• Fire Science and Engineering
• /
• v.26 no.6
• /
• pp.38-44
• /
• 2012

The purpose of fire protection for the nuclear power plants (NPPs) is to ensure safe shutdown state of the reactor, to minimize the release of radioactive materials to the environment, to provide physical safety of the on-site personnel, and to limit the property damage. Fire protection and extinguishing equipments are one of the important protection measures based on the defense-in-depth concept, which can promptly detect and control and extinguish those fires that do occur, thereby limiting fire damage. However, a separate evaluation process might be additionally necessary for the construction permit and operating license because the fire protection laws of the NEMA for installation standards of the fire protection systems is not fully characterized for the NPPs. It is also not easy to implement the regulations such as the performance based design concept for fire protection system of the NPPs which are characterized for a relatively low density of employee. This study suggests a guideline for the improvement of the technical standards for fire protection systems of the NPPs by evaluating the fundamental problems drawn by reviewing laws and regulatory guides relevant to fire protection and by evaluating the applicability of the KEPIC FPN in domestic nuclear power plants.

• The Korean Journal of Applied Statistics
• /
• v.29 no.7
• /
• pp.1385-1398
• /
• 2016

A pressurizer is a major equipment system in a nuclear power plant (NPP) and controls the reactor cooling system pressure within the allowable range. Faults in the pressurizer can be critical to the NPP; therefore, early fault detection in the pressurizer is significant for NPP safety. This study applies Mahalanobis Taguchi system (MTS), which is one of the promising pattern classification methods, based on the Mahalanobis distance concept and Taguchi quality engineering theory to the early fault detection problem of the pressurizer pressure control system. We conducted experiments using data from full scope NPP simulator based on a pressurizer pressure transmitter faults scenario to validate the faults detection performance of MTS. As a result, MTS can rapidly detect the faults compared to conventional faults detection based on single sensor monitoring.

• Journal of Energy Engineering
• /
• v.25 no.4
• /
• pp.124-132
• /
• 2016

In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

• Nuclear Engineering and Technology
• /
• v.27 no.5
• /
• pp.645-660
• /
• 1995

The loss of Residual Heat Removal System (RHRS) event during reduced inventory operation for the Korean Standard Nuclear Power Plants (KSNPPS) is simulated by RELAP5/MOD3 and RELAP5/MOD3.1 Tn cases are considered : Base case for an intact Reactor Coolant System (RCS) with no tent and a vent case for an open system. Comparative simulations of base case are peformed by RELAP5/MOD3 and RELAP5/MOD3. 1 computer codes. The results of too simulations are generally in good qualitative and quantitative agreement. However, since the results of RELAP5/MOD3 simulation reveals the deficiency of RELAP5/MOD3 wall heat model, the RELAP5/AOD3.1 computer code is used for the simulation of the vent case. The analysis result of base case show that two steam generators are insufficient to remove decay heat at one day after shutdown, where the RCS is closed. The RCS pressure increased continuously and reached the RCS temporary boundaries design pressure of 0.24 MPa around 4,000 seconds. In the vent case with a flow capacity equivalent to three times the capacity of Pressurizer Safety Valve (PSV), it is shown that the RCS Pressure does not reach 0.24 MPa and core uncovery does not occur until 10,000 seconds. The detailed discussions on the results of this study suggest the feasibility of RELAP5/AOD3.1 as an analysis tool for the simulation of the loss of RHRS event at reduced inventory operation. The results of this study also provide insight for the determination of proper vent capacity.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.24 no.1 s.173
• /
• pp.206-214
• /
• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tube s. the integrity evaluation must be carried out. and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire: integrity evaluation process. For this reason. an integrity evaluation system, which provides efficient of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). This system does not only provide various databases including the 3-D finite element analysis results on pressure tubes, inspection data and design specifications but also is compatible with other commercial database software. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Nuclear Engineering and Technology
• /
• v.44 no.8
• /
• pp.919-928
• /
• 2012

The applications of computers and communication system and network technologies in nuclear power plants have expanded recently. This application of digital technologies to the instrumentation and control systems of nuclear power plants brings with it the cyber security concerns similar to other critical infrastructures. Cyber security risk assessments for digital instrumentation and control systems have become more crucial in the development of new systems and in the operation of existing systems. Although the instrumentation and control systems of nuclear power plants are similar to industrial control systems, the former have specifications that differ from the latter in terms of architecture and function, in order to satisfy nuclear safety requirements, which need different methods for the application of cyber security risk assessment. In this paper, the characteristics of nuclear power plant instrumentation and control systems are described, and the considerations needed when conducting cyber security risk assessments in accordance with the lifecycle process of instrumentation and control systems are discussed. For cyber security risk assessments of instrumentation and control systems, the activities and considerations necessary for assessments during the system design phase or component design and equipment supply phase are presented in the following 6 steps: 1) System Identification and Cyber Security Modeling, 2) Asset and Impact Analysis, 3) Threat Analysis, 4) Vulnerability Analysis, 5) Security Control Design, and 6) Penetration test. The results from an application of the method to a digital reactor protection system are described.

• Nuclear Engineering and Technology
• /
• v.46 no.4
• /
• pp.547-556
• /
• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.14 no.7
• /
• pp.3121-3126
• /
• 2013

The integrity evaluation of pipes in nuclear power plant are essential for the safety of reactor vessel, and integrity must be assured when flaws are found. Accurate stress intensity analyses and crack growth rate data of surface-cracked components are needed for reliable prediction of their fatigue life and fracture strengths. Fatigue design and life assessment are the essential technologies to design the structures such as pipe, industrial plant equipment and so on. The effect of crack spacing on stress intensity factor K values was studied using three-dimensional finite element method (FEM). For the case of cylinder under internal pressure, a significant increase in K values observed at the deepest point of the surface crack. Also, this paper describes the fatigue analysis for cracked structures submitted to bending loads.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.22 no.4
• /
• pp.245-251
• /
• 2018

In this paper, we address some issues in existing seismic hazard closed-form equations and present a novel seismic hazard equation form to overcome these issues. The presented equation form is based on higher-order polynomials, which can well describe the seismic hazard information with relatively high non-linearity. The accuracy of the proposed form is illustrated not only in the seismic hazard data itself but also in estimating the annual probability of failure (APF) of the structural systems. For this purpose, the information on seismic hazard is used in representative areas of the United States (West : Los Angeles, Central : Memphis and Kansas, East : Charleston). Examples regarding the APF estimation are the analyses of existing platform structure and nuclear power plant problems. As a result of the numerical example analyses, it is confirmed that the higher-order-polynomial-based hazard form presented in this paper could predict the APF values of the two example structure systems as well as the given seismic hazard data relatively accurately compared with the existing closed-form hazard equations. Therefore, in the future, it is expected that we can derive a new improved APF function by combining the proposed hazard formula with the existing fragility equation.

• Journal of the Korean Society of Visualization
• /
• v.14 no.2
• /
• pp.52-60
• /
• 2016

The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.