• Journal of the Korean Society for Heat Treatment
• /
• v.10 no.3
• /
• pp.198-208
• /
• 1997

The-steady-state creep mechanism and behavior of Zircaloy-4 used as cladding materials in PWR have been investigated in air environment over the temp, ranges from 600 to $645^{\circ}C$ and stress ranges from 4 to $7kg/mm^2$. The stress exponents for the creep deformation of this alloy, n were decreased 4.81, 4.71, 4.64, and 4.56 at 600, 615, 630 and $645^{\circ}C$, respectively; the stress exponents decreased with increasing the temperature and got closer to about 5. The apparent activation energies, Q, were 62.1, 60.0, 57.9 and 55.4 kcal/mole at stresses of 4, 5, 6, $7kg/mm^2$, respectively; the activation energies decreased with increasing the stress and were close to those of volume self diffusion of Zr in Zr-Sn-Fe-Cr system. In results, it can be considered that the creep deformation for Zircaloy-4 was controlled by the dislocation climb over the ranges of this experimental conditions. Larson-Miller parameter, P, for the crept specimens was obtained as P=(T+460)(logt,+23). The failure plane observed by SEM slightly showed up intergranular fracture at this experiment ranges. However, it was essentially dominated by the dimple phenomenon, which was a characteristics of the transgranular fracture.

• 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.

• Korean Journal of Materials Research
• /
• v.18 no.6
• /
• pp.326-333
• /
• 2008

Since the 1990s, the second generation of Zirconium alloys containing main alloy compositions of Nb, Sn and Fe have been used as a replacement of Zircaloy-4 (Zr-Sn-Fe-Cr), a first-generation Zirconium alloy, to meet severe and rigorous reactor operating conditions characterized by high-burn-up, high-power and high-pH operations. In this study, the mechanical properties and creep behaviors of Zr-Sn-Fe-Cr and Zr-Nb-Sn-Fe alloys were investigated in a temperature range of $450{\sim}500^{\circ}C$ and in a stress range of $80{\sim}150\;MPa$. The mechanical testing results indicate that the yield and tensile strengths of the Zr-Nb-Sn-Fe alloy are slightly higher compared to those of Zr-Sn-Fe-Cr. This can be explained by the second phase strengthening of the $\beta$-Nb precipitates. The creep test results indicate that the stress exponent for the steady-state creep rate decreases with the increase in the applied stress. However, the stress exponent of the Zr-Sn-Fe-Cr alloy is lower than that of the Zr-Nb-Sn-Fe alloy in a relatively high stress range, whereas the creep activation energy of the former is slightly higher than that of the latter. This can be explained by the dynamic deformation aging effect caused by the interaction of dislocations with Sn substitutional atoms. A higher Sn content leads to a lower stress exponent value and higher creep activation energy.

• Journal of the Korean Society for Precision Engineering
• /
• v.33 no.4
• /
• pp.279-286
• /
• 2016

To enhance the safety of nuclear reactors after the Fukushima accident, researchers are developing various types of accident tolerant fuel (ATF) to increase the coping time and reduce the generation of hydrogen by oxidation. Coated cladding, an ATF concept, can be a promising technology in view of its commercialization. We applied 3D printing technology to the fabrication of coated cladding as well as of coated pellets. Direct metal tooling (DMT) in 3D printing technologies can create a coated layer on the tubular cladding surface, which maintains stability during corrosion, creep, and wear in the reactor. A 3D laser coating apparatus was built, and parameter studies were carried out. To coat pellets with erbium using this apparatus, we undertook preliminary experiments involving metal pellets. The adhesion test showed that the coated layer can be maintained at near fracture strength.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.242-242
• /
• 2002

Two MIMAS MaX fuel rods base-irradiated in a commercial PWR have been reinstrumented and irradiated at a test reactor. The fabrication data for two MOX roda are characterized together with base irradiation information. Both Rods were reinstrumented to be fitted with thermocouple to measure centerline temperature of fuel. One rod was equipped with pressure transducer for rod internal pressure whereas the other with cladding elongation detector. The post irradiation examinations for various items were performed to determine fuel and cladding in-pile behavior after base irradiation. By using well characterized fabrication and re-instrumentation data and power history, the fuel performance code, COSMOS, is verified with measured in-pile and PIE information. The COMaS code shows good agreement for the cladding oxidation and creep, and fission gas release when compared with PIE dad a after base irradiaton. Based on the re-instrumention information and power history measured in-pile, the COSMOS predicts re-instrumented in-pile thermal behaviour during power up-ramp and steady operation with acceptable accuracy. The rod internal pressure is also well simulated by COSMOS code. Therfore, with all the other verification by COSMOS code up to now, it can be concluded that COSMOS fuel performance code is applicable for the design and license for MaX fuel rods up to high burnup.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05b
• /
• pp.175-180
• /
• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.614-620
• /
• 2020

