Journal Search Engine
View PDF Download PDF Export Citation Korean Bibliography PMC Previewer
ISSN : 1738-1894(Print)
ISSN : 2288-5471(Online)
Journal of Nuclear Fuel Cycle and Waste Technology Vol.20 No.2 pp.161-170
DOI : https://doi.org/10.7733/jnfcwt.2022.013

Effect of Steady-State Oxidation on Tensile Failure of Zircaloy Cladding

Taeho Kim1, Kyoung Joon Choi2, Seung Chang Yoo3, Yunju Lee3, Ji Hyun Kim3*
1Korea Atomic Energy Research Institute, 111, Daedeok‑daero 989beon‑gil, Yuseong‑gu, Daejeon 34057, Republic of Korea
2Korea Railroad Research Institute, 176, Cheoldobangmulgwan-ro, Uiwang-si, Gyeonggi-do 16105, Republic of Korea
3Ulsan National Institute of Science and Technology, 50, UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan 44919, Republic of Korea
* Corresponding Author.
Ji Hyun Kim, Ulsan National Institute of Science and Technology, E-mail: kimjh@unist.ac.kr, Tel: +82-52-217-2913

March 28, 2022 ; May 11, 2022 ; June 17, 2022

Abstract


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.



초록


    1. Introduction

    Numerous types of Zr alloys have been utilized as nuclear fuel cladding materials for pressurized water reactors for several years due to their exceptional mechanical properties and low thermal neutron capture cross-section. Because cladding materials shield radioactive materials such as uranium and fission products, the degradation of zirconium fuel cladding is directly related to the safety of nuclear power plants as well as spent nuclear fuel. Additionally, the mechanical properties of the zirconium alloy are determined by the zirconium oxide that forms on its surface. Moreover, the phase of zirconium oxide is probably correlated with its structure [1-16]. To ensure the safety of spent nuclear fuel and nuclear power plants, it is, therefore, necessary to fully comprehend the mechanical property and oxidation characteristics of zirconium fuel cladding. Investigations utilizing Raman spectroscopy and transmission electron microscopy (TEM) revealed that monoclinic and tetragonal phases of zirconium oxide exist [17]. The tetragonal zirconium oxide predominates in regions where the macroscopic compressive stress is typically higher near the oxide/metal interface than it is in the interior zirconium oxide layer. As the oxidation process continues, the proportion of monoclinic zirconium oxide increases while the proportion of tetragonal zirconium oxide decreases [18]. By decreasing the proportion of tetragonal phase zirconium oxide, properties such as corrosion resistance and mechanical properties may be altered. The transformation from the tetragonal phase to the monoclinic phase is an intrinsic property of zirconium oxide, and many previous results indicate that its phase transformation to the monoclinic zirconium oxide phase correlates with the mechanical properties of zirconium alloys [19, 20]. The majority of previous zirconium oxide-focused literature relies on ex-situ experiments [21]; however, removing the specimen from the oxidation environment could influence the structure and phase of the oxide. To investigate oxides, in-situ Raman spectroscopy was utilized in this study. In addition, a tube tensile test and a rupture test were conducted to investigate the changes in mechanical behavior brought on by oxidation. To characterize the fractography of oxidized zirconium alloy after a tensile test, an ex-situ investigation involving scanning electron microscopy (SEM) was carried out.

    2. Experimental

    2.1 Materials and Specimen Preparation

    This study utilized a plate and tube of the zirconium alloy [4, 16, 22], and the chemical composition of zirconium alloy is presented in Table 1.

    Prior to the oxidation experiment, the plate zirconium alloy specimen for in-situ Raman specimens had dimensions of 40 × 40 × 0.65 mm3 and was meticulously polished. To first grind the specimen, 400 to 800 grit SiC paper was utilized. Next, diamond pastes with a maximum particle size of 1 μm and colloidal SiO2 were utilized to reduce the mechanical transformation of specimens. As specimens for tube rupture and tube tensile tests, zirconium alloy tubes were utilized. The outer diameter, length, and thickness of tubes are respectively 9.5, 130, and 0.6 mm. The tube was then pressurized to approximately 4.5 MPa prior to the oxidation test, and this pressure was maintained during oxidation. The schematic of the zirconium tube specimen for oxidation and tube rupture test was illustrated in Fig. 1. This procedure was intended to simulate the internal pressure of the fuel cladding in a pressurized water reactor. For the tube tensile test, the type 2 reduced method of ASTM E8-E8M were used to design the tensile specimen. Fig. 2 explains the construction of a tensile specimen.

