Results 1 - 10 of 12
Results 1 - 10 of 12. Search took: 0.034 seconds
|Sort by: date | relevance|
[en] As a result, severe failures occur in the piping and equipment of nuclear power plants (NPPs). Numerous parameters such as the geometry, pH, flow velocity, steam quality, dissolved oxygen (DO), temperature, and materials have an influence on FAC. This paper describes the effect of chromium content in material on FAC of carbon steel at 150 .deg. C. Since the Surry Unit 2 accident in 1986, flow accelerated corrosion (FAC) has been extensively studied. However, many accidents caused by FAC have been reported such as an accident at Mihama Unit 3 in 2004 and at domestic plants. During the FAC, a protective oxide layer on carbon steel dissolves into flowing water leading to a thinning of the oxide layer and accelerating corrosion of base material
[en] Since the Surry Unit 2 accident in 1986, flow accelerated corrosion (FAC) has been extensively studied. However, many accidents caused by FAC have been reported such as an accident at Mihama Unit 3 in 2004 and at domestic plants. During the FAC, a protective oxide layer on carbon steel dissolves into flowing water leading to a thinning of the oxide layer and accelerating corrosion of the base material. As a result, severe failures occur in the piping and equipment of nuclear power plants (NPPs). Numerous parameters such as the geometry, pH, flow velocity, steam quality, dissolved oxygen (DO), temperature, and materials have an influence on FAC. This paper describes the effect of chromium content on the FAC of carbon steel at 150 .deg. C
[en] In this paper, the status of the test facility, the basic concept of the ASME QME-1 test, and the future work are presented. The measured performance parameter, such as pressure, temperature, flow rate, thrust, stroke, and time, is used for the maintenance method and the evaluation of the aging and fatigue in the operating valve. In these days, the survival of the safety related valve in severe accident is important issue. In Korea Institute of Machinery and Materials (KIMM), the functional facility and related Quality Assure (QA) system are made. In addition, the qualification research was made for the functional test of the valve as ASME QME-1. The methodology is based on the ASME QME-1 1997. But the future work for the ASME QME-1 2007 was performed. In real situation, the engineering work is needed for the functional test of the nuclear valve in QA system. So the research work for the qualification is needed. The functional qualification of the valve in the nuclear power plant is performed in KIMM The development of the research and facility required in the functional qualification is in progress. Now the functional qualification is possible domestically, that was impossible in previous time due to the lack of qualification technology and facility
[en] Radiation technology is closely related to the industrial growth and the creation of employment in Korea. The techniques as radiation or/and radioactivity measurement, and the practical skills achieving a higher level analysis are required. In this study, practice manual for liquid scintillation counter were developed by job analysis. Raw data applied in job analysis are collected by on/off line survey by 420 workers employed in KOREA. Importance-priority analysis was performed to make duties and competency unit that consists of knowledge, skills as each task. Refined data was reviewed by expert who experienced actual duties on site. Classification was conducted by focus group interview to deduct duties and competency unit. From the radiation devices in measurement and analysis, liquid scintillation counter was preferentially selected because of the high demands for training. Investigation of build-up status to liquid scintillation counter in KOREA was conducted. Then technical specification and operating procedure of 2 main devices were analyzed and integrated by practice manual. Duties and competency unit were applied to integrated materials respectively. To validate effectiveness, test curriculum was designed by the advanced course to workers who engaged in radiation measurement and analysis. The developed manual is structured to take advantage of test training. This manual will be a practical handbook that can improve the knowledge, skills of radiation workers in Korea. (paper)
[en] After the Surry Unit 2 accident in 1986, flow accelerated corrosion (FAC) has been extensively studied. However, many accidents caused by FAC have been reported, such as the accident at Mihama Unit 3 in 2004 and at domestic plants. The mechanism of FAC is that a protective oxide layer on carbon steel dissolves into the flowing water, leading to a thinning of the oxide layer. As a result, severe failures occur in piping and equipment in nuclear power plants (NPPs). Numerous parameters such as geometry, pH, flow velocity, steam quality, dissolved oxygen (DO), temperature, and materials have an influence on FAC. This paper describes the effect of dissolved oxygen on FAC of carbon steel at 150 .deg. C.
