• Nuclear Engineering and Technology
• /
• v.34 no.3
• /
• pp.259-267
• /
• 2002

NFPA-803 has been referred as the Fire Protection Standard at the Nuclear Power Plants of Pressurized Water Reactor. This Standard has been used as the fire protection regulation, containing prescriptive requirements with deterministic methodology. Recently, with cumulative efforts by the U.S. Nuclear Regulatory Commission and Utilities in America to establish a new Standard, including a quantitative evaluation methodology, NFPA-805, the Performance-Based Standard for FIRE Protection for Light Water Reactor Electric Generating Plants was issued and approved by the American National Standards Institute as an American National Standard with an effective date of February 9, 2001. This paper presents an analysis result from the computer modeling for the fire simulation In addition, it proposes the idea that this kind of analytic method can be available for the facilities design of fire prevention and protection fields, as well as an evaluation for the fire suppression system with a quantitative analysis for the thermal phenomena in fire compartments in Nuclear Power Plants.

• Nuclear Engineering and Technology
• /
• v.37 no.6
• /
• pp.605-610
• /
• 2005

A method to estimate the relative sensitivity of a self-powered rhodium detector for an upcoming cycle is developed by combining the rhodium depletion data from a nuclear design with the site measurement data. This method can be used both by nuclear power plant designers and by site staffs of Korean standard nuclear power plants for determining which rhodium detectors should be replaced during overhauls.

• Journal of Electrical Engineering and Technology
• /
• v.13 no.6
• /
• pp.2561-2567
• /
• 2018

The extension of DC battery backup time in the DC power supply system of nuclear power plants (NPPs) remains a challenge. The lead-acid battery is the most popular at present. And it is generally the most popular energy storage device. However, extension of backup time requires too much space. The lithium-ion battery has high energy density and advanced gravimetric and volumetric properties. The aim of this paper is development of the sizing formula of stationary lithium-ion batteries. The ongoing research activities and related industrial standards for stationary lithium-ion batteries are reviewed. Then, the lithium-ion battery sizing calculation formular is proposed for the establishment of industrial design standard which is essential for the design of stationary batteries of nuclear power plants. An example of calculating the lithium-ion battery capacity for a medium voltage UPS is presented.

• Nuclear Engineering and Technology
• /
• v.33 no.1
• /
• pp.12-24
• /
• 2001

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.8 no.1
• /
• pp.25-32
• /
• 2012

Korea Hydro & Nuclear Power Co. (KHNP) has developed and implemented the maintenance effectiveness monitoring (MR) programs for the operating nuclear power plants. The MR program is developed by reflecting design characteristics of the operating nuclear power plants to monitor the plant performance for improving the safety and reliability. Recently, KHNP has built a new nuclear power plant, and developed the MR program to establish the advanced maintenance system by reflecting unique design characteristics based on the OPR1000 standard model. So, the MR program developed in this study has another characteristics in comparison with the OPR1000 standard model, and we will verify the suitability of the MR program through evaluating initial performance of the plant. The safety and reliability of the new plant will be improved by developing and implementing the MR program.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
• /
• v.25 no.4
• /
• pp.71-78
• /
• 2011

This paper contains the selection of operation parameters for protection and control of steam turbine in a Korea Standard Type Nuclear Power Plant. The safety of nuclear reactor must be ensured which generates nuclear energy and produces steam. Also, the safety of turbine, which consume the nuclear energy as a core machine, must be ensured. For the purpose of this, we describe how the operating parameters were selected, reviewed, implemented into the operator console and finally put into actual operation of the system.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.05a
• /
• pp.60.2-60.2
• /
• 2004

• Nuclear Engineering and Technology
• /
• v.47 no.1
• /
• pp.1-10
• /
• 2015

After the Fukushima Daiichi nuclear power plant (NPP) accident, new regulatory requirements were enforced in July 2013 and a backfit was required for all existing nuclear power plants. It is required to take measures to prevent severe accidents and mitigate their radiological consequences. The Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) has been conducting numerical studies and experimental studies on relevant severe accident phenomena and countermeasures. This article highlights fission product (FP) release and hydrogen risk as two major areas. Relevant activities in the S/NRA/R are briefly introduced, as follows: 1. For FP release: Identifying the source terms and leak mechanisms is a key issue from the viewpoint of understanding the progression of accident phenomena and planning effective countermeasures that take into account vulnerabilities of containment under severe accident conditions. To resolve these issues, the activities focus on wet well venting, pool scrubbing, iodine chemistry (in-vessel and ex-vessel), containment failure mode, and treatment of radioactive liquid effluent. 2. For hydrogen risk: because of three incidents of hydrogen explosion in reactor buildings, a comprehensive reinforcement of the hydrogen risk management has been a high priority topic. Therefore, the activities in evaluation methods focus on hydrogen generation, hydrogen distribution, and hydrogen combustion.

• Proceedings of the KIEE Conference
• /
• 2008.07a
• /
• pp.685-686
• /
• 2008

For the purpose of reducing the plant trip/transient by the failure of a single component during plant operation or maintenance, the list of critical components with Single Point Vulnerability (SPV) on KSNP (Korean Standard Nuclear Power Plant), the standardized methodology of SPV evaluation and the plan to improve reliability of the equipment have been established. In addition, SPV component lists for the other domestic operating Nuclear Power Plants have been made, and the proper procedure for SPV management will be developed.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.266-266
• /
• 2001