    2.2 Experimental System

    For simulating the primary water chemistry of a nuclear power plant, particularly a pressurized water reactor, an autoclave and high-temperature and -pressure condition loop were created. The explanation is elaborated upon in the authors’ previous study [16]. During the oxidation experiment, the following conditions of the oxidation environment were maintained: temperature of 360°C, the pressure of 19 MPa, dissolved oxygen (DO) concentration of less than 5 ppb, and LiOH and H3BO3 concentrations of 2 and 1,200 ppm, respectively. During the oxidation, a dissolved hydrogen (DH) concentration of 30 cm3·kg−1 was also measured.

    2.3 Experimental Procedure

    In this study, oxidized Zr-Nb-Sn specimens were subjected to in-situ Raman spectroscopy, tube rupture testing, and tube tensile testing in a primary water environment. Plate and tube specimens of Zr-Nb-Sn were corroded in primary water chemistry conditions for a maximum of 100 d. The in-situ Raman spectra were measured 30, 50, 80, and 100 d after the start of oxidation. The detailed explanation of the whole system used for the in-situ Raman spectroscopic analysis is explained the authors’ previous studies [16, 23-25]. After 300 h, 50 d, and 100 d of oxidation initiation, the rupture and tensile properties of the tubes were measured. Using SEM, the microstructure characteristics of oxidized specimens were determined by analyzing the tensile test fractography.

    The tube rupture test was conducted until a high-pressure water-filled zirconium tube specimen ruptured. Using an Instron 8801 universal testing machine, as shown in Fig. 3, at a strain rate of 0.015 s−1 and a proportionally reduced specimen in accordance with ASTM E8-E8m, tensile tests were conducted.

    3. Results and Discussion

    3.1 Mechanical Tests With Oxidized Tubes

    In this study, rupture and tensile tests were conducted on oxidized tubes after 300 h, 50 d, and 100 d of exposure to high-temperature water. Fig. 4 depicts the tube rupture experiment system, and Fig. 5 illustrate the results of a rupture test on a tube. The black line depicts the internal pressure of the as-received specimen as a function of time, while the red, blue, and magenta lines depict the internal pressure of 300 h, 50 d, and 100 d oxidation specimens, respectively. As illustrated in Fig. 5, the rupture pressure remains constant as oxidation time increases. The maximum oxidation time is 100 d, which is insufficient to influence the rupture pressure, which is heavily influenced by the oxide and metal matrix thickness [3].

    Fig. 6 displays the results of a tube tensile test performed at room temperature. Table 2 summarizes the yield strength, tensile strength, and elongation data from tensile tests. As shown in Fig. 6 and Table 2, yield and tensile strengths increased slightly. However, the 300-h oxidized sample had a different elongation than the sample as received. After 50 d of oxidation, however, the yield and tensile strength increased. The yield tensile strengths slightly increased to 753 and 1,002 MPa, respectively, as the oxidation time increased to 50 d.

    When the oxidation time attained 100 d, the yield and tensile strengths decreased to 555 and 706 MPa, respectively, due to softening. To investigate the relationship between mechanical properties and microstructure, SEM was used for fractographic analysis. Figs. 7 depict the fractography of an alloy of oxidized zirconium. As shown in Figs. 7, the surface morphologies of the as-received specimen and the 300-h oxidized specimen were identical. The size of the dimples in both SEM images was approximately 5–10 μm.

    As depicted in Fig. 8(a), after 50 d of oxidation, the dimple size increased to approximately 30–40 μm, resulting in an increase in brittleness and yield tensile strength. In addition, fractography revealed smooth grains, which could increase the brittleness of the specimen [26]. In Fig. 8(b), which depicts the 100 d oxidation specimen, the grain boundary fracture leads to the softening of the zirconium alloy. In addition, Fig. 8(b) contains lots of cracks on the fractography, with length up to 50 μm.

