• Fire Science and Engineering
• /
• v.30 no.4
• /
• pp.20-26
• /
• 2016

When the fire takes place in Nuclear Powr Plants(NPPs), the reactor should achieve and maintain safe shut-down conditions and minimize the radioactive material released to the environment. The U.S. Nuclear Regulatory Commission (NRC) has issued numerous generic communications related to fire protection over the past 20 years, after it issued its requirements in the Fire Protection Rule set forth in Title 10, Section 50.48 of the Code of Federal Regulations (10 CFR 50.48) and Appendix R to the 10 CFR 50. The and Nuclear Energy Institute (NEI) has developed a Methodology for Risk Informed Fire Safe-Shutdown Analysis, which is related to the Deterministic Method for Multiple Spurious Operations solutions. The aim of this study was to identify, achieve, and maintain Post-Fire Safe-Shutdown of the Main Control Room (MCR) of the CANDU reactor, even though one train of the multiple Safety Structures, Systems, and Components (SCCs) fail by the technical specification and analysis method.

• Journal of the Korean Society of Systems Engineering
• /
• v.16 no.2
• /
• pp.27-37
• /
• 2020

The Fukushima Daiichi accident in 2011 revealed some vulnerabilities of existing Nuclear Power Plants (NPPs) under extended Station Blackout (SBO) accident conditions. One of the key Severe Accident Management (SAM) strategies developed post Fukushima accident is the In-Vessel Retention (IVR) Strategy which aims to retain the structural integrity of the Reactor Pressure Vessel (RPV). RELAP/SCDAPSIM/MOD3.4 is selected to predict the thermal-hydraulic response of APR1400 undergoing an extended SBO. To assess the effectiveness of the IVR strategy, it is essential to quantify the underlying uncertainties. In this work, both the epistemic and aleatory uncertainties are considered to identify the success window of the IVR strategy. A set of in-vessel relevant phenomena were identified based on Phenomena Identification and Ranking Tables (PIRT) developed for severe accidents and propagated through the thermal-hydraulic model using Wilk's sampling method. For this work, a Systems Engineering (SE) approach is applied to facilitate the development process of assessing the reliability and robustness of the APR1400 IVR strategy. Specifically, the Kossiakoff SE method is used to identify the requirements, functions and physical architecture, and to develop a design verification and validation plan. Using the SE approach provides a systematic tool to successfully achieve the research goal by linking each requirement to a verification or validation test with predefined success criteria at each stage of the model development. The developed model identified the conditions necessary for successful implementation of the IVR strategy which maintains the vessel integrity and prevents a melt-through.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.15 no.6
• /
• pp.11-18
• /
• 2011

Accurate modal identification analysis is required to reasonably perform a seismic qualification of safety-related electric equipment installed in nuclear power plants (NPPs). This study evaluates a variation of the modal properties of an electric equipment cabinet structure in NPPs according to the excitation levels. For the study, an actual electric equipment cabinet was selected as a specimen and was dynamically tested by using a portable exciter in accordance with the level of input vibration energy. Tests were classified into two sets: with-door cases, and without-door cases. Frequency response functions were computed from the signals of the acceleration responses and input motions measured from the vibration tests. A polynomial curve fitting algorithm was used to extract the modal properties from the frequency response functions. This study reviews the variation of the modal properties according to the variation of the excitation levels. The results of the study show that the modal frequencies and the modal dampings of the object specimen varies nonlinearly according to the excitation level of the test motion. Attaching the door increases the modal damping of the cabinet.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.30 no.4
• /
• pp.330-337
• /
• 2010

In this paper, 3-dimensional electromagnetic numerical analysis is performed about the eddy current(EC) array probe characteristic which is the next generation probe for accurate diagnosis of steam generator(SG) in nuclear power plants(NPPs). ASME(American Society of Mechanical Engineers) Standard and X-probe combo calibration standard tube are selected for acquisition of eddy current testing(ECT) signals and this result of compared with the real test signals for reasonability of result. Based on the analysis result of calibration standard tube, ECT signals that are about the defects of pitting, stress corrosion cracking(SCC), multiple SCC and wear is obtained. Material of specimen was Inconel 600 which is usually used for SG tubes in NPPs. The operation frequency of 300 kHz were used. The signal characteristics could be observed according to the various defects. The results in this paper can be helpful when the ECT signals from EC array probe are evaluated and analyzed.

