• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.25 no.6
• /
• pp.965-971
• /
• 2001

To estimate the reliability of 12Cr alloy steel, the material of turbine blade in the steam power plant, Its corrosion susceptibility and fatigue characteristics in NaCl and Na$_2$SO$_4$solution with the difference of concentration and temperature was investigated. The polarization tests recommended in ASTM G5 standard for corrosion susceptibility in the various corrosive solutions was estimated. It showed that the higher temperature, the faster corrosion rates and corrosion rates were the fastest in 3.5 wt.% NaCl and 1M Na$_2$SO$_4$solution. From these results, the degradation conditions were determined in distilled water, 3.5 wt.% NaCl and 1M Na$_2$SO$_4$solution at room temperature, 60$\^{C}$ and 90$\^{C}$ during 3, 6 and 9 months. Its surface had a few pits for long duration. But, it was not susceptible in sulfide and chloride condition of several temperatures. If the degraded 12Cr alloy steel and non-degraded one were compared with fatigue characteristics, Any differences were not found regardless of temperature and degradation period.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2004.11a
• /
• pp.311-316
• /
• 2004

For the location estimation in the conventional LPMS(Loose Parts Monitoring System), it is popular to employ a group delay among the acoustic sensors installed within a 3 ft range from the impact source. However, there exists inherent error in determining the arrival time differences of the generated wave group among the neighboring sensors. To overcome this problem in this study, the two dimensional approach has been proposed and applied to effectively estimate the arrival time differences by using a continuous wavelet transform which is one of the linear time-frequency analysis methods. The experiment has been performed to both the plate model and the real steam generator in a nuclear power plant. It is expected that the reliability of the location estimation could be enhanced when the proposed time-frequency method is introduced into the LPMS system.

• Journal of Radiation Industry
• /
• v.11 no.3
• /
• pp.123-130
• /
• 2017

Concerns about high radiation exposure to the hands of radiation workers who may contact with radioactive contamination on surfaces in a nuclear power plant (NPP) had been raised, and the Korean regulatory body required the extremity dose estimation during contact tasks with radioactive materials. Korean NPPs conducted field tests to identify the incident radiation to the hands of radiation workers who may contact with radioactive contamination during maintenance periods. The results showed that the radiation fields for contact tasks are dominated by high energy photons. It was also found that the radiation doses to the hands of radiation workers in Korean NPPs were much less than the annual dose limits for extremities. This approach can be applicable to measure and estimate the extremity dose to the hands of medical workers who handle the radioactive materials in a hospital.

• Nuclear Engineering and Technology
• /
• v.51 no.2
• /
• pp.402-409
• /
• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Journal of the Korean Society of Manufacturing Process Engineers
• /
• v.20 no.7
• /
• pp.89-96
• /
• 2021

In this study, a mini hydro cyclone was designed and manufactured to achieve an inlet flow rate of 2 L/min in the experiment, which was conducted using alumina powder with a specific gravity of 3.97. This hydro cyclone was studied for using in steam and water analysis system (SWAS) of thermal power plant and was manufactured by 3D printing. Numerical analysis was performed with Solidworks Flow Simulation, utilizing the reynolds stress method (RSM) of fluid multiphase flow analysis models. Experimental and numerical analysis were performed under the three conditions of inlet velocity 2.0, 4.0, and 6.0 m/s. The separation efficiency was over 80% at all inlet velocity conditions. At the inlet velocity 4m/s, the separation efficiency was the best, and it was confirmed that the efficiency was more than 90%.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.38 no.6
• /
• pp.655-662
• /
• 2014

This study conducted tensile tests on an Alloy 690TT tube at room temperature (RT) and at $343^{\circ}C$ using tube- and ring-type specimens to investigate the stress-strain behavior and tensile properties of a steam generator (SG) tube in the axial and circumferential directions at RT and at the design temperature of a nuclear power plant (NPP). The results of the axial tensile test showed that yield point phenomena appeared at both RT and $343^{\circ}C$, and serrated flow in the stress-strain curve appeared at $343^{\circ}C$. Yield and tensile strengths for both directions were clearly lower at $343^{\circ}C$ compared to RT; however, the elongations were approximately the same at both test temperatures. Regardless of the test temperature, the strengths in the circumferential direction were lower by approximately 5~10 % than those in the axial direction. In addition, the test data revealed that the reduction in the yield and tensile strengths of the Alloy 690TT SG tube with the test temperature was more significant than that estimated by the temperature correction factor of ASME Sec.II.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.19 no.8
• /
• pp.44-54
• /
• 2018

