• Journal of the Korean Society of Safety
• /
• v.32 no.1
• /
• pp.140-144
• /
• 2017

Fire protection for nuclear power plants should be designed according to the concept of "Defense in Depth" to achieve the reactor safety shutdown. This concept focuses on fire prevention, fire suppression and safe shutdown. Fire prevention is the first line of "Defense in Depth" and the licensee should establish administrative measures to minimize the potential for fire to occur. Administrative measures should include procedures to control handling and use of combustibles. Electrical cables is the major contributor of fire loads in nuclear power plants, therefore electrical cables should be fire retardant. Electrical cables installed in nuclear power plants should pass the flame test in IEEE-383 standard in accordance with NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants". To assure the fire retardant of electrical cables during design life, both aged and unaged cable specimens should be tested in accordance with IEEE-383. It can be generally thought that the flammability of electrical cables has been increased by wearing as time passed, however the results from fire retardant tests performed in U.S.A and Korea indicate the inconsistent tendency of aging and consequential decrease in flammability. In this study, it is expected that the effective methodology for validation of fire retardant performance would be identified through the review of the results from fire retardant tests.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2006.11a
• /
• pp.353-354
• /
• 2006

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.526.2-526
• /
• 2001

• Nuclear Engineering and Technology
• /
• v.53 no.6
• /
• pp.2038-2045
• /
• 2021

The main objective of this work is to propose a reliable routine standard operation procedures (SOP) for structural health monitoring and diagnosis of nuclear power plants (NPPs). At present, NPPs have monitoring systems that can be used to obtain the quantitative health record of containment (CTMT) buildings through system identification technology. However, because the measurement signals are often interfered with by noise, the identification results may introduce erroneous conclusions if the measured data is directly adopted. Therefore, this paper recommends the SOP for signal screening and the required identification procedures to identify the dynamic characteristics of the CTMT of NPPs. In the SOP, three recommend methods are proposed including the Recursive Least Squares (RLS), the Observer Kalman Filter Identification/Eigensystem Realization Algorithm (OKID/ERA), and the Frequency Response Function (FRF). The identification results can be verified by comparing the results of different methods. Finally, a preliminary CTMT healthy record can be established based on the limited number of earthquake records. It can be served as the quantitative reference to expedite the restart procedure. If the fundamental frequency of the CTMT drops significantly after the Operating Basis Earthquake and Safe Shutdown Earthquake (OBE/SSE), it means that the restart actions suggested by the regulatory guide should be taken in place immediately.

• Journal of the Korean Society of Safety
• /
• v.26 no.5
• /
• pp.99-104
• /
• 2011

An ILRT Interval for a nuclear power plant in Korea was extended from once in five years to once in ten years. Therefore, it is necessary to evaluate risk impact for ILRT interval extensions. In this paper, input data were generated for the reference plants, KSNP, using raw data such as meteorological data, population distribution data and source term data. And, using MACCS II code the risk impact assessment was performed based on the two methodologies of NUREG-1493 and NEI Interim Report. The risk impact derived from an ILRT interval extension was identified not to be significant. It is considered to apply this study and results to making an accident management plan and safety goal, and to the field of public acceptance.

• Nuclear Engineering and Technology
• /
• v.33 no.1
• /
• pp.34-45
• /
• 2001

We have needs to develop a performance analysis system that can be used in domestic nuclear power plants to determine performance status of turbine cycle. We developed new NOPAS system to aid performance analysis of turbine cycle . Procedures of performance calculation are improved using several adaptations from standard calculation algorithms based on ASME (American Society of Mechanical Engineers) PTC (Performance Test Code). Robustness in the performance analysis is increased by verification & validation scheme for measured input data. The system also provides useful aids for performance analysis such as graphic heat balance of turbine cycle and components, turbine expansion lines, automatic generation of analysis reports.