• Economic and Environmental Geology
• /
• v.56 no.5
• /
• pp.501-514
• /
• 2023

Forest fires can generate numerous pollutants through the combustion of vegetation and cause serious environmental problems. The global warming and climate change will increase the frequency and scale of forest fires across the world. In Korea, many nuclear power plants (NPPs) are located in the East Coast where large-scale forest fires frequently occur. Therefore, understanding the sorption and transport characteristics of radionuclides in the forest fire areas is required against the severe accidents in NPPs. This article reviewed the physiochemical changes and contamination of groundwater and soil environments after forest fires, and discussed sorption and transport of radionuclides in the subsurface environment of burned forest area. We considered the geochemical factors of subsurface environment changed by forest fire. Moreover, we highlighted the need for studies on changes and contamination of subsurface environments caused by forest fires to understand more specific mechanisms.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.18 no.3
• /
• pp.141-150
• /
• 2014

A methodology to evaluate the seismic performance of interface piping systems that cross the isolation interface in the seismically isolated nuclear power plant (NPP) was developed. The developed methodology was applied to the safety-related interface piping system to demonstrate the seismic performance of the target piping system. Not only the seismic performance for the design level earthquakes but also the performance for the beyond design level earthquakes were evaluated. Two artificial seismic ground input motions which were matched to the design response spectra and two historical earthquake ground motions were used for the seismic analysis of piping system. The preliminary performance evaluation results show that the excessive relative displacements can occur in the seismically isolated piping system. If the input ground motion contained relatively high energy in the low frequency region, we could find that the stress response of the piping system exceed the allowable stress level even though the intensity of the input ground motion is equal to the design level earthquake. The structural responses and seismic performances of piping system were varied sensitively with respect to the intensities and frequency contents of input ground motions. Therefore, for the application of isolation system to NPPs and the verification of the safety of piping system, the seismic performance of the piping system subjected to the earthquake at the target NPP site should be evaluated firstly.

• Journal of the Korea institute for structural maintenance and inspection
• /
• v.8 no.3
• /
• pp.169-176
• /
• 2004

The Mechanistic model, considering expansion stress, coefficient of diffusion etc. to time, is applied to predict the deterioration of concrete structures of the nuclear power plant(NPP) due to sulfate attack. Mix design for the test was three kinds of specified compressive strength 385, 280 and $210kgf/cm^2$ which are used to construct NPPs and cement was type I and V. The immersion test was performed with 10% $Na_2SO_4$ solution to cement type and strength for a year. The coefficient of diffusion on each concrete mix is calculated based on the results of immersion test, and it is used for predicting the sulfate attack of the concrete structures of NPP. The coefficient of diffusion of the target concrete ranged $0.5763{\sim}3.9002{\times}10^{-12}m^2/sec.$, and the sulfate attack rate of concrete structures of the NPP was predicted as 0.1~7.1 mm/year.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.23 no.6
• /
• pp.289-299
• /
• 2019

The tendency to use a probabilistic design method rather than a deterministic design method for the design of nuclear power plants (NPPs) will increase because their safety should be considered and strictly controlled in relation to various causes of damage. The distance between a seismically isolated NPP structure and a moat wall is called the clearance to stop. The clearance to stop is obtained from the 90th percentile displacement response of a seismically isolated NPP subject to a beyond design basis earthquake (BDBE) in the probabilistic design method. The purpose of this study is to analyze the effects of heating and buckling effects on the 90th percentile displacement response of a lead-rubber bearing (LRB) subject to a BDBE. The analysis results show that considering the heating and buckling effects to estimate the clearance to stop is conservative in the evaluation of the 90th percentile displacement response. If these two effects are not taken into account in the calculation of the clearance to stop, the underestimation of the clearance to stop causes unexpected damage because of an increase in the collision probability between the moat wall and the seismically isolated NPP.

• Journal of the Korea institute for structural maintenance and inspection
• /
• v.23 no.3
• /
• pp.136-142
• /
• 2019

Nuclear power plant concrete structures are in contact with the coast, and durability due to chloride attack is very important because it is used as cooling water by taking seawater. For this purpose, a 3-year long-term saltwater immersion test was carried out to evaluate chloride ion diffusion coefficient and age apponent (m) The m values of the foundation with 4,000 class was 0.35 ~ 0.39, similar to KCI or ACI suggested values. essential service water constructions and tunnels of 5,000 class were 0.44 ~ 0.53 and 6,000 class, and 0.62 of reactor containment buildings were similar to the proposed values of FIB. As a result of the prediction of the service life with the measured age coefficient, all the safety related concrete structures of the nuclear power plants satisfied the service life of more than 60 years.

• Proceedings of the KSME Conference
• /
• 2008.11a
• /
• pp.192-192
• /
• 2008

As the lifetime of nuclear power plants (NPPs) reaches design life, the probability for fatal accidents increases. Most of accidents are known to be caused by degradation of mechanical components. Pressure tubes are the most important components in CANDU reactor. They are subjected to various aging mechanisms such as delayed hydride cracking (DHC), irradiation and corrosion, etc. Therefore, the integrity of pressure tube is key concern in CANDU reactor. Up to recently, conventional deterministic approaches have been utilized to evaluate the integrity of components. However, there are many uncertainties to prevent a rational evaluation. The objective of this paper is to assess the failure probability of pressure tube in CANDU. To do this, probability fracture mechanics (PFM) analysis based on the Genetic Algorithm (GA) is performed. For the verification of the analysis, a comparison of the PFM analysis using a commercial code and mathematical method is carried out.