• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.44-50
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• 2017

The shaft seal of the reactor coolant pump is installed on the upper side of the rotating shaft of the pump to seal the reactor coolant from flowing out between the rotating shaft and the non-rotating parts. In this study, the loading conditions for the normal operation and faulted conditions are identified and structural integrity evaluation is performed using the finite element stress analysis for the sealing apparatus of the APR 1400 reactor coolant pump. It is confirmed that the stress analysis results satisfy the design criteria at all loading conditions.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.169-179
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• 2000

A detailed numerical analysis of the evolution of thermal stratification in a curved piping system in a nuclear power plant is performed. A finite volume based thermal-hydraulic computer code has been developed employing a body-fitted, non-orthogonal curvilinear coordinate for this purpose. The cell-centered, non-staggered grid arrangement is adopted and the resulting checkerboard pressure oscillation is prevented by the application of momentum interpolation method. The SIMPLE algorithm is employed for the pressure and velocity coupling, and the convection terms are approximated by a higher-order bounded scheme. The thermal-hydraulic computer code developed in the present study has been applied to the analysis of thermal stratification in a curved duct and some of the predicted results are compared with the available experimental data. It is shown that the predicted results agree fairly well with the experimental measurements and the transient formation of thermal stratification in a curved duct is also well predicted.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.54-60
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• 2011

A spacer grid, which supports nuclear fuel rods laterally and vertically with a friction grip, is one of the most important structural components in a PWR fuel. The form of grid strap and supporting parts such as grid spring and dimple is known to be closely related with the mechanical/structural performance of spacer grid and nuclear fuel assembly. In this study, reviewing various research results for enhancing the performance of the spacer grid, some structural design considerations and research directions on the spacer grid assembly are suggested for further study.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.48-53
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• 2011

A 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. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. In this study, as a part of the evaluation on the high-temperature structural integrity of the small-scale PHE prototype, we carried out macroscopic high-temperature structural analysis of the small-scale PHE prototype under the gas loop test conditions considering the pipeline stiffness.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.30-39
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• 2011

URANUS (Ubiquitous, Robust, Accident-forgiving, Non-proliferating, Ultra-lasting and Sustainer) has been developed with 35MWe (100MWth) operating without primary coolant pump, capitalizing on natural circulation capability of lead-bismuth eutectic (LBE) for long-life small and robust power units. To ensure the structural integrity, the large safety margin against Safe Shutdown Earthquake, 0.3g, and furthermore the cost effectiveness for URANUS, three-dimensional seismic base isolation design has been developed. The analytical model has been developed and seismic time history analyses have been carried out. The advantage for using three-dimensional seismic base isolation for URANUS has been discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.47-53
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• 2010

In this research project, development of methodology to measure the thickness of pipes in the wide range using magnetic field. The magnetic field spreading in the sensor and the plate was modeled in the cases of the various thicknesses in plate. Based on the analysis, sensors were designed, manufactured and tested to optimize the specifications of the sensor. The sensor can be used in high temperature through calibration. And the uncertainty of the sensor was estimated.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.42-46
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• 2010

The evaluation method for the failure frequency of reactor vessel under pressurized thermal shock(PTS) is developed using probabilistic fracture mechanics. The probabilistic reactor integrity evaluation code, named R-PIE code, is developed. The validity and uncertainty of the R-PIE code is investigated. The reactor failure frequencies under PTS for Kori-1 nuclear power plant and other type of domestic nuclear power plants are evaluated. The reference PTS temperature for domestic nuclear power plants is obtained for the rule making against PTS failure.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.16-19
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• 2010

In this study, the axial ODSCC occurrence of domestic OPR 1000 steam generator tube was caused by the tube weakness and the sludge accumulation in the secondary side of steam generator. Inconel 600 HTMA used as tube material is related to most of tube leakage accidents in the world and also these ODSCCs were detected mainly at the 5th TSP(Tube Support Plate) to the 8th TSP of hot leg side. These elevations(5th TSP to 8th TSP) pave the way for the sludge accumulation. As a result of EC(Eddy Current) Bobbin and RPC data analysis, ODSCCs were occurred at contact points of tube and tube support plate. The more accumulated sludge, the higher occurrence frequency of ODSCC.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.9-15
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• 2010

US NRC issued rulemaking of 10CFR50.55a to perform the Perservice and Inservice inspection for Reactor Vessel Head Penetration Nozzle of US Nuclaer plant. The rulemaking was required the EPRI Demonstration to verify the NDE technique performing special Ultrasonic examination. In order to meet this requirement, the UT and ECT procedures was demonstrated and the NDE personnel were qualified by EPRI. In this paper, the NDE technique and analysis method are described the Preservice inspection for the Palo Verde #1/2/3 Replacement Reactor Vessel Head Penetration Nozzle using the qualified procedures and personnel.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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• pp.1-15
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• 2017

After Kyeongju earthquake occurred in September 12, 2016, the seismic safety of nuclear power plants became important issue in our country. The seismic safety of nuclear power plant against beyond design basis earthquake became very important to secure the public safety. In this paper, the current status of the seismic safety assessment methodology is reviewed and some aspects for the reliability improvement of the seismic safety assessment results are proposed. Seismic margin analysis and probabilistic seismic safety assessment have been used for the seismic safety evaluation of a nuclear power pant. The basic procedure and the related issues and proposals for the probabilistic seismic safety assessment are investigated.