• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.3
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• pp.231-235
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• 2016

There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.40 no.9
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• pp.821-826
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• 2016

The leak before break (LBB) concept is used in pipe line design for nuclear power plants. For application of the LBB concept, leak rates through cracks should be evaluated accurately. Usually leak late analyses are performed for through-thickness cracks with constant cross-sectional area. However, the cross-sectional area at the inner pipe surface of a crack can be different from that at the outer surface. In this paper, leak rate analyses are performed for the cracks with linearly-varying cross-sectional areas. The effect of varying the cross-sectional area on leak rates was examined. Leak rates were also evaluated for cracks in bi-material pipes. Finally, the effects of crack surface morphology parameters on leak rates were examined.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.609-613
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• 2003

The confinement of a storage system to accommodate spent fuel and radioactive material must be designed and estimated so that the storage system is safe during a storage period. The confinement can be estimated by calculation of the allowable leakage rate in compliance with ANSI Nl4.5 or ISO 12807. Accordingly, the allowable leakage rate was estimated in compliance with ISO 12807 in the case of storage of 24 PWR spent fuels and 24-conditioned spent fuels. In the case of the 24 PWR spent fuel assemblies, the allowable leakage rate was estimated as $7.43{\times}10_{-11}m_3/s$, in the case of the 24 conditioned spent fuel assemblies, the allowable leakage rate was estimated as $1.80{\times}10_{-10}m_3/s$. Therefore, the confinement condition in the storage of the conditioned spent fuel is easier than that for the storage of the PWR spent fuel.

• Journal of the Korean Professional Engineers Association
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• v.31 no.1
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• pp.17-28
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• 1998

누설 검출기들 중에서 검출 감도가 가장 민감한 것은 '헬륨 양이온 질량 분석계 누설 검출기(줄인 이름, '헬륨 누설검출기')'이다. 이 글은 헬륨 누설 검출기의 주요 부분을 이루고 있는 질량 분석계의 측정 원리를 요약하여 소개한 것이다. 누설 검사의 항목들은 검사의 필요에 따라 대개 시험 물체에서 누설의 크기와 시험 방법을 바탕으로 하여 대규모 누설, 총량 누설, 소규모 누설 그리고 미세 누설 등으로 나뉘어지며, 미세 누설들은 누설 검출기의 최소 검출 가능 누출률이 $10^{-8}$ std ㎤ s$^{-1}$ 보다 작은 범위에 있는 것들로서, 그들의 누설 시험은 헬륨 누설 검출기를 써서 하게 된다. 따라서, 헬륨 누설 검출기의 심장은 바로 질량 분석계이며, 질량 분석계의 원리를 이해하는 것은 곧, 헬륨 누설 검출기를 다루는 기술의 바탕을 다지는 것이라고 말할 수 있을 것이다. 여기서는 헬륨 질량 분석계의 원리를 쉽게 풀이하고, 그런 검출기를 써서 누설검사를 하는 데 바탕이 될 모습들을 함께 요약한다.약한다.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.108-108
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• 2002

• The Journal of Korean Institute of Communications and Information Sciences
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• v.36 no.5A
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• pp.543-550
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• 2011

In order to improve readability, a leakage-signal canceller has been proposed in UHF-band RFID readers. The proposed canceller is composed of two blocks: an environment monitoring block for detection of Tx and RX power levels and a leakage cancellation block. The leakage canceller consists of directional-couplers, digital attenuators and phase shifters based on the feed-forward scheme. The leakage canceller is located between antenna and reader. An improved experimental characterization of the scheme is presented with results from the reader operating in the Korean RFID frequency band, conforming the validity of the approach with more than 29 dB of leakage cancellation.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2018.11a
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• pp.61-62
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• 2018

On this Study, it conducts the Standards investigation about air leakage of Buildings for performance improvement of Smoke control in Fire. In domestic standards, it is incomplete the Criteria standards of air leakage. Therefore it is considered that won't limit the smoke spread due to stack effect. But in the case of the United States, it conducts the Hot smoke air-Leakage test and it consider that can prevent the Smoke Spread in Compartment.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.9-13
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• 2009

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS-56.8(1994) in Korea. Two methods, the make-up flow rate and the pressure decay, are used for LLRT. Though ANSI/ANS-56.8 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the test period for type "C" LLRT is differently applied to each NPPs. Therefore, this study presents a unified test criteria for data stabilization and test duration through experiments to improve the test reliability for type "C".

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.103-111
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• 2020

Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g^-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.175-175
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• 1999