• Journal of Energy Engineering
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• v.12 no.4
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• pp.281-288
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• 2003

The MATRA-LMR code has been developed based on a subchannel analysis method for LMR (Liquid Metal Reactor) core subassembly thermal hydraulic design and analysis. The code was improved to allow a seven assembly calculation and can account for inter-assembly heat transfer based on a lumped parameter model. This paper describes the main modifications and improvements of the code and shows reference calculation results which compared single assembly calculation with seven assembly calculation cased for driver and blanket subassemblies of the KALIMER 150 MWe breakeven conceptual design core. KAL- IMER is a pool-type sodium cooled reactor with a thermal output of 392.0 MWth, which have inherently safe, environmentally friendly, proliferation-resistant and economically viable reactor concepts.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.199-200
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.527-528
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• 2004

• CDE review
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• v.3 no.3
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• pp.19-24
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• 1997

이 글에서는 다음의 내용에 대하여 다루었다. 회의적 반응을 뒷받침하는 이유들, 역행할 수 없는 추세, 높은 가격의 정당화, Collaborative Engineering, 지식 기반 시스템, 불평의 원인, 누가 생존할 것인가?, Parametric Technology and SDRC, Dassault Systems, EDS/Unigraphics, Hewlett-Packard사의 CoCreate, Intergraph, Matra Datavision, Autodesk, 무엇을 할 것인가?

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.566-567
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.590-591
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.529-530
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.67-68
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• 2005

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2303-2308
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• 2007

Experiments were performed to determine the thermal (or turbulent) diffusion coefficient (TDC) and to investigate the critical heat flux (CHF) performance in the 5${\times}$5 rod bundle with 5 unheated rods which are supported by Hybrid Mixing Vane. In this study, HFC-134a fluid was used as working fluid and the fluid temperature were measured in the important subchannels. To determine the TDC value, the measured fluid temperatures were compared with the predicted values obtained from the MATRA code. The best optimized value of ${\beta}$ was found to be 0.02 by considering prediction statistics, i.e., average and standard deviations of the differences between the experimental results and code calculations. Using the best optimized value of ${\beta}$ as 0.02, the MATRA code predicts the test results of the fluid temperature within ${\pm}$1.0 % of error. According to the experimental results on CHF of 5 non-heating guide tubes, the case with non-heating guide tube showed a little good performance in terms of CHF.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1071-1082
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• 2016

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.