• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.313-313
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.248.2-248
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.266.2-266
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.344.1-344
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.372.1-372
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• 1999

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.502-516
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• 2002

Wolsong NPP are CANDU6 type reactors and there are 4 CANDU6 type reactors in commercial operation. CANDU type reactors require on-power refuelling by two remote controlled F/Ms (Fuelling Machines). Most of channels, fuel bundles are float by channel coolant flow and move toward downstream, however in about 30% of channels the coolant flow are not sufficient enough to carry fuel bundles to downstream. For those channels a special device, FARE (Flow Assist Ram Extension) device, is used to create additional force to push fuel bundles. It has been showing that during fuelling operation of some channels the channel coolant flow rate is reduced below specified limit (80% of normal), and consequently trip alarm signal turns on. This phenomenon occurs on selected channels that are instrumented for the channel flow and required to use the FARE device for refuelling. Hence it is believed that the FARE device causes the problem. It is also suspected that other channels that do not use the FARE device for refuelling might also go into channel flow low state. The analysis revealed that the channel How low occurs as the FARE device is introduced into the core and disappears as the FARE device is removed from the core. This paper presented the FARE device behavior, detailed fuelling operation sequence with the FARE device and effect on channel flow low phenomena. The FARE device components design changes are also suggested, such as increasing the number or now holes in the tube and flow slots in the ring orifice.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.192-192
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• 2008

As the lifetime of nuclear power plants (NPPs) reaches design life, the probability for fatal accidents increases. Most of accidents are known to be caused by degradation of mechanical components. Pressure tubes are the most important components in CANDU reactor. They are subjected to various aging mechanisms such as delayed hydride cracking (DHC), irradiation and corrosion, etc. Therefore, the integrity of pressure tube is key concern in CANDU reactor. Up to recently, conventional deterministic approaches have been utilized to evaluate the integrity of components. However, there are many uncertainties to prevent a rational evaluation. The objective of this paper is to assess the failure probability of pressure tube in CANDU. To do this, probability fracture mechanics (PFM) analysis based on the Genetic Algorithm (GA) is performed. For the verification of the analysis, a comparison of the PFM analysis using a commercial code and mathematical method is carried out.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.267-276
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• 2017

In any reactor physics analysis, the instantaneous power distribution in the core can be calculated when the actual bundle-wise burnup distribution is known. Considering the fact that CANDU (Canada Deuterium Uranium) utilizes on-power refueling to compensate for the reduction of reactivity due to fuel burnup, in the CANDU fuel management analysis, snapshots of power and burnup distributions can be obtained by simulating and tracking the reactor operation over an extended period using various tools such as the $^*SIMULATE$ module of the Reactor Fueling Simulation Program (RFSP) code. However, for some studies, such as an evaluation of a conceptual design of a next-generation CANDU reactor, the preferred approach to obtain a snapshot of the power distribution in the core is based on the patterned-channel-age model implemented in the $^*INSTANTAN$ module of the RFSP code. The objective of this approach is to obtain a representative snapshot of core conditions quickly. At present, such patterns could be generated by using a program called RANDIS, which is implemented within the $^*INSTANTAN$ module. In this work, we present an alternative approach to derive the patterned-channel-age model where a simulated-annealing-based algorithm is used to find such patterns, which produce reasonable power distributions.

• Journal of Energy Engineering
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• v.4 no.1
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• pp.103-114
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• 1995

본 연구는 원자력발전소용 시뮬레이션 언어인 DSNP(Dynamic Simulator for Nuclear Power-plants)언어를 이용하여 CANDU-6 발전소 운전 모사 프로그램을 구성함으로써 핵심계통인 1차 냉각재 계통(PHTS)과 2차 계통 일부가 정상 및 과도조건에서 보일 수 있는 운전 상태를 연구하였다. DSNP 프로그램은 원자로심과 증기발생기에서의 열전달 모델, 열수송계통 펌프 모델 및 가압기 열수력 모델을 포함하고 있으며, 파이프(pipe)라는 단위 구성체를 이용하여 1차 냉각재계통을 노드화하여 계통 모사가 실현된다. 정상상태 100% 전출력 운전시 대표적인 운전변수를 기준으로 DSNP 결과와 CANDU-6 발전소 설계치를 비교해 본 결과 서로 매우 근사한 값을 나타내었으며, 이는 과도상태 모사의 초기조건으로 합당한 것으로 판단된다. 본 연구에서 선택된 과도상태 모사시 DSNP 프로그램은 매우 안정된 최종정상상태를 얻음에 따라 원자로의 기계 물리학적 변화를 합리적으로 모사하고 있음을 알 수 있었다. 최종 정상상태 회귀 이전의 동적 거동을 원자로 설계자료인 예비 안전성 평가 보고서(PSAR)와 비교한 결과 단기적 거동은 PSAR 결과와 다소 다른 점이 있었으나 전체적으로 합리적인 운전변수 값을 얻을 수 있었다. 단기적 거동에 대한 입증은 원자로 운전 자료를 통하여 가능할 것으로 사료된다. 이상과 같이 본 연구를 통해 구성한 DSNP 프로그램은 보완 및 개선의 여지가 있으나 현재의 수준으로도 CANDU-6 발전소의 일부 과도상태 모사가 가능한 것으로 판단된다.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4042-4051
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• 2021

In this work, we applied a multilevel modeling technique to estimate the diametral creep in the pressure tubes of Korean Canada Deuterium Uranium (CANDU) units. Data accumulated from in-service inspections were used to develop the model. To confirm the strength of the multilevel models, a 2-level multilevel model considering the relationship between channels for a CANDU unit was compared with existing linear models. The multilevel model exhibited a very robust prediction accuracy compared to the linear models with different data pooling methods. A 3-level multilevel model, which considered individual bundles, channels, and units, was also implemented. The influence of the channel installation direction was incorporated into the three-stage multilevel model. For channels that were previously measured, the developed 3-level multilevel model exhibited a very good predictive power, and the prediction interval was very narrow. However, for channels that had never been measured before, the prediction interval widened considerably. This model can be sufficiently improved by the accumulation of more data and can be applied to other CANDU units.