• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.1291-1292
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.05a
• /
• pp.434-434
• /
• 1999

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.741-742
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.368-368
• /
• 2002

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.10a
• /
• pp.101-102
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 1999.10a
• /
• pp.365.2-365
• /
• 1999

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1414-1422
• /
• 2017

Remote response technology has advanced to the extent that a robot system, if properly designed and deployed, may greatly help respond to beyond-design-basis accidents at nuclear power plants. Particularly in the aftermath of the Fukushima accident, there is increasing interest in developing disaster robots that can be deployed in lieu of a human operator to the field to perform mitigating actions in the harsh environment caused by extreme natural hazards. The nuclear robotics team of the Korea Atomic Energy Research Institute (KAERI) is also endeavoring to construct disaster robots and, first of all, is interested in finding out to what extent safety benefits can be achieved by such a disaster robotic system. This paper discusses a new approach based on the probabilistic risk assessment (PRA) technique, which can be used to quantify safety benefits associated with disaster robots, along with a case study for seismic-induced station blackout condition. The results indicate that to avoid core damage in this special case a robot system with reliability > 0.65 is needed because otherwise core damage is inevitable. Therefore, considerable efforts are needed to improve the reliability of disaster robots, because without assurance of high reliability, remote response techniques will not be practically used.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.11 no.2
• /
• pp.61-67
• /
• 2015

3D FEM modeling based on 3D CAD data has been performed to evaluate the efficiency of CP system in a real operating nuclear power plant. The results of it successfully produced sophisticated profiles of electrolytic potential and current distributions in the soil of an interested area. This technology is expected to be a breakthrough for detection technology of damages on buried pipes when it comes into combining with a brand of area potential earth current (APEC) and ground penetrated radar (GPR) technologies. 2D current distribution and 2D current vectors on the earth surface from the APEC survey will be used as boundary conditions with exact 3D geometry data resulting in visualization of locations and extents of corrosion damages on the buried pipes in nuclear power plants.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.772-777
• /
• 1997

Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and in the hot leg and SG during the TMLB' scenrio. The objective of this study is to develop a natural circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models.

• Nuclear Engineering and Technology
• /
• v.50 no.6
• /
• pp.907-914
• /
• 2018

In this article, we conducted molecular dynamics simulations to investigate the effect of applied strain and temperature on irradiation-induced damage in alpha-zirconium. Cascade simulations were performed with primary knock-on atom energies ranging between 1 and 20 KeV, hydrostatic and uniaxial strain values ranging from -2% (compression) to 2% (tensile), and temperatures ranging from 100 to 1000 K. Results demonstrated that the number of defects increased when the displacement cascade proceeded under tensile uniaxial hydrostatic strain. In contrast, compressive strain states tended to decrease the defect production rate as compared with the reference no-strain condition. The proportions of vacancy and interstitial clustering increased by approximately 45% and 55% and 25% and 32% for 2% hydrostatic and uniaxial strain systems, respectively, as compared with the unstrained system, whereas both strain fields resulted in a 15-30% decrease in vacancy and interstitial clustering under compressive conditions. Tensile strains, specifically hydrostatic strain, tended to produce larger sized vacancy and interstitial clusters, whereas compressive strain systems did not significantly affect the size of defect clusters as compared with the reference no-strain condition. The influence of the strain system on radiation damage became more significant at lower temperatures because of less annealing than in higher temperature systems.