    3.2 In-situ Raman Spectroscopic Analysis for Zr-Nb-Sn Alloy

    Fig. 9 [27] depicts the in-situ Raman spectra of the zirconium alloy plate specimen oxidized for 30, 50, 80, and 100 d at a constant concentration of 30 cm3·kg−1 dissolved hydrogen. Increasing the signal-to-noise ratio and obtaining more accurate Raman spectra [16, 23, 24] were accomplished using the various techniques described in previous research. The high-intensity peaks that correspond to the sapphire window and boric acid are depicted in Fig. 9 as a vertical black bar. Positioned at the tip of the optical probes, the sapphire window determines the Raman peak positions at 414, 430, 575, and 747 cm−1 [16, 28]. The peaks between 868 and 870 cm−1 are associated with the boric acid dissolved in the primary water [16]. Tetragonal zirconium oxide phase peaks could be observed at 330 and 380 cm−1 after 30 d of oxidation [15, 16]. At 50 d of oxidation, the tetragonal zirconium oxide peak at 330 cm−1 merged with the monoclinic zirconium oxide peak. Furthermore, the peak at 380 cm−1 for tetragonal zirconium oxide disappeared. Other monoclinic peaks 475, 530–561 (broad), 616, and 637 cm−1 remained until 100 d of oxidation [15, 16].

    This indicates that in-situ Raman spectroscopy can be used to investigate the phase of zirconium oxide. At an early stage of oxidation, the tetragonal phase of zirconium oxide may be close to the oxide/metal interface. Nonetheless, as oxidation time increased, the stress relaxation due to OH ion could contribute to the tetragonal to monoclinic zirconium oxide phase transformation, resulting in an increase in the monoclinic zirconium oxide ratio [16].

    3.3 Ex-situ Analysis for Zirconium Oxide Using TEM

    After milling with a focused ion beam (FIB), TEM analysis was conducted to determine the phase of zirconium oxide. Before FIB, the specimen surface was coated with carbon to prevent surface contamination and ion beam damage to the sample itself. Fig. 10 [27] is a bright-field TEM micrograph of the oxide/metal interface of an oxidized zirconium alloy. After 100 d of oxidation, the thickness of zirconium oxide on the specimen was 2.44 μm. To investigate the zirconium oxide phase at different positions within the oxide and to confirm the zirconium oxide phase, the fast Fourier transform (FFT) method, which represents the grain shape of the oxide, was utilized. In Fig. 10, position “A” represents the columnar grain, while position “B” represents the equiaxed grain. These two positions were used for the FFT analysis. Position “A” was analyzed as the monoclinic zirconium oxide phase, whereas position “B” was analyzed as the tetragonal zirconium oxide phase, as indicated by the FFT analysis results. This indicates that the zirconium oxide phase transformation from oxide surface to metal/oxide interface has occurred.

    3.4 Relationship Between Zirconium Oxide Phase and Mechanical Property

    Using in-situ Raman spectroscopy, the oxide phase of zirconium alloy was analyzed. Phase transformation occurred as the oxidation time increased. The tetragonal-tomonoclinic transformation occurred at the oxide-to-oxideto- metal interface, and the volume expansion associated with the transformation caused intergranular cracks in oxidized zirconium alloys [29-31]. It is known that monoclinic and tetragonal zirconium oxides have distinct grain sizes. The monoclinic zirconium oxide is primarily columnar and has a length of approximately 500 nm, whereas the tetragonal zirconium oxide is primarily equiaxed and has a radius of approximately 30 nm [9, 10]. This relationship between oxygen vacancies and water molecules was explained by the previous zirconium degrading mechanism. In the degradation mechanism [31], hydroxyl ions enter the inner portion of the material through grain boundary diffusion and fill oxygen vacancies. The formation of proton defects and a decrease in the concentration of oxygen vacancy leads to the transformation from tetragonal to monoclinic because the tetragonal phase is no longer stable.

    During this degradation, the grain boundaries are susceptible to hydroxyl ion attacks, which can explain intergranular cracking, and the oxygen vacancy concentration in the space charge layer rises as grain size decreases [31, 32]. In other words, the zirconium degradation might slow down when the grain size is relatively small or fine. When the oxidation time is short, a greater proportion of tetragonal zirconium oxide is present in the zirconium oxide, which explains the relatively high mechanical property. However, as oxidation time increases, the tetragonal zirconium oxide transforms into monoclinic zirconium oxide, and the material degrades due to hydroxyl ion attacks. This can lead to grain boundary cracking, as depicted in Fig. 8, and it softens zirconium alloy. As oxidation time increases, the yield and tensile strengths of zirconium alloy decrease.