In a previous study, delayed hydride cracking (DHC) assessment of pressurized water reactor (PWR) spent fuel during dry storage using the threshold stress intensity factor (K_IH) was performed. However, there were a few limitations in the analysis of the cladding properties, such as oxide thickness and mechanical properties. In this study, those models were modified to include test data for irradiated materials, and the cladding creep model was introduced to improve the reliability of the DHC assessment. In this study, DHC susceptibility of PWR spent fuel during dry storage depending on the axial elevation was evaluated with the improved assessment methodology. In addition, the sensitivity of affecting parameters such as fuel burnup, hydride thickness, and crack aspect ratio are presented.

• Journal of the Korean Society of Fisheries and Ocean Technology
• /
• v.28 no.1
• /
• pp.71-78
• /
• 1992

Zr-4 used for a cladding and an end plug of reactor component has creep deformation under operation at high temperature. Creep is regarded as the time dependent deformation of a material under constant applied stress. Although the major source of the deformation of zirconium component in water-cooled reactors is irradiation creep, the thermal creep may give a rise to significant deformation in reactor component especially at relatively high temperatures and at various constant stresses, and therefore it must be predicted accurately. Stress relaxation is the time dependent change of stress at constant strain and it is a process related intimately to creep. In this paper, the creep behavior and stress relaxation of Zr-4 is examined at the temperature of 50$0^{\circ}C$ that is 40% of the absolute melting temperature of Zr-4 under the stress below yield stress and under the various constant strains. The results obtained are summarized as follows: 1) With an increase of stress, the steady state creep rate increases and the creep rupture time decreases. 2) The steady state creep rate $\varepsilon$(%/s) for the stress $\sigma$sub(c) (kgf/mm super(2)) of Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 increases outstandingly. All the empirical equations computed for Zr-4 are in accord with Norton's model equation($\varepsilon$=K$\sigma$ sub(c) super (n)). The constants of materials computed are as follows: K=3.9881$\times$10 super(-5), n=1.9608 3) The rupture time T sub(r) (hr) decreases linearly with the increase of stress on the log-log scaled graph. The empirical equations computed for Zr-4 are in accord with Bailey's model equation (T sub(r)=K sub(1)$\sigma$sub(c) super(m)). The constants of materials computed are as follows: K sub(1)=1.2875$\times$10 super(16), m=-3.467 4) It seems clear that the strain could be quantitatively dependent on the high temperature creep properties such as creep stress, rupture time, steady state creep rate and total creep rate. It is found that these relationships are linear on the log-log graph. 5) In stress relaxation test, as the critical constant strain that can be allowed to the specimen is larger, stress relaxation becomes more rapid, and as the constant strain is smaller, the stress relaxation becomes slower.

• Journal of the Korea Concrete Institute
• /
• v.12 no.4
• /
• pp.59-67
• /
• 2000

With increased height of structure, the effect of column shortening need special consideration in the design and construction of high-rise buildings. The shortening of each column affects nonstructural members such as partitions, cladding, and M/E systems, which are not designed to carry gravity forces. The slabs and beams will tilt due to the cumulative differential shortening of adeacent vertical members. The main purpose of estimating the total shortening of vertical structural member is to compensate the differential shortening between adeacent members. This paper presents effect of parameters for phenomenon of column shortening in vertical members. The paper presents effect of parameters for phenomenon of column shortening in vertical members. The conclusions obtained from this study are follow as ; Strength of concrete and steel ratio effected on column shortening caused by elastic and inelastic shortening. Also, it is known that Ultimate-shrinkage-Value, Specific-Creep-Value, and volume to surface ratio effected on inelastic shortening only. Particularly, Ultimate-Shrinkage-Value and Specific-Creep-Value effected considerable on the amount of total column shortening.

• Proceedings of the KWS Conference
• /
• 2004.05a
• /
• pp.123-125
• /
• 2004

Flat specimens of polycrystal Ni-base superalloy with U-shape notch on both sides were laser cladded by injection of IN738LC powder onto surface. The quality of cladding was investigated by microstructural analysis and high temperature tension test, creep test at 950$^{\circ}C$. Effects of heat treahnent and the angle between the tension axis and the direction of weldment were also investigated.