• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.321-326
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• 1996

During the plant operation, steam generator U-tube assembly will potentially be subject to adverse environmental conditions which can cause damages to them. This report addresses the major design factors of CANDU type steam generator which are intended to minimize the potential tube damages. Such factors include U-tube material, high circulation ratio, tube-to-tubesheet joint, tube support design. Also a few suggestions are presented for the design and performance improvement of CANDU type steam generators.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.24 no.3
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• pp.343-351
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• 2011

Nuclear containment building is the last barrier for being secure from any nuclear power plant accident. Therefore, it is very important to understand the ultimate capacity of nuclear containment building to loads associated with severe accidents. LOCA (loss of coolant accident) is considered as the basic accidental load and CANDU-type containment building is considered as a target structure in order to conduct the numerical analysis for the structural safety of a containment building. The CANDU-type containment building is a prestressed concrete shell structure which has the dome and the cylindrical wall and is reinforced with bonded tendons. In this paper, the evaluation of ultimate internal pressure capacity was carried out by nonlinear analysis of a prestressed concrete containment building using 3-dimensional structural analysis system.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.46-50
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• 1996

Tritium dose management is an important aspect of the radiation protection program at CANDU type reactor sites. This paper describes the bioassay and dosimetry of tritium at CANDU reactor sites, especially for Wolsung Nuclear Power Plant. It presents a compilation of information drawn from published papers, technical reports, international and national guidelines as well as practical experience both in Korean and Canadian CANDU Nuclear Power Plants. The implementation of this program would provide a technical basis for demonstrating to workers, managers and regulators that tritium bioassay measurements, dose calculations and records should be of acceptable quality and should meet overall radiation protection program objectives.

• Journal of Radiation Protection and Research
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• v.19 no.1
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• pp.23-35
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• 1994

A computer code DRL was developed to calculate the derived release limits at CANDU type nuclear power plants. The derived release limits resulting from DRL code is to set guidelines for the release of radionuclides in airborne and water-borne effuents during normal operations of a CANDU type nuclear power plant. The DRL code generally follows the methodology Prescribed in the CSA standard N288.1-M87 and uses the Parameter values recommended in the same standards. The DRL code was used to calculate a set of preliminary derived release limits for the Wolsong NPP.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.638-644
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• 2017

The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

• Proceedings of the KIEE Conference
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• 2009.07a
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• pp.666_667
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• 2009

원자력발전소의 비상전력계통은 소외정전 시 발전소 안전정지에 필요한 안전관리 설비에 비상전력을 공급하는 계통으로 비상전력계통 모선에 비상디젤발전기(Emergency Diesel Generator : EDG)가 설치되어 있다. 본 논문에서는 국내 중수로형 원자력 발전소에 대해서 최초로 주기적 안전성 평가(Periodic Safety Review : PSR)을 수행한 결과를 기술하였다.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.257-264
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• 2013

In Canada Deuterium Uranium (CANDU)-type nuclear power plants, the reactor is composed of 380 fuel channels and refueling is performed on one or two channels per day. At the time of refueling, the fluid force of the cooling water inside the channel is exploited. New fuel added upstream of the fuel channel is moved downstream by the fluid force of the cooling water, and the used fuel is pushed out. Through this process, refueling is completed. Among the 380 fuel channels, outer rows 1 and 2 (called the FARE channel) make the process of using only the internal fluid force impossible because of the low flow rate of the channel cooling water. Therefore, a Flow Assist Ram Extension (FARE) tool, a refueling aid, is used to refuel these channels in order to compensate for the insufficient fluid force. The FARE tool causes flow resistance, thus allowing the fuel to be moved down with the flow of cooling water. Although the existing FARE tool can perform refueling in Korean plants, the coolant flow rate is reduced to below 80% of the normal flow for some time during refueling. A Flow rate below 80% of the normal flow cause low flow rate alarm signal in the plant operation. A flow rate below 80% of the normal flow may cause difficulties in the plant operation because of the increase in the coolant temperature of the channel. A new and improved FARE tool is needed to address the limitations of the existing FARE tool. In this study, we identified the cause of the low flow phenomena of the existing FARE tool. A new and improved FARE tool has been designed and manufactured. The improved FARE tool has been tested many times using laboratory test apparatus and was redesigned until satisfactory results were obtained. In order to confirm the performance of the improved FARE tool in a real plant, the final design FARE tool was tested at Wolsong Nuclear Power Plant Unit 2. The test was carried out successfully and the low flow rate alarm signal was eliminated during refueling. Several additional improved FARE tools have been manufactured. These improved FARE tools are currently being used for Korean CANDU plant refueling.

• Nuclear Engineering and Technology
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• v.27 no.2
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• pp.155-164
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• 1995

The steam generator water level control system is studied as a class of randomly sampled nonlinear control systems. The sampling interval and the loop amplification factor are considered as random variables in order to take the operator behavior in account. Stochastic stability using Lyapunov method is used without determining such Lyapunov function. The derived stability criterion is verified with time-domain simulation using the data of CANDU type nuclear power plant, Wolsung 1.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.70-74
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• 2014

An Operating Experience Report(OER) has written about events and accidents happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. In this paper, it is analyses that the foreign NPPs' OERs on JIT published by the International Nuclear Agency(WANO, INPO, COG, BE). The analysis introduced in this paper is performed along with the various factors such as type of work, root-cause, and equipment. The root-cause analysis about the OERs shows that the Human-error is the major factor in foreign NPPs, but on the other hand equipment problem is the main part of the Domestic NPPs. The ratio of the foreign NPP's OERs on JIT according to the type of work was applied to KHNP-JIT developed nowadays for the first time in KOREA.

• Proceedings of the KIEE Conference
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• 1997.11a
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• pp.567-569
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• 1997

A control system using a PLC has been developed for a single tower drier(STD) in a CANDU type nuclear power plant. This system will replace the existing STD control system which was implemented with mechanical timers and relays. The new control system makes it possible for an operator to perform more precise time and dew point control for the STD, thanks to the high efficiency and flexibility of the PLC. The operational cost for the control system is much reduced compared to the existing system.