    4. Conclusion

    Using in-situ Raman spectroscopy, tube rupture tests, and tube tensile tests, the oxide characteristics and mechanical properties of oxidized spent fuel cladding material, zirconium alloy, were investigated in this study. Both tetragonal and zirconium oxide phases exist in the early stages of oxidation. As oxidation time increases, the monoclinic zirconium oxide phase may become predominant, leading to the degradation and softening of zirconium alloy. The observation from this study can be applied to the basic understanding of mechanical properties of cladding materials with various oxidation conditions. The following conclusions can be drawn from this study:

    1. Utilizing a tube rupture and a tensile test, the mechanical properties of the spent fuel cladding material were analyzed. As the oxidation time increases, the rupture pressure remains constant. Nonetheless, as oxidation time increases, yield and tensile strengths diminish.

    2. As oxidation time increases, the monoclinic phase of zirconium oxide becomes dominant, as determined by in-situ Raman spectroscopy. The tetragonal phase zirconium oxide peaks are visible in the 30-d oxidation Raman spectrum, but they merge or disappear as the oxidation time increases.

    3. According to TEM and FFT analyses, the oxide phase distribution differs from the grain position in the oxide. Near the oxide/metal interface, tetragonal zirconium oxide exists, while monoclinic zirconium oxide exists in the middle of zirconium oxide.

    4. The transformation of tetragonal zirconium oxide to monoclinic zirconium oxide increases the proportion of monoclinic zirconium oxide. The diminished tetragonal ratio may account for the decline in zirconium alloy mechanical properties.

    Acknowledgements

    This work was supported by the Nuclear Safety Research Program through the Korea Foundation Of Nuclear Safety (KoFONS) using the financial resource granted by the Nuclear Safety and Security Commission (NSSC) of the Republic of Korea (No. 2106022), and the KAERI Institutional Project (Project No. 522320-22).

    Figures

    JNFCWT-20-2-161_F1.gif

    Schematic of the fuel cladding tube specimen.

    JNFCWT-20-2-161_F2.gif

    Schematic and image of the fuel cladding tensile test specimen.

    JNFCWT-20-2-161_F3.gif

    Instron 8801 universal testing machine for this study. The tube tensile sample is held by grips. The samples after the tensile test are also presented.

    JNFCWT-20-2-161_F4.gif

    Images of tube burst experiment. Burst tubes are also presented.

    JNFCWT-20-2-161_F5.gif

    Results of the oxidized fuel cladding tube rupture test.

    JNFCWT-20-2-161_F6.gif

    Results of the oxidized fuel cladding tube tensile test.

    JNFCWT-20-2-161_F7.gif

    (a) Fractography of the as-received fuel cladding specimen, (b) fractography of the 300-h oxidized fuel cladding tube specimen.

    JNFCWT-20-2-161_F8.gif

    (a) Fractography of the 50-d oxidized fuel cladding specimen, (b) fractography of the 100-d oxidized fuel cladding specimen.

    JNFCWT-20-2-161_F9.gif

    In-situ Raman spectra of oxidized plate specimen at different oxidation times, 30, 50, 80, and 100 d [27].

    JNFCWT-20-2-161_F10.gif

    Bright field TEM micrograph of the cross section of the 100-d oxidized specimen [27].

    Tables

    Chemical composition of the fuel cladding alloy (Zr-Nb-Sn)

