• Geomechanics and Engineering
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• v.10 no.1
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• pp.109-124
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• 2016

In this paper, process of dynamic powder compaction is investigated experimentally using impact of drop hammer and die tube. A series of test is performed using aluminum powder with different grain size. The energy of compaction of powder is determined by measuring height of hammer and the results presented in term of compact density and rupture stress. This paper also presents a mathematical modeling using experimental data and neural network. The purpose of this modeling is to display how the variations of the significant parameters changes with the compact density and rupture stress. The closed-form obtained model shows very good agreement with experimental results and it provides a way of studying and understanding the mechanics of dynamic powder compaction process. In the considered energy level (from 733 to 3580 J), the relative density is varied from 63.89% to 87.41%, 63.93% to 91.52%, 64.15% to 95.11% for powder A, B and C respectively. Also, the maximum rupture stress are obtained for different types of powder and the results shown that the rupture stress increases with increasing energy level and grain size.

• Proceedings of the KSME Conference
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• 2003.11a
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• pp.373-378
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• 2003

It is important that overheat protection of super heated tube in boiler operation and maintenance. The overheat of super heat tube can make damage and rupture of tube material, which causes accidental shutdown of boiler. The super heated tube overheat is almost due to the lack of uniformity of gas temperature distribution. There are two ways to protect overheat of super heated tube. The one is to control hot gas operation pattern which is temperature or flow distribution. the other is to control super heated steam flow distribution. The former is difficult than the later, because of control device design. In this paper steam flow control method which uses orifices is proposed to protect overheat of super heat tube.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.39-46
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• 2014

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.519-520
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.523-524
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.183-183
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.150.2-150.2
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.77-78
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.167.1-167
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• 1999

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.870-880
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• 2016

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.