63}Be_{0.37}$ were made by melt-spinning method to be used as brazing filler metals for joining Zircaloy-4 nuclear fuel cladding tubes, and their crystallization behavior as well as microstructure of the brazed zone were examined. The crystallization behavior was investigated in teams of thermal stability, crystallization temperature and activation energy. The crystallization of the $Ti_{1-x}Be_x$ alloys proceeded in two steps by the formation of ${\alpha}$-Ti at a lower temperature and of TiBe at a higher temperature. The crystallization temperature and activation energy of $Ti_{1-x}Be_x$ alloys were higher and larger than those of $Zr_{1-x}Be_x$ alloys and PVD Be. Those resulted thinner joining layer with $Ti_{1-x}Be_x$ alloys, which kept sound thickness of Zircaloy-4 nuclear fuel cladding tubes after brazing. But in the brazed zones made by $Ti_{1-x}Be_x$ filler metals, a little solid-solution layers composed of Zr and Ti were formed toward the Zr cladding tube and Zr was detected in the brazed zones. Microstructure of brazed zone was changed from globular to dentrite with decreasing Be content in the $Ti_{1-x}Be_x$ filler metal.

• Nuclear Engineering and Technology
• /
• v.8 no.1
• /
• pp.19-27
• /
• 1976

The strain ageing behaviour of Zircaloy-4 has been studied in the temperature range 175$^{\circ}C$ to 575$^{\circ}C$ for both quenched and annealed specimens. The strain ageing in quenched Zircaloy-4 was found in the temperature range 175$^{\circ}C$ to 50$0^{\circ}C$ and its Peak occured at 3$25^{\circ}C$ while the strain ageing in annealed specimens occured in the temperature range 175-575$^{\circ}C$, showing two peaks, one at 323$^{\circ}C$ and a higher one at 45$0^{\circ}C$. The peak at 3$25^{\circ}C$ in both quenched and annealed specimens is considered to be due to the segregation of interstitial oxygen atoms to cell walls during ageing. The peak at 45$0^{\circ}C$ in annealed specimens is considered to he due to the interaction of dislocations with Fe atoms. It has been found that strain ageing stress at ~30$0^{\circ}C$ in zirconium alloys is proportional to the square root of oxygen content.

• Nuclear Engineering and Technology
• /
• v.19 no.2
• /
• pp.122-129
• /
• 1987

Experiments and numerical analysis have been performed to investigate the effect of preoxidation by oxidizing Zircaloy-4 specimens at a higher temperature after a period of exposure at a lower temperature. The oxidation experiments were performed between $700^{\circ}C$ and 85$0^{\circ}C$ after Preoxidation at $650^{\circ}C$ in a steam and water mixture for 600 seconds and 1,800 seconds. As the thickness of preoxidized layer increased, the oxidation rate of preoxidized specimens at higher temperature became lower than that of as-received claddings. A transition region of oxidation rate exist in the preoxidized specimens, and the region disappeared rapidly as the oxidation temperature increased. This effect appeared more clearly at lower temperatures. According to the results of numerical analysis performed in this study, the growth rate of oxide layer thickness and weight gains were similarly affected by the thickness of preoxidized layer.

• 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.54 no.5
• /
• pp.1549-1559
• /
• 2022

As an accident tolerant fuel (ATF) concept, oxide dispersion strengthened Zircaloy-4 (ODS Zry-4) cladding has been developed to enhance the mechanical properties of cladding using laser processing technology. In this study, a simulation technique was established to investigate the mechanical properties and effects of Y₂O₃ particles for the ODS Zry-4. A 3D representative volume element (RVE) model was developed considering the parameters of the size, shape, distribution and volume fraction (VF) of the Y₂O₃ particles. From the 3D RVE model, the Young's modulus, coefficient of thermal expansion (CTE) and creep strain rate of the ODS Zry-4 were effectively calculated. It was observed that the VF of Y₂O₃ particles had a significant effect on the aforementioned mechanical properties. In addition, the predicted properties of ODS Zry-4 were applied to a simulation model to investigate cladding deformation under a transient condition. The ODS Zry-4 cladding showed better performance, such as a delay in large deformation compared to Zry-4 cladding, which was also found experimentally. Accordingly, it is expected that the simulation approach developed here can be efficiently employed to predict more properties and to provide useful information with which to improve ODS Zry-4.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.1
• /
• pp.69-75
• /
• 2013

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls was demonstrated to investigate the effect of the oxidation process on the reaction rate of the chlorination reaction. In chlorination reaction experiments performed for 6 h, where reaction products were collected every 2 h, it was observed that a significant decrease in the reaction rate was caused by the oxidation process ($500^{\circ}C$, 10 h under an air atmosphere) within the reaction period of 0 - 2 h. The amount of reaction residue increased from 0.95 to 1.65wt% of initial weights in the fresh and Zry-500-10 (Zry-4 hulls oxidized at $500^{\circ}C$ for 10 h under an air atmosphere) hulls, respectively. The purity of the recovered Zr was identical at 99.61wt% for the fresh Zry-4 and Zry-500-10 hulls. Quantitative analysis of the chlorination reaction rate was performed by varying the reaction time from 0.5 to 1.0, 2.0, and 4.0 h. The fitting results showed that the relationship between weight loss and reaction time can be interpreted by a linear line with a slope of 23.35wt%/h for the fresh Zry-4 case, while two linear lines were necessary to fit the results of Zry-500-10. In addition, the slope values were 17.12 and 27.16wt%/h for (0 - 20) and (20 - 100)wt% loss regions, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.1
• /
• pp.55-61
• /
• 2013

The recovery of Zr from Zircaloy-4 (Zry-4) cladding hulls using a chlorination method was demonstrated for complete conversion of Zr into $ZrCl_4$. A chlorination reaction was performed by reacting Zry-4 hulls for 8 h under a 70 cc/min $Cl_2$ + 70 cc/min Ar flow at $380^{\circ}C$. The initial weight of the reactant (51.7 g) decreased to 0.49 g after 8 h of operation, which is only 0.95wt% of the initial weight. The weight of the total reaction products was 121.7 g with a high Zr purity of 99.80wt%. Fe and Sn were identified as major (0.18wt%) and minor (0.02wt%) impurities of the reaction products, respectively. It was also shown that Zr exhibited a high recovery ratio of 96.95wt% with a relatively small experimental loss of 2.34wt%. Observation of the reaction residues revealed that the chlorination reaction was dominant along the longitudinal direction, and surface oxide layers remained as reaction residues. The high purity and recovery ratio of Zr proposed the feasibility of the chlorination technique as an effective hull waste treatment method.

• Nuclear Engineering and Technology
• /
• v.23 no.1
• /
• pp.1-11
• /
• 1991

The objective of this study is to investigate the effect of weld line on the mechanical strengths and the process of weld line elimination in the Zircaloy-4 resistance upset welding for the fabrication of heavy water reactor fuel rods. The weld current and the amount of upset increased linearly with the main heat, in which two relations between them were derived. It was found that the threshold to obtain sound weld was 50% of main heat in terms of weld upset size, mechanical strengths and weld line elimination. The weld microstructure of resistance upset welds of Zircaloy-4 comprsied basketweave, Widmanstatten and martensite respectively by changing the main heats. Dimples on uniaxially fractured surface at weld line in the Zr-4 welds were larger and deeper compared with those on biaxially fractured surface. It was also found that the process of the weld line elimination in the resistance upset weld of Zircaloy-4 could be divided into three stages in terms of the presence of many pores, their shrinkage and elimination, and the shrinkage of the original weld interface with increasing weld currents.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.780-781
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.931-932
• /
• 2004