• 원자력산업
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• 제16권3호통권157호
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• pp.4-13
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• 1996

한국중공업(주)는 최근 캐나다에 이어 세계에서 두 번째로 700MW급 중수로형 원자로(CALANDRIA)를 국산화 개발하는데 성공, 월성 4호기 건설현장으로 출하하였다. 지난 94년 5월 제작에 착수하여 19개월만에 완료된 이 국산 원자로는 스테인리스와 튜브 소재인 지르코늄 등 초합금강으로 제작된 계약금액이 120억원에 이르는 고부가가치 제품으로, 이번 국산화에 따라 수입대체효과는 물론, 앞으로 중국을 비롯한 동남아시아 지역의 원자력발전소 수출에 새로운 전기가 될 것으로 기대된다. 그 간의 개발경위와 국산화의 의의 등을 살펴본다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.499-504
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• 2001

A validation of a 3D CFD model for predicting local subcooling of moderator in the vicinity of calandria tubes in a CANDU reactor is performed. The small scale moderator experiments performed at Sheridan Park Experimental Laboratory(SPEL) in Ontario, Canada[1] is used for the validation. Also a comparison is made between previous CFD analyses based on 2DMOTH and PHOENICS, and the current model analysis for the same SPEL experiment. For the current model, a set of grid structures for the same geometry as the experimental test section is generated and the momentum, heat and continuity equations are solved by CFX-4.3, a CFD code developed by AEA technology. The matrix of calandria tubes is simplified by the porous media approach. The standard $k-\varepsilon$ turbulence model associated with logarithmic wall treatment and SIMPLEC algorithm on the body fitted grid are used and buoyancy effects are accounted for by the Boussinesq approximation. For the test conditions simulated in this study, the flow pattern identified is a buoyancy-dominated flow, which is generated by the interaction between the dominant buoyancy force by heating and inertial momentum forces by the inlet jets. As a result, the current CFD moderator analysis model predicts the moderator temperature reasonably, and the maximum error against the experimental data is kept at less than $2.0^{\circ}C$ over the whole domain. The simulated velocity field matches with the visualization of SPEL experiments quite well.

• Nuclear Engineering and Technology
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• 제30권5호
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• pp.403-415
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• 1998

The CANDU 6 reactor assembly consists of an array of 380 pressure tubes, which are installed horizontally in a large cylindrical vessel, the Calandria, containing the low pressure heavy water moderator. The pressure tube is located inside the calandria tube and the annulus between these tubes, which forms a closed loop with $CO_2$ gas recirculating, is called the Annulus Gas System(AGS). It is designed to give an alarm to the operator even for a small pressure tube leak by a very sensitive dew point meter so that he can take a preventive action for the pressure tube rupture incident. To judge whether the operator action time is enough or not in the design of Wolsong 2,3 & 4, the Leak Before Break(LBB) assessment is required for the analysis of the pressure tube failure accident. In order to provide the required data for the LBB assessment of Wolsong Units 2, 3, 4, a series of leak detection capability tests was performed by injecting controlled rates of heavy water vapour. The data of increased dew point and rates of rise were measured to determine the alarm set point for the dew point rate of rise of Wolsong Unit 2. It was found that the response of the dew point depends on the moisture injection rate, $CO_2$ gas flow rate and the leak location. The test showed that CANDU 6 AGS can detect the very small leaks less than few g/hr and dew point rate of rise alarm can be the most reliable alarm signal to warn the operator. Considering the present results, the first response time of dew point to the AGS $CO_2$ flow rate is approximated.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.464-470
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• 1997

Severe accident analysis has been performed for the Wolsung nuclear power plants in Korea to investigate severe accident phenomena of CANDU-600 reactors as a part of Level II PSA study. The accident sequence analyzed in this paper is loss of active heat sinks(LOAH) which is caused by loss of off-site power, diesel generators, and DC power. ISAAC (Integrated Severe Accident Analysis Code)computer code developed by KAERI (Korea Atomic Energy Research Institute) was used in this analysis. This paper describes the important thermal-hydraulics and source term behaviors in the primary system and inside containment, and the failure mechanism of calandria vessel and containment. In addition, some insights for accident management program(AMP) are also given.