    Results of the oxidized fuel cladding tube tensile test

    References

    1. W. Qin, C. Nam, H.L. Li, J.A. Szpunar, “Tetragonal Phase Stability in ZrO2 Film Formed on Zirconium Alloys and its Effects on Corrosion Resistance”, Acta Mater., 55(5), 1695-1701 (2007).
    2. E. Polatidis, P. Frankel, J. Wei, M. Klaus, R.J. Comstock, A. Ambard, S. Lyon, R.A. Cottis, and M. Preuss, “Residual Stresses and Tetragonal Phase Fraction Characterisation of Corrosion Tested Zircaloy-4 Using Energy Dispersive Synchrotron X-ray Diffraction”, J. Nucl. Mater., 432(1-3), 102-112 (2013).
    3. P. Platt, S. Wedge, P. Frankel, M. Gass, R. Howells, and M. Preuss, “A Study Into the Impact of Interface Roughness Development on Mechanical Degradation of Oxides Formed on Zirconium Alloys”, J. Nucl. Mater., 459, 166-174 (2015).
    4. B. de Gabory, A.T. Motta, and K. Wang, “Transmission Electron Microscopy Characterization of Zircaloy-4 and ZIRLOTM Oxide Layers”, J. Nucl. Mater., 456, 272- 280 (2015).
    5. X. He, H. Yu, G. Jiang, G. Dang, D. Wu, and Y. Zhang, “Cladding Oxidation Model Development Based on Diffusion Equations and a Simulation of the Monoclinic- Tetragonal Phase Transformation of Zirconia During Transient Oxidation”, J. Nucl. Mater., 451(1-3), 55-65 (2014).
    6. A. Couet, A.T. Motta, and R.J. Comstock, “Hydrogen Pickup Measurements in Zirconium Alloys: Relation to Oxidation Kinetics”, J. Nucl. Mater., 451(1-3), 1-13 (2014).
    7. N. Ni, D. Hudson, J. Wei, P. Wang, S. Lozano-Perez, G.D.W. Smith, J.M. Sykes, S.S. Yardley, K.L. Moore, S. Lyon, R. Cottis, M. Preuss, and C.R.M. Grovenor, “How the Crystallography and Nanoscale Chemistry of the Metal/oxide Interface Develops During the Aqueous Oxidation of Zirconium Cladding Alloys”, Acta Mater., 60(20), 7132-7149 (2012).
    8. H.X. Zhang, D. Fruchart, E.K. Hlil, L. Ortega, Z.K. Li, J.J. Zhang, J. Sun, and L. Zhou, “Crystal Structure, Corrosion Kinetics of New Zirconium Alloys and Residual Stress Analysis of Oxide Films”, J. Nucl. Mater., 396(1), 65-70 (2010).
    9. A. Yilmazbayhan, A.T. Motta, R.J. Comstock, G.P. Sabol, B. Lai, and Z. Cai, “Structure of Zirconium Alloy Oxides Formed in Pure Water Studied With Synchrotron Radiation and Optical Microscopy: Relation to Corrosion Rate”, J. Nucl. Mater., 324(1), 6-22 (2004).
    10. A. Yilmazbayhan, E. Breval, A.T. Motta, and R.J. Comstock, “Transmission Electron Microscopy Examination of Oxide Layers Formed on Zr Alloys”, J. Nucl. Mater., 349(3), 265-281 (2006).
    11. B. Cox, “Some Thoughts on the Mechanisms of In-reactor Corrosion of Zirconium Alloys”, J. Nucl. Mater., 336(2-3), 331-368 (2005).
    12. P. Bouvier, J. Godlewski, G. and Lucazeau, “A Raman Study of the Nanocrystallite Size Effect on the Pressure- Temperature Phase Diagram of Zirconia Grown by Zirconium-based Alloys Oxidation”, J. Nucl. Mater., 300(2-3), 118-126 (2002).
    13. M. Oskarsson, E. Ahlberg, and K. Pettersson, “Oxidation of Zircaloy-2 and Zircaloy-4 in Water and Lithiated Water at 360°C”, J. Nucl. Mater., 295(1), 97-108 (2001).
    14. S. Gou, B. Zhou, C. Chen, M. Yao, J. Peng, X. Liang, J. Zhang, and Q. Li, “Investigation of Oxide Layers Formed on Zircaloy-4 Coarse-grained Specimens Corroded at 360°C in Lithiated Aqueous Solution”, Corros. Sci., 92, 237-244 (2015).
    15. J.E. Maslar, W.S. Hurst, W.J. Bowers, and J.H. Hendricks, “In Situ Raman Spectroscopic Investigation of Zirconium-Niobium Alloy Corrosion Under Hydrothermal Conditions”, J. Nucl. Mater., 298(3), 239-247 (2001).
