• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.1 s.173
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• pp.206-214
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Journal of the KSME
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• v.33 no.5
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• pp.449-455
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• 1993

지금까지의 CANDU 사고이력과 관련된 문제점을 살펴보면 핵연료 채널의 부적절한 설계 및 설치 그리고 부적절한 압력관 가동조건 등에 많은 문제점이 있었다. 이러한 의미에서 CANDU의 안전성은 압력관의 건전성으로부터 확보된다 하여도 과언이 아니다. 그러나 CANDU에서 차지 하는 중요성에 비추어 압력관의 사용환경은 매우 열약하다. 따라서 가동중 압력관 건전성 위협 요인에 대한 정기적인 검사, 시험 및 평가는 CANDU 안전성확보의 첫걸음이 된다. 특히 건전 성평가에 필요한 주요자료가 압력관 인출시험결과로부터 확보됨을 고려할 때 압력관 인출시험을 국내에서 수행할 수 있는 능력을 확보하는 것 또한 우리에게 부과된 과제라 할 것이다.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.342.2-342
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.253-253
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.266.2-266
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.225-225
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.68-68
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• 2004

• Nuclear Engineering and Technology
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• v.30 no.5
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.295-299
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• 1996

This analysis is to investigate the benefits and disadvantages of increasing the pressure tube wall thickness for CANDU reactor. Creep analysis of the pressure tube was performed for slightly enriched uranium (SEU) to establish the reduction in axial elogation and diametral creep provided by a thicker wall pressure tube.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.3
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• pp.455-463
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• 2003

The aim of this study is to investigate the hydrogen embrittlement of Zr-2.5Nb CANDU pressure tube at room temperature. The transverse tensile and fracture toughness tests were performed at various hydrogen concentrations using transverse tensile specimens and CCT (curved compact tension) specimens. These specimens were directly machined from the pressure tube retaining original curvatures. Based on the results of these tests. the hydrogen embrittlement phenomenon was clearly observed and fracture toughness parameters of Zr-2.5Nb pressure tube materials such as, $K_{J(0.2)}$.$J_{ML}$.dJ/da, were dramatically decreased with the increasement of the hydrogen concentration. From microscopic observation by SEM and TEM, it was also revealed that various shapes dimples, fissures and quasi-cleavage were found at the hydrogen-absorbed materials with hydrides while traditional shape dimples were generally located at the as-received materials Through the comparison of the hydride and fissure lengths with the hydrogen concentration the new evaluation method of hydrogen embrittlement was suggested.