• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.31-32
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.237.2-237
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• 1999

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.183-184
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• 2018

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1115-1126
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• 2009

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.283-288
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• 1996

Direct safety injection into the reactor vessel downcomer annulus(DVI) is a fundamental feature of the KNGR(Korean Next Generation Reactor) four-train safety injection system. The numerical analysis of thermal mixing of ECC(Emergency Core Cooling) water through DVI with the water in the RVDC(Reactor Vessel Downcomer) annulus has been performed, in order to study the impact of nozzle location on the pressurized thermal shock and safety analysis. The results of this study show that the thermal mixing due to the natural circulation induced by the limiting accident conditions is sufficient to prevent temperature in the RVDC from dropping to the level of concern for PTS. When the DVI nozzle is located right above the cold leg, the temperature distribution at the outlet of flow field is most uniform. The tool used for numerical analysis is CFDS-FLOW3D.

• Journal of Welding and Joining
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• v.31 no.5
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• pp.41-46
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• 2013

Different microstructures in the weld zone of a metal structure such as a fusion zone or heat affected zone are formed as compared to the parent material. Thus, the mechanical properties in the weld zone are different from those in the parent material. As the basic data for reliably understanding the structural characteristics of a welded PCHE specimen to be made of Alloy 617, the mechanical properties in the weld zone and parent material for a Alloy 617 plate are measured using an instrumented indentation technique in this study.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.169-174
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• 1998

An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculation were carried out with the RFSP code, provided by cell averaged hel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift art a time. The simulation results show that the maximum channel and bundle powers were maintained below the licence limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period.

• Proceedings of the Korean Institute of Navigation and Port Research Conference
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• v.1
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• pp.315-320
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• 2006

There have been several problems in treating shipboard sewage due to special environmental conditions of ship, such as limited space, rolling and pitching, change of temperature and so on. It was suggested that Sequence Batch Reactor (SBR) might be suitable process for overcome these problems in terms of small size, high capacity of treating wastewater and full automation. In this study a SBR process was employed for biological treatment of organic wastes in the shipboard sewage. This process was able to remove nitrogen and phosphorus as well as organic matter efficiently. More than 95% of chemical oxygen demand(COD) were removed. In addition, about 97% of total nitrogen (T-N) was reduced. The total phosphorus(T-P) reduction averaged 93%. A disturbance operation caused by the treatment of Methylene Blue Active Substances(MBAS) was not observed.

• Nuclear Engineering and Technology
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• v.11 no.4
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• pp.287-293
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• 1979

Mixed standard-FLIP fuel loading patterns in the TRIGA Mark-III reactor were analyzed. It was judged that the mixed loading pattern with the standard fuel in the B-ring and the FLIP fuel in other rings was mostly desirable in view of fuel temperature, cooling condition with the natural convection, or effective thermal flux utilization in the central thimble. In addition, tile maximum useful flux in tile reactor beamports versus the loading patterns was evaluated.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.167-174
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• 2011

A pressurizer in a small integral type pressurized water reactor is located inside the upper region of the reactor vessel, and uses a space between the upper head of the reactor vessel and the upper region of the upper guide structure which is partially filled with fluid depending on the operating power. This new design requires a comprehensive investigation of vibration characteristics. This study investigates the modal characteristics of a pressurizer which uses a simplified cylindrical shell model, focusing on how having fluid in the shell affects vibration and response characteristics. In addition, an analysis of sloshing is performed and the response characteristics are addressed.