• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.48-55
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• 2013

In this paper, W-DCAP-HTR(Web-based Design Compatibility Assessment Program for High Temperature Reactor) which will be used to check the design criteria for high temperature reactor is newly proposed. To do this, the assessment procedure of the ASME Sec.III Div.5 such as time-dependent primary stress limit, accumulated inelastic strain, and creep-fatigue damage evaluation were investigated. Furthermore, the trend of candidate materials for high temperature reactor was also reviewed. Then, all assessment procedures for high temperature reactor have been computerized to enhance the efficiency and to reduce the possibility of human error during calculating procedure by hand calculation. It can be directly conducted by adopting the actual thermal and structural analysis results. The validation of W-DCAP-HTR has been demonstrated by benchmark analysis.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3571-3580
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• 2023

The application of the induction bending process to pipe systems in various industrial fields is increasing. Recently, efforts have also been made to apply this bending process to nuclear power plants because it can innovatively reduce welded parts of the curved pipes, such as elbows. However, there have been no cases of the application of induction bending to the piping of nuclear power plants. In this study, the applicability of the P91 induction bending piping for the sodium-cooled fast reactor PGSFR was validated through high temperature low cycle fatigue tests and creep tests using P91 induction bending pipe specimens. The tests confirmed that the materials sufficiently satisfied the fatigue life and the creep rupture life requirements for P91 steel at 550 ℃ in the ASME B&PV Code, Sec. III, Div. 5. The results show that the effects of heating and bending by the induction bending process on the material properties were not significant and the induction bending process could be applicable to piping system of PGSFR well.

• Journal of the Korean Institute of Gas
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• v.17 no.2
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• pp.50-58
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• 2013

Various process piping usually carries out high flammable and explosible gas under high pressure and high temperature. Due to frequent change of design and structure it becomes more complicated and compactly located. The safety management level is relatively low since it is considered as simply designed component. In this study a safety assessment procedure is proposed for complicated piping system around a mixing drum in which natural gas and by-product gases were mixed. According to ASME code, pipe stress analysis was conducted for determining design margin at some key locations of the piping. These high stress locations can be used as major inspection points for managing the pipe integrity. Sensitivity analysis with outside temperature of the pipe and support constraint condition. Possible effect of hydroen gas to the pipe steel during the previous use of the by-product gas was also discussed.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.5
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• pp.381-388
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• 2009

In some portions of nuclear piping systems, stratification phenomena may occur due to the density difference between hot and cold stream. When the temperature difference is large, the stratified flow under diverse operating conditions can produce high thermal stress, which leads to unanticipated piping integrity issues. The objectives of this research are to examine controvertible numerical factors such as model size, grid resolution, turbulent parameters, governing equation, inflow direction and pipe wall. Parametric three-dimensional computational fluid dynamics analyses were carried out to quantify effects of these parameters on the accuracy of temperature profiles in a typical nuclear piping with complex geometries. Then, as a key finding, it was recommended to use optimized mesh of real piping with the conjugated heat transfer condition for accurate thermal stratification analyses.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.2
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• pp.1-6
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• 2011

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established a small-scale gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype made of Hastelloy-X to be tested in the small-scale gas loop. Results from the elastic structural analysis on the PHE prototype were reported in the previous article. In order to investigate the macroscopic structural characteristics and behavior of the PHE prototype under the test condition of the small-scale gas loop far more in detail, elastic-plastic high-temperature structural-analysis of the PHE prototype was carried out in this study.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.15-23
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• 2018

A compact heat exchanger is one of critical components in a very high temperature gas-cooled reactor (VHTR). Alloy 617 (Ni-Cr-Co-Mo) is considered as one of leading candidates for this application due to its excellent thermal stability and strengths in anticipated operating conditions. On the basis of current ASME code requirements, sixty sheets of this alloy are prepared for diffusion welding, which is the key technology to have a reliable compact heat exchanger. Optical microscopic analysis show that there are no cracks, incomplete bond, and porosity at/near the interface of diffusion weldment, but Cr-rich carbides and Al-rich oxides are identified through high resolution electron microscopic analysis. In high-temperature tensile testing, superior yield strengths of the diffusion weldment compared to the code requirement are obtained up to 1223 K ($950^{\circ}C$). However, both tensile strength and ductility drop rapidly at higher temperature due to the insufficient grain boundary migration across the interface of diffusion weldment. Best fit curves for minimum yield strength and average tensile strength are drawn from the experimental tensile results of this study.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.1
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• pp.33-38
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute has established a small-scale gas loop for the performance test on VHTR components and recently has manufactured a medium-scale PHE prototype made of Hastelloy-X. A performance test on the PHE prototype is scheduled in the gas loop. In this study, high-temperature structural analysis modeling, and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints in the previous research were carried out under the gas loop test condition. The results obtained in this study will be compared with performance test results.

• Journal of the Earthquake Engineering Society of Korea
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• v.26 no.1
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• pp.39-48
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• 2022

Many piping systems installed in the power plant are directly related to the safety and operation of the plant. Various dampers have been applied to the piping system to reduce the damage caused by earthquakes. In order to reduce the vibration of the piping system, this study developed a steel coil damper (SCD) with a straightforward structure but excellent damping performance. SCD reduces the vibration of the objective structure by hysteretic damping. The new SCD damper can be applied to high-temperature environments since it consists of steel members. The paper introduces a design method for the elastoplastic coil spring, which is the critical element of SCD. The practical applicability of the design procedure was validated by comparing the nonlinear force-displacement curves calculated by design equations with the results obtained from nonlinear finite element analysis and repeated loading test. It was found that the designed SCD's have a damping ratio higher than 25%. In addition, this study performed a set of seismic tests using a shaking table with an existing piping system to verify the vibration control capacity on the piping system by SCD. Test results prove that the SCD can effectively control the displacement vibration of the piping system up to 80%.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.44-50
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• 2011

Turbulent mixing of hot and cold coolants is one of the possible causes of high cycle thermal fatigue in piping systems of nuclear power plants. A typical situation for such mixing appears in turbulent flow through a T-junction. Since the high cycle thermal fatigue caused by thermal striping was not considered in the piping fatigue design in several nuclear power plants, it is very important to evaluate the effect of thermal striping on the integrity of mixing tees. In the present work, before conducting detailed evaluation, three thermal striping evaluation methodology suggested by EPRI, JSME and NESC are analyzed. Then, a by-pass pipe connected to the shutdown cooling system heat exchanger is investigated by using these evaluation methodology. Consequently, the resulting thermal stresses and the fatigue life of the mixing tee are reviewed and compared to each other. Futhermore, the limitation of each methodology are also presented in this paper.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.326.1-326.1
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• 2004