• The Transactions of the Korean Institute of Electrical Engineers P
• /
• v.63 no.2
• /
• pp.61-68
• /
• 2014

Function of battery bank stores energy for DC load in general, and DC power system of the nuclear power plant is used to supply DC loads for safety- featured instrumentation and control such as inverter, class 1E power system control and indication, and station annunciation. Class 1E DC power system must provide a power for the design basis accident conditions, and adequate capacity must be available during loss of AC power and subsequent safe shutdown of the plant. In present, batteries of Class 1E DC power system of the nuclear power plant uses lead-acid batteries. Class 1E batteries of nuclear power plants in Korea are summarized in terms of specification, such as capacity, discharge rate, bank configuration and discharge end voltage, etc. This paper summarizes standards of determining battery size for the nuclear power plant, and analyzes duty cycle for the class 1E DC power system of nuclear power plant. Then, battery cell size is calculated as 2613Ah according to the standard. In addition, this paper analyzes performance test results during past 13 years and shows performance degradation in the battery bank. Performance tests in 2001 and 2005 represent that entire battery cells do not reach the discharge-end voltage. Howeyer, the discharge-end voltage is reached in 14.7% of channel A (17 EA), 13.8% of channel B (16 EA), 5.2% of channel C (6 EA) and 16.4% of channel D (19 EA) at 2011 performance test. Based on the performance test results analysis and size calculation, battery capacity and degradation by age in Korearn nuclear power plant is discussed and would be used for new design.

• Nuclear Engineering and Technology
• /
• v.33 no.5
• /
• pp.468-482
• /
• 2001

As a process of Direct Containment Heating (DCH) issue resolution for Korean Standard Nuclear Power Plants (KSNPs), a containment load/strength assessment with two different approaches, the probabilistic and the deterministic, was performed with all plant-specific and phenomena-specific data. In case of the probabilistic approach, the framework developed to support the Zion DCH study, Two-Cell Equilibrium (TCE) coupled with Latin Hypercubic Sampling (LHS), provided a very efficient tool to resolve DCH issue. In case of the deterministic approach, the evaluation methodology using the sophisticated mechanistic computer code, CONTAIN 2.0 was developed, based on findings from DCH-related experiments or analyses. For three bounding scenarios designated as Scenarios V, Va, and VI, the calculation results of TCE/LHS and CONTAIN 2.0 with the conservatism or typical estimation for uncertain parameters, showed that the containment failure resulted from DCH loads was not likely to occur. To verify that these two approaches might be conservative , the containment loads resulting from typical high-pressure accident scenarios (SBO and SBLOCA) for KSNPs were also predicted. The CONTAIN 2.0 calculations with boundary and initial conditions from the MAAP4 predictions, including the sensitivity calculations for DCH phenomenological parameters, have confirmed that the predicted containment pressure and temperature were much below those from these two approaches, and, therefore, DCH issue for KSNPS might be not a problem.

• Proceedings of the Korean Institute of Building Construction Conference
• /
• 2012.11a
• /
• pp.201-202
• /
• 2012

Recently, High-Strength steel reinforcement has been studied throughout the internal and external. One of the advantages using High-Strength steel reinforcement in construction is the economic effect due to the decreasing of its quantity. Also, another good effect is the increases of workability by reason of reducing the congestion. But, realistically it is not used in nuclear power plant construction site because of the restriction of design standard. The purpose of this report secures the reliability and changes the code through the performance evaluation test of the wall using the high-strength steel reinforcement in nuclear power plant.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.61 no.6
• /
• pp.859-864
• /
• 2012

The computer network plays an important role in modern digital controllers within a safety system of a nuclear power plant. For the reliable and realtime data communication between controllers, this paper proposes a modified high-reliable MVB(multi-function vehicle bus) as a main control network for a safety system of a nuclear power plant. The proposed network supports the state-based communication in order to ensure the deterministic communication latency, and very fast network recovery when the bus master fails compare to the standard MVB. This paper also shows the implementation results using a FPGA-based testbed.