• 대한기계학회논문집B
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• 제30권9호
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• pp.825-833
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• 2006

To investigate the moderator coolability for CANDU-6 reactors, a test facility (HU-KINS) has been manufactured as a 1/8 scaled-down of a calandria tank. In the design of the test facility, a scaling law was developed in such a way to consider the thermal-hydraulic characteristics of a CANDU-6 moderator. The proposed scaling law takes into consideration of the energy conservation, the dynamic similitude such as dimensionless numbers, Archimedes number (Ar) and Reynolds number (Re), and thermal-hydraulic properties similitude. Using this proposed scaling law, the thermal-hydraulic scaling analyses of similar test facilities such as the SPEL (1/10 scale) and the STERN (1/4 scale), have been identified. As a result, in the case of the SPEL, while the energy conservation is well defined, the similarities of Ar and the heat density are not well considered. As for the similarity of the STERN, while both the energy conservation and the characteristics of Ar are well defined, the heat density is not. In the meanwhile, the HU-KINS test facility with 1/8 length scaled-down is well similitude in compliance with all similarities of the energy conservation, the fluid dynamics and thermal-hydraulic properties. To verify the adequacy of the similarities in terms of thermal-hydraulics, a computational fluid dynamic (CFD) analysis has been conducted using the CFX-5 code. As the results of the CFD analyses, the predicted flow patterns and variation of axial properties inside the calandria tank are well consistant with those of previous studies performed with FLUENT and this implies that the present scaling method is acceptable.

• 에너지공학
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• 제21권4호
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• pp.419-426
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• 2012

본 연구에서는 기기 스케일 2상 유동(Two-phase flow) 해석 코드 CUPID를 사용하여 CANDU 원자로의 칼란드리아 용기 내부 감속재의 열수력 거동을 분석하기 위한 사전연구를 수행하였다. 먼저, Stern 연구소에서 수행한 단상유동 실험 3종류를 이용하여 CUPID 코드를 검증하였다. 칼란드리아 관다발 영역 격자생성의 복잡성을 피하기 위하여 다공성 매질 모델을 해당 영역에 적용하였고, 다공성 매질 영역의 유동 저항은 실험에서 얻은 관계식을 이용하여 계산하도록 하였다. 계산결과, CUPID 코드는 칼란드리아 용기 내부의 강제 및 자연 대류의 혼합 유동 양식을 성공적으로 예측하였다. 다음으로 2상 유동이 발생하는 경우를 해석하였다. 이들 계산을 통해 CUPID 코드의 CANDU 원자로 감속재 해석 능력을 보였다. 또한, 국부 과냉각 여유도를 예측하는데 사용할 수 있는 유입유량 대비 칼란드리아 용기의 국부 최대 감속재 온도 그래프를 제시하였다.

• 에너지공학
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• 제23권4호
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• pp.95-105
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• 2014

CUPID 코드는 기기 스케일(Component scale)의 2상 유동(Two-phase flow) 해석 코드로서 다양한 2상 유동 조건의 실험 자료를 이용하여 검증되어 왔다. 특히, CUPID 코드의 CANDU형 원자로 감속재 탱크 내부 유동 해석능력을 평가하기 위해 1/4 규모 축소 실험장치의 실험결과를 이용하여 검증한 바가 있다. 본 연구에서는 이전 연구를 바탕으로 CUPID 코드를 사용하여 실제 원자로 감속재탱크 내부의 열수력 거동을 해석하였다. 감속재 탱크의 내부 구조는 아주 복잡하기 때문에 다공질 매질 방법을 적용하였으며 탱크 입구노즐 또한 기기 스케일 코드의 취지에 부합하게 아주 단순화하여 모델하였다. 해석결과의 정확성을 결정하는 가장 중요한 요소는 입구노즐의 모델 방법에 있는 것으로 나타났다. 입구노즐을 단순하게 모델하여 입구유량을 경계조건으로 부여하고 발전소 정상운전조건으로 계산한 결과, 부력에 의한 열성층화 현상이 발생하였다. 이는 전혀 타당하지 않은 것으로 입구 유동의 모멘텀을 정확하게 모의하지 않아 발생한 것이 나타났다. 이를 개선하고자 입구 유량과 운동량을 동시에 보존시킬 수 있도록 입구 노즐 면적을 축소하고 속도는 증가시켜서 계산한 결과, 사실적인 내부 유동장을 얻을 수 있었다. 결론적으로 계산 비용효과가 뛰어난 다공질 매질 방법에 입각하여 CUPID 코드를 실규모 감속재 탱크 열유동 해석에 적용할 수 있음을 보였고, 입구노즐의 적절한 모델이 가장 중요한 요소임을 확인하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.275-276
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.715-716
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• 2005

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.120-120
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• 2002

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.602-603
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• 2005

• Nuclear Engineering and Technology
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• 제30권3호
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• pp.262-272
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• 1998

The experiences and causes of pressure tube cracking accidents in the CANDU reactors and the development of the fuel channel at AECL(Atomic Energy Canada Limited) have been described. Most of the accidents were caused by Delayed Hydride Cracking(DHC). In the cases of the Pickering units 3&4 and the Bruce unit 2, excessive residual stresses induced by an improper rolled joint process played a role in DHC. In the Pickering unit 2, cracks formed by contact between the pressure and calandria tubes due to the movement of the garter spring were the direct cause of the failure. To extend the life of a fuel channel, several R&D programs examining each component of the fuel channel have been carried out in Canada. For a pressure tube, the main concern is focused on changing the fabrication processes, e.g., increasing cold working rate, conducting intermediate annealing and adding a third element like Fe, V, and Cr to the tube material. In addition to them, chromium plating on the end fitting and increasing wall thickness at both ends of the calandria tube are considered. There has also been much interest in the improvement of fuel channel performance in our country and several development programs are currently under way.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.589-596
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• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.