• Journal of the Korean Society of Safety
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• v.30 no.1
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• pp.144-149
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• 2015

This study aims to quantitatively evaluate failure pressure of wall-thinned elbow under combined load along with internal pressure, by conducting real-scale burst test and finite element analysis together. For quantitative evaluation, failure pressure data was extracted from the real-scale burst test first, and then finite element analysis was carried out to compare with the test result. For the test, the wall-thinning defect of the extrados or intrados inside the center of 90-degree elbow was considered and the loading modes to open or close the specimen maintaining a certain load or displacement were applied. Internal pressure was applied until failure occurred. As a result, when the bending load was applied under the load control condition, the intrados of the defect was more affected by failure pressure than the extrados, and the opening mode was more vulnerable to failure pressure than the closing mode. When the bending load was applied under the displacement control, it was hardly affected by failure pressure though it was slightly different from the defect position. The result of the finite element analysis showed a similar aspect with the test. Moreover, when major factors such as material properties and pipeline thickness were calibrated to accurate values, the analytical results was more similar to the test results.

• Journal of the Korean Society of Safety
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• v.30 no.4
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• pp.168-173
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• 2015

A steam generator tubing is one of the main pressure boundary of the reactor coolant system in the nuclear power plants. Structural integrity refers to maintaining adequate margins against failure of the tubing. Burst pressure of a tube at tube support plate can be higher than that for a free-span tube because failure behaviors could be interfered from the tube support plate. Alternative repair criteria for out-diameter stress corrosion cracking indications in tubes to the drilled type tube support plate were developed, however, there are very limited information to the eggcrate type tube support plate. This paper discussed reinforcement effect of steam generator tube burst pressure with axial out-diameter stress corrosion cracking within an eggcrate type tube support plate. A series of tube burst tests were performed under the room temperature and it was found out that there is no significant but marginal effects.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.10
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• pp.1043-1047
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• 2012

The European nuclear design requirements, which should be satisfied by nuclear reactors in Europe, usually recommend a so-called core catcher, which is a molten core ex-vessel cooling facility, to manage a severe accident at a nuclear reactor. Two different types of core catcher concepts are compared to determine their abilities to manage severe accidents and cool core melts. The study reveals that direct cooling is better for cooling capacity and is convenient to construct, while indirect cooing is better for the management of a severe accident.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.11-15
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• 2008

In order to transport spent resin in a high integrated container made of high density polyethylene, a method for determining transportation grade by radioactivity analysis was developed. Ratios of radioisotopes in spent resin were derived from radioactivity analysis on spent resin. Associated curie-to-dose factors were determined to estimate radioisotope inventory from surface dose rates of spent resin. From the results, Activity limit of type A package was derived to be 1.19 TBq for HIC, and the corresponding surface dose rate was found to be 124.2 mSv/h.

• Journal of Energy Engineering
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• v.26 no.2
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• pp.50-63
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• 2017

Safety culture degradation signs in nuclear power plants with complex and diverse systems can lead to their equipments performance deterioration. If these signs are neglected, they become potential causes of accidents. Therefore, it is necessary to monitor safety culture in the point of view of organization and management as well as to evaluate safety performance of nuclear power plants. Therefore, This paper suggested a methodology to evaluate safety culture weakness contributing the accidents' root causes in the case accidents occur at nuclear power plants. After reviewing methodologies using at domestic and international industry, the methodology suitable for domestic nuclear power plants was determined.

• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.123-130
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• 2015

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1231-1242
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• 2020

The Kori Unit 1 will be decommissioned after a permanent shutdown in June 2017. South Korea has a 0.1 mSv/yr exposure limit standard for limited or unlimited site release. This is South Korea's first commercial NPP; therefore, if the containment building is reused as a memorial hall, it will contribute to the improvement of public understanding and enhance the public's acceptance of NPPs. Also, existing Kori Unit 1 nuclear power plant manpower resources can be reused after decommissioning and resident staff and memorial hall visitors can activate nearby commercial areas. Therefore, such a reuse scenario may also prevent an economic recession. The exposure dose was calculated using the following scenarios: worker in the containment building, visitor in the containment building, and worker in buildings other than the containment building. The exposure dose in the buildings was calculated by the RESRAD-BUILD developed by the Argonne National Laboratory (ANL). The preliminary exposure dose and derived concentration guideline level (DCGL) were derived.

• Journal of Radiation Protection and Research
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• v.36 no.2
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• pp.99-106
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• 2011

Tritium is the one of the most important radionuclide for workers in nuclear power plants (NPPs) and the public, from the dosimetric point of view. Humans are likely to have internal radiation exposure by tritium inhalation. Radiation exposure by tritium accounts for approximately 7% and 60~90% of the total radiation exposure of NPP workers and the public during normal operation, respectively. Thus, many researches have been conducted to estimate the internal dose by tritium precisely in the world. In terms of tritium dosimetry, this paper provides the current status of research for tritium metabolism and dosimetry.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.26-32
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• 2012

The systems of the nuclear power plant are designed based on the User Requirement Document, and Korea Hydro & Nuclear Power Co. (KHNP) implements preventive maintenance activities to keep the specific design function of the system consistently. To monitor the preventive maintenance effectiveness, KHNP has also developed maintenance effectiveness monitoring (MR) program based on NUMARC 93-01 since 2003, and has implemented the program in all operating plants. Recently, KHNP has upgraded MR programs by reflecting implementing experiences ; reestablishing the performance monitoring level, improving analysis for standby function and performance criteria for passive components, reestablishing the availability performance criteria and the performance criteria for the same type of components. These upgraded MR programs will contribute to enhance safety and improve equipment reliability through monitoring maintenance effectiveness.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.23 no.1
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• pp.49-55
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• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.