    16. T. Kim, J. Kim, K.J. Choi, S.C. Yoo, S. Kim, and J.H. Kim, “Phase Transformation of Oxide Film in Zirconium Alloy in High Temperature Hydrogenated Water”, Corros. Sci., 99, 134-144 (2015).
    17. V.Y. Gertsman, Y.P. Lin, A.P. Zhily, and J.A. Szpunar, “Special Grain Boundaries in Zirconia Corrosion Films”, Philos. Mag. A, 79(7), 1567-1590 (1999).
    18. J. Lin, H. Li, C. Nam, and J.A. Szpunar, “Analysis on Volume Fraction and Crystal Orientation Relationship of Monoclinic and Tetragonal Oxide Grown on Zr- 2.5Nb Alloy”, J. Nucl. Mater., 334(2-3), 200-206 (2004).
    19. Y. Ding and D.O. Northwood, “TEM Study of the Oxide- Metal Interface Formed During Corrosion of Zr- 2.5wt.%Nb Pressure Tubing”, Mater. Charact., 30(1), 13-22 (1993).
    20. H. Anada and K. Takeda, “Microstructure of Oxides on Zirclaloy-4, 1.0Nb Zircaloy-4, and Zircaloy-2 Formed in 10.3- MPa Steam at 673 K”, in: Zirconium in the Nuclear Industry: Eleventh International Symposium, E.R. Bradley, G.P Sabol, eds., 35-54, STP 1295, ASTM International, Pennsylvania (1996).
    21. H.S. Hong, S.J. Kim, and K.S. Lee, “Long-term Oxidation Characteristics of Oxygen-added Modified Zircaloy- 4 in 360°C Water”, J. Nucl. Mater., 273(2), 177- 181 (1999).
    22. N. Ni, S. Lozano-Perez, J.M. Sykes, G.D.W. Smith, and C.R.M. Grovenor, “Focused Ion Beam Sectioning for the 3D Characterisation of Cracking in Oxide Scales Formed on Commercial ZIRLOTM Alloys During Corrosion in High Temperature Pressurised Water”, Corros. Sci., 53(12), 4073-4083 (2011).
    23. J. Kim, K.J. Choi, C.B. Bahn, and J.H. Kim, “In Situ Raman Spectroscopic Analysis of Surface Oxide Films on Ni-base Alloy/Low Alloy Steel Dissimilar Metal Weld Interfaces in High-temperature Water”, J. Nucl. Mater., 449(1-3), 181-187 (2014).
    24. J. Kim, S.H. Kim, K.J. Choi, C.B. Bahn, I.S. Hwang, and J.H. Kim, “In-situ Investigation of Thermal Aging Effect on Oxide Formation in Ni-base Alloy/Low Alloy Steel Dissimilar Metal Weld Interfaces”, Corros. Sci., 86, 295-303 (2014).
    25. J.H. Kim and I.S. Hwang, “Development of an in Situ Raman Spectroscopic System for Surface Oxide Films on Metals and Alloys in High Temperature Water”, Nucl. Eng. Des., 235(9), 1029-1040 (2005).
    26. H. Bethge and J. Heydenreich, “Electron Microscopy in Solid State Physics”, VEB Deutscher Verlag der Wissenschaften, Germany (1982).
    27. T. Kim, K.J. Choi, S.C. Yoo, Y. Lee, and J.H. Kim, “Influence of Dissolved Hydrogen on the Early Stage Corrosion Behavior of Zirconium Alloys in Simulated Light Water Reactor Coolant Conditions”, Corros. Sci., 131, 235-244 (2018).
    28. M. Kadleíková, J. Breza, and M. Veselý, “Raman Spectra of Synthetic Sapphire”, Microelectron. J., 32 (12), 955-958 (2001).
    29. X. Guo, “Hydrothermal Degradation Mechanism of Tetragonal Zirconia”, J. Mater. Sci., 36(15), 3737-3744 (2001).
    30. J.K. Lee and H. Kim, “Surface Crack Initiation in 2YTZP Ceramics by Low Temperature Aging”, Ceram. Int., 20(6), 413-418 (1994).
    31. X. Guo, “Property Degradation of Tetragonal Zirconia Induced by Low-Temperature Defect Reaction With Water Molecules”, Chem. Mater., 16(21), 3988-3994 (2004).
    32. X. Guo and Z. Zhang, “Grain Size Dependent Grain Boundary Defect Structure: Case of Doped Zirconia”, Acta Mater., 51(9), 2539-2547 (2003).
    1. SEARCH
    2. JNFCWT

      Online Submission

    3. Korean Radioactive
      Waste Society (KRS)

    4. Editorial Office
      Contact Information

      - Tel: +82-42-861-5851, 866-4205
      - Fax: +82-42-861-5852
      - E-mail: krs@krs.or.kr