Stress and fatigue of the distributor, an equipment of the high-pressure evaporator for the HRSG, were evaluated according to ASME Boiler & Pressure Vessel Code Section VIII Division 2. First, from the results of the piping system analysis model, reaction forces of the tubes connected to the distributor were derived and used as the nozzle load applied to the detailed analysis model of the distributor afterward. Next, the detailed model to analyze the distributor was constructed, the distributor being statically analyzed for the design condition with the steam pressure and the nozzle load. As a result, the maximum stress occurred at the bore of the horizontal nozzle, and the primary membrane stress at the shell and nozzle was found to be less than the allowable. Next, for the transient operating conditions given for the distributor, thermal analysis was performed and the structural analysis was carried out with the steam pressure, nozzle load, and thermal load. Under the transient conditions, the maximum stress occurred at the vertical downcomer nozzle, and of which fatigue life was evaluated. As a result, the cumulative usage factor was less than the allowable and hence the distributor was found to be safe from fatigue failure.

• Journal of Radiation Protection and Research
• /
• v.24 no.4
• /
• pp.215-223
• /
• 1999

Characteristics of radiation field in the steam generator(S/G) water chamber of a PWR were investigated and the anticipated effective dose rates to the worker in the S/G chamber were evaluated by Monte Carlo simulation. The results of crud analysis in the S/G of the Kori nuclear power plant unit 1 were adopted for the source term. The MCNP4A code was used with the MIRD type anthropomorphic sex-specific mathematical phantoms for the calculation of effective doses. The radiation field intensity is dominated by downward rays, from the U-tube region, having approximate cosine distribution with respect to the polar angle. The effective dose rates to adults of nominal body size and of small body size(The phantom for a 15 year-old person was applied for this purpose) appeared to be 36.22 and 37.06 $mSvh^{-1}$) respectively, which implies that the body size effect is negligible. Meanwhile, the equivalent dose rates at three representative positions corresponding to head, chest and lower abdomen of the phantom, calculated using the estimated exposure rates, the energy spectrum and the conversion coefficients given in ICRU47, were 118, 71 and 57 $mSvh^{-1}$, respectively. This implies that the deep dose equivalent or the effective dose obtained from the personal dosimeter reading would be the over-estimate the effective dose by about two times. This justifies, with possible under- or over- response of the dosimeters to radiation of slant incidence, necessity of very careful planning and interpretation for the dosimetry of workers exposed to a non-regular radiation field of high intensity.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.30 no.4
• /
• pp.386-395
• /
• 2010

The steam generator management program(SGMP) has recently defined the procedures for the qualification of eddy current hardware and technique. These procedures provide two basic methods for qualification. The first way is to qualify the equipment or the probe by using the flaw mechanism and method of the pulled tubes from the heat exchangers or the artificial flawed tubes. The second way is to verify the equivalency with the characteristics of the qualified equipment or probe. In this case, the qualified equipment or probe may be modified to substitute or replace instruments or probes without re-qualification provided that the range of essential variables defined in the examination technique specification sheet are met. This study is to describe the result of the comparative performance evaluation of bobbin coil eddy current probes manufactured by KEPCO Research Institute and probes manufactured by a foreign manufacturer. As a result of this study, although there were minor differences between the two kinds of probes, it was evaluated that the two kinds of probes were almost identical in the significant performance characteristics described in the KEPCO Research Institute guideline.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.24 no.4
• /
• pp.384-389
• /
• 2004

As the increase of the operation year of nuclear power plants, the probabilities of the degradation of the major facilities and materials in the nuclear power plants are increased. The integrity of those facilities shall be monitored and verified by the non-destructive examination methods with the regulation codes, so called inservice inspection(ISI). The ISI of Yonggwang unit 6 was performed in four different parts, 1) non-destructive examinations for the components, piping weldments and structures, 2) automated ultrasonic examinations for pressure vessels, 3) visual examinations for the interior structures of the reactor, 4) eddy current examinations for the steam generator tubes. As the results, there was no severe indication and all detected indications were evaluated as non-relavent. Especially for the examinations of the piping weldments, PD(Performance Demonstration) was applied as a W examination method defined in the 1995 edition of ASME Code Sec. XI. The implementation of the PD for the piping weld results in an improvement of the reliability of the UT examinations.