• Nuclear Engineering and Technology
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• 제34권2호
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• pp.176-185
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• 2002

A 4-bundle shift refuelling method of CANDU6 F/H (Fuel Handling) System is analyzed to assess the operational flexibility and capacity of F/H system. The current 8-bundle shift refuelling scheme requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The analysis showed that the 4-bundle shift refuelling method increases F/M (Fuelling Machine) duty cycle and operator load. However, the fuellin’g method change from the 8- to 4-bundle shift refuelling ill not require additional team of operators. A marginal increase in the maintenance cost may be resulted in by the change of fuelling method and the increase of fatigue usage factors requires some components to be replaced during the FM lifetime.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.1008-1013
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• 1995

The CANFLEX fuel bundle has been developed by KAERI/AECL jointly to facilitate the use of various fuel cycles in CANDU-6 reactor. As one of the design evaluations, the structural analysis of the fuel bundles by hydraulic drag force is performed to evaluate the fuel integrity in the period of the refuelling in CANDU-6. The structural integrity is evaluated by FEM modelling for the complicated bundles configuration in channel. It is noted that the present analysis method is newly developed for the structural integrity evaluation. The analysis results show that the fuel bundle is shown to keep its structural integrity during the refuelling.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.643-648
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• 1996

Recent failures of Heat Transport (HT) system Liquid Relief Valves (LRVs) at several CANDU 6 plants resulted in heavy water spills. A multi-discipline KAERI task team along with AECL representatives was set up to assess the specific implications on the Wolsong 2/3/4 station. The study applied the knowledge gained from the generic study by conducting a fundamental review of the specific design of Welsong 2/3/4. The purpose of the study was to demonstrate compliance with the relevant codes and licensing regulations and to identify the improvements to the design and operating procedures applicable to Wolsong 2/3/4. This paper presents the key recommendations of this study.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.719-724
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• 1998

A preliminary fuel channel analysis for 35% reactor inlet header (RIH) break in CANDU reactor loaded with the CANFLEX-RU fuel bundles has been performed. The predicted results are compared with those for the reactor compared with those for the reactor loaded with standard 37-element bundles. The maximum fuel centerline and sheath temperatures for the CANFLEX-RU bundle channel were lower by 338 and 122 $^{\circ}C$, respectively, than those for the standard bundle because of the Bower maximum linear power of the CANFLEX-RU bundle In spite of the 0.4 FPS higher power pulse of the CANFLEX-RU bundle case. Fuel integrity margin to fuel breakup for the CANFLEX-RU bundle is about 50 J/g higher than that for the standard bundle. The PT/CT contact for the CANFLEX-RU bundle occurred 2 s later than that for the standard bundle. The PT/CT contact temperature for the CANFLEX-RU bundle was 2 $^{\circ}C$ lower than that for the standard bundle. These provide the CANFLEX-RU bundle with the negligibly enhanced safety margin for the fuel channel integrity in CANDU 6 reactor, compared with the standard bundle.

• Nuclear Engineering and Technology
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• 제36권1호
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• pp.53-63
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• 2004

The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same($0.008\%$), which showed that the CCMS could monitor core behaviors well.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.209-210
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.545-546
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.142-142
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.400-400
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.160.2-160
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• 2001