• Nuclear Engineering and Technology
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• 제28권4호
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• pp.423-430
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• 1996

The improved and advanced Annulus Gas System(AGS) has been developed for Wolsong 2 to satisfy the requirements of the regulatory body. The Atomic Energy Control Board(AECB) required a shorter detection time following a small leak from a pressure tube and/or calandria tube. This paper describes licensing requirements, functional requirements and detail design description for the AGS. The Wolsong unit No. 1 AGS was designed to operate as a stagnant system normally requiring only pressure regulation and having provisions for purging. no improved AGS involves the adoption of gas recirculation in AGS, duplication of dew point indicators with additional instrumentation and sampling provisions to prompt operator action. The improved system operates in the recirculation mode with continuous dew point measurement for leak detection. An AGS with improved detection instrumentation is provided.

• 대한전자공학회:학술대회논문집
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• 대한전자공학회 1999년도 하계종합학술대회 논문집
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• pp.1083-1086
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• 1999

중수로(CANDU) 형 월성 원자력발전소의 칼란드리아 압력관 전면부를 감시점검하기 위한 열영상 관측시스템을 설계/제작하였다. 중수로는 가동중에 핵연료를 교체한다. 칼란드리아 전면부에는 380 개의 압력관 채널이 위치하고 있다. 핵연료를 교체할 시에 핵연료 교체장비가 칼란드리아 압력관 채널의 ENDCAP을 열고 핵연료를 장전하는 과정에서 발생할 지도 모르는 중수누출, 핵연료교체장비의 이상상태를 점검하는데 목적이 있다. 열영상카메라는 상용 CCD 카메라에 비해 영상의 해상도가 떨어진다. CCD 카메라는 수증기 누출과 같은 육안검사에 활용하고, 열영상카메라는 압력관 채널의 온도변화 등을 점검하기 위해 CCD/열영상카메라의 융합구조로 설계/제작하였다.

• 대한전자공학회:학술대회논문집
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• 대한전자공학회 1999년도 하계종합학술대회 논문집
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• pp.1087-1090
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• 1999

중수로(CANDU) 형 월성 원자력발전소의 칼란드리아 압력관 전면부를 감시점검하기 위한 열영상 관측프로그램을 개발하였다. 국내의 사용자들의 요구에 부응할 수 있는 다양한 기능을 부가 하였다. 부가된 기능에는 임의의 포인트, 영역, 라인, 경계선을 마우스로 지정하여, 선택된 지점, 영역, 라인 및 경계선의 특징을 추출할 수 있는 기능을 갖추고 있다. 또한 KAERI Thermo Inspector 의 기능을 살린 일반영상/열영상의 매핑기능을 부가하였다. 일반영상에 비해 상대적으로 해상도가 떨어지는 열영상대신에 CCD 영상의 관측포인트를 지정하면 열영상카메라좌표계의 매핑된 지점의 이상상태를 판정할 수 있는 특성을 갖는다.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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• pp.66-71
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• 2004

Estimation of the CANDU pressure tube deflection is important since the deflection may cause significant structural failure due to hydrogen diffusion and blister. However, there is no appropriate engineering model to estimate it exactly. The purpose of this paper is to propose a new analysis method and program to resolve this issue. For development of proper analysis method, a series of finite element analyses has been carried under elastic-creep condition. In addition, for effective estimation of the creep deflection, an analysis program named PC-DAS was developed based on the proposed method. Comparison of simple case study results with corresponding reference ones showed good agreement. Therefore, the proposed method and program can be utilized as one of valuable toolkit for integrity assessment of CANDU pressure tube.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.