[en] In this study, the axial-compressor-design and performance/flow analysis program is developed. A mean-line analysis was used to determine optimum arrangement of overall geometry and its off-design performance is predicted by stage-stacking method. Three dimensional blade shape is generated using radial equilibrium equation and vortex methods. Various blade shape is generated and their performance is compared. Finally the quasi-three dimensional flow analysis is applied to investigate the detailed flow phenomena
[en] The final goal of this study is to enhance the reliability of safety assessment of a HLW repository through the development of safety case based on KURT environment. We secured the bases for the risk-based safety assessment by making element analysis and developing the integrated analysis method for complex scenarios, making multiple source term analysis, and upgrading the existing safety assessment program. We built the foundation for enhancing the reliability of safety assessment of a HLW repository through safety case analysis of each stage of waste disposal, establishment of safety assessment framework, and making a safety case report. The impact analysis of temperature and metal ions on the radionuclide migration, the impact of corrosion products and gas in the boundary of waste package and buffer materials on the radionuclide migration, and the effects of microbes on the sorption behavior of uranium onto bentonite and granite were analyzed through experimental studies. We developed the TRLFS to develop methodologies for quantifying uranium contents in groundwater. By using TRLFS, we identified the chemical species of uranium in groundwater and analyzed the oxidation-reduction reaction of uranium in KURT groundwater. We also developed radionuclide transport model considering geochemical reaction and long-term geochemical reaction model of uranium in KURT groundwater. They can be used to support the development of safety case for the HLW repository. The results of this study are expected to contribute the reliability enhancement of safety assessment of a HLW repository through being used in the development of safety case.
[en] The final goal of this project is to improve the confidence in disposal safety assessment of high-level waste through KURT-based Safety Case development. In order to achieve this goal, in the Phase I, a previous safety assessment code is updated and then a risk-based safety assessment system is obtained by development of models for multiple source term evaluation and complex exposure analysis. Besides, the data for the complex behaviors of radionuclides necessary for disposal safety assessment are obtained and analyzed by experimental researches on the complex behaviors of radionuclides in the interfaces of engineered barrier system such as container, buffer, and rock. Long-term behavior of radionuclide in natural barrier is analyzed and evaluated using natural uranium existing in the KURT groundwater and then contribute to the improvement of confidence in disposal safety. In the Phase II, In order to improve the confidence in disposal safety through Safety Case development based on the assumed repository, KURT, and A-KRS disposal system for pyro-processing waste, complementary safety indicators and their evaluation models are developed and an assessment system for the Safety Case is also developed. Besides, the behaviors of radionuclides in the interface of natural barrier are investigated and their data are provided for the Safety Case development. The confidence in disposal safety assessment is also improved by evaluating long-term behaviors of uranium in the fractured rock of KURT. Our project published a Safety Case (AKRS-16) report for conceptual disposal system for pyro-processing high-level waste based on the KURT site by using the research results. The results obtained from our project will be directly applied to the disposal system development for radioactive wastes from recycling process, the development of a total performance assessment system, and the implementation of national programs for the management of high-level waste. We expect that our results are also contribute to the confidence improvement of public peoples in disposal safety of high-level waste.
[en] Flow-Induced Vibration of steam generator tubes may result in fretting wear damage at the tube-to-support locations. KSNP(Korean Standard Nuclear Power plant) steam generators experienced fretting wear in the upper part of U-bend above the central cavity region of steam generators. This region has conditions susceptible to the flow-induced vibration, such as high flow velocity, high void fraction, and longer unsupported span. To improve its performance, APR1400 steam generator is designed with additional supports in this region to reduce unsupported span and to reduce peak velocity in the central cavity region. In this paper, we examined its performance improvement using ATHOS code. The thermal-hydraulic condition in the region of secondary side of APR1400 steam generator is obtained using the ATHOS3 code. The effective mass for modal analysis is calculated using the void fraction, enthalpy, and operating pressure information from ATHOS3 code result. With the effective mass distribution along the tube, natural frequency and mode shape is obtained using ANSYS code. Finally, stability ratios and real mean squared displacements for selected tubes of the APR1400 steam generator are computed with and without the improvement. From these results, the effectiveness of the improvement and the current design criteria are examined
[en] This study was to investigate the effect of conduction characteristics between rows of the fin-tube heat exchanger. Experiments were performed for the 7mm tube diameter heat exchangers using air-enthalpy type calorimeter. Fin patterns of the heat exchangers were slit and louver types. Equivalent fin spacing with 18FPI was used for all samples, and the number of tube rows were 2. In order to investigate the conduction effects, one sample was physically separated between two rows, but the other was connected. The air velocity was varied from 0.7 to 2.5 m/s with 0.3 m/s interval. Heat transfer for each row are evaluated. It was observed that conduction effects between rows on the overall heat transfer performance was considerable and have to be considered for determination of heat transfer coefficients for individual row