• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.

• Journal of Radiation Protection and Research
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• v.35 no.2
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• pp.57-62
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• 2010

Maintenance on the water chamber of steam generator, the change of pressurizer heater, the removal of pressure tube feeder, and so on during outage in nuclear power plants (NPPs) has a likelihood of high radiation exposure to whole body of workers even short time period due to the high radiation exposure rates. In particular, it is expected that hands would receive the highest radiation exposure because of its contact with radiation materials. In this study, field tests on extremity dose assessment of radiation workers for contact works with high radiation exposure were conducted during the maintenance periods in Korean pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs). In this field test, radiation workers were required to wear additional TLDs on the back and wrist, and an extremity dosimeter on fingers including a main TLD on the chest, while performing maintenance. As a result, it was found that the equivalent dose for fingers was distributed in the fixed range of deep dose and the equivalent dose for wrists.

• Proceedings of the KAIS Fall Conference
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• 2008.11a
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• pp.365-367
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• 2008

원자로 냉각계통의 압력경계를 구성하고 있는 재료들의 부식은 재료 표면에 형성되는 산화막, 금속재료의 구성성분이 용해되어 생성된 가용성 화학종 및 산화물 입자 형태의 부식생성물들을 발생시킨다. 금속합금의 부식에 의한 가용성 화학종 및 입자들의 방출은 원자로 냉각계통에서 노심과 증기발생기를 순환하면서 연료피복관 위에 침전되어 여러 가지 문제를 야기한다. 크러드는 구조재료의 부식에 기인하여 발생한 부식생성물들이 냉각수에 부유하여 떠다니거나 피복관 표면에 침적하여 형성되며 주로 니켈과 철 산화물로 구성되어 있다. 원자로 냉각계통에서 크러드를 최소화하기 위하여 수화학 조건들을 제어하지만 장주기 고연소도 노심에서 AOA 현상을 일으키는 주된 원인이 되고 있다. 피복관 위에 침적되는 크러드는 붕소의 잠복위치를 제공할 뿐만 아니라 냉각수의 압력강하를 증가시키고 피복관의 부식 및 파손 원인을 제공하며 방사선 준위가 증가하도록 한다. 따라서 본 연구에서는 반응속도론적 관점에서 원자로 정지시의 용출 크러드 특성에 대한 연구를 수행하였다.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.33 no.9
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• pp.97-103
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• 2005

In this paper, effects on the heat transport limitation of heat pipe by the wick structural factors were theoretically analyzed for the sintered-copper wick heat pipe. Uniformity of particle size and sintering process were acted as dominant factors on the pore distribution and wick porosity, and small deviations of the wick thickness and the pore size greatly affected the heat transport limitations of the heat pipe. Especially, slight variations of the wick thickness, mean particle radius and capillary radius along the vapor temperatures and inclination angles remarkably changed the capillary limitation of the heat pipe.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.113.2-113.2
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• 2005

• 월간 기계설비
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• s.252
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• pp.50-51
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• 2011

서울시는 최근 자원회수시설에서 쓰레기 소각 시 발생하는 질소산화물 저감방식을 개선해 매년 54억원의 예산을 절감하게 됐다고 밝혔다. 이번 아이디어는 독일 쉬텔링거 모어 소각장의 열교환시스템을 벤치마킹한 것으로 마포 자원회수시설 담당 공무원(김창환 주무관)이 지난 해 6월 독일을 방문한 뒤 제안하면서 탄생하게 됐다. 개선 방식은 질소산화물을 제거하기 위해 설치된 SCR촉매탑의 가온시스템을 LNG를 이용한 닥터버너 방식에서 '소각증기 사용 열교환 방식'으로 전환시킨 것이다.

• Journal of the Korean Society of Safety
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• v.15 no.3
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• pp.57-65
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• 2000

The stress corrosion crack growth is simulated assuming a small axial surface crack inside a S/G tube. Internal pressure and residual stresses are considered as applied forces. Stress intensity factors along crack front, variation of crack shape and crack growth rate are obtained and discussed. It is noted that the aspect ratio of the crack is not depend on the initial crack shape but depend on the residual stress distribution.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.05a
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• pp.191-192
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• 2011

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.229-230
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• 2016