• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.1
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• pp.63-68
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• 2013

Some essential components in nuclear power plants are periodically inspected using non-destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in-service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non-magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Journal of the Earthquake Engineering Society of Korea
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• v.9 no.6 s.46
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• pp.13-19
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• 2005

Floor response spectra for dynamic response of subsystem such as equipment, or piping in nuclear power plants are usually generated without considering dynamic interaction between main structure and subsystem. This study describes the analytic method in which equipment response spectra can be obtained through dynamic analysis considering equipment-structure Interaction(ESI). In this method, dynamic response of the equipment by this method is based on a dynamic substructure method in which the equipment-structure system is partitioned into the single-degree-ol-freedom system(SDOF) representing the equipment and the equipment support impedance representing the dynamic charactenstics of the structure ai the equipment support. A family of equipment response spectra is developed by applying this method to calculate the maximum responses of a family of SDOF equipment systems with wide banded equipment frequency, damping ratio, and mass. The method is validated by comparing the floor response spectrum from this method with the floor response spectrum generated from the rigorous analysis including equipments on the containment building of a prototypical nuclear power plant. in order to Investigate ESI effect in the response of equipment, response values from the method and the conventional approach without considering ESI are compared for the equipment having the mass less than 1% of the total structural mass. Response spectra from the method showed lower spectral amplitudes than those of the conventional floor response spectra around controlling frequencies.