• Journal of Radiation Protection and Research
• /
• v.28 no.3
• /
• pp.215-223
• /
• 2003

Against major release of radioactive material in nuclear power plant, Emergency Planning Zone(EPZ)s are typically established around nuclear power plants to effectively perform the public protective measures. The domestic methodology to determine the size of the EPZ is similar to that of Japan established in 1980, where calculations were based on the conservative accident source term. The objective of this study is to re-evaluate the validity of established EPZ, the area within the radius of $8{\sim}10km$ around domestic nuclear power plants, using the source terms covering full spectrum of accidents obtained from PSA study of ULJIN 3&4. To evaluate the risks of health effects, the computer code MACCS2(MELCOR Accident Consequence Code System2) was used. The result shows that the existing EPZ can reduce the probability of early fatality adequately for most of the source term categories(STCs) except for STC-14 and STC-19. In case of STC-14 and 19, the evacuation distance of 16km and 13km, respectively, are required. These distances can be reduced by improving emergency preparedness since the sensitivity studies for the public protective actions show that the magnitude of early fatality is largely affected by the time delays in notification and evacuation.

• Nuclear Engineering and Technology
• /
• v.53 no.3
• /
• pp.1013-1019
• /
• 2021

This paper reviewed the temporal changes in the public opinions on nuclear energy in Korea with a big data analysis of nuclear energy-related articles and their comments posted on the portal site NAVER. All articles that included at least one of "nuclear energy," "nuclear power plant (NPP)," "nuclear power phase-out," or "anti-nuclear" in their titles or main text were extracted from those posted on NAVER in January 2010-December 2019. First, we performed annual word frequency analysis to identify what words had appeared most frequently in the articles. For that period, the most frequent words were "NPP," "nuclear energy," and "energy." In addition, "safety" has remained in the upper ranks since the Fukushima NPP accident. Then, we performed sentiment analysis of the pre-processed articles. The sentiment analysis showed that positive-tone articles have been reported more frequently than negativetone over the entire analysis period. Last, we performed sentiment analysis of the comments on the articles to examine the public's intention regarding nuclear issues. The analysis showed that the number of negative comments to articles each month-irrespective of positive or negative tone-was always larger than that of positive comments over the entire analysis period.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.19 no.4
• /
• pp.370-377
• /
• 2009

A simple preliminary analysis is often useful to check a validity of design alternatives before the detailed analysis phase in the viewpoint of efficiency. This paper describes a preliminary analysis procedure for the selection among basket design candidates for the spent fuel shipping cask of Korean standard nuclear power plant. As the cask should maintain the structural integrity in hypothetical accident condition, the case of 9 m drop is significantly considered as the worst scenario among the accident conditions in structural design viewpoint in this paper. As basket design options, totally four different types are considered and analyzed in the point of structural integrity at drop impact and weldability for fabrication. As a result, an insertion round plate type with densely spaced supports turns out to be the best in both of the viewpoints, though the weld plate type shows a bit more design margin.

• Plant Journal
• /
• v.12 no.4
• /
• pp.37-43
• /
• 2016

Power system which provides electricity to the accident mitigation load for nuclear power plant should be verified to maintain the proper voltage level under the various loading and source conditions. For this purpose, it was needed to collect the voltage data of safety related buses during operation of the Reactor Coolant Pump(RCP) motor and Component Cooling Water Pump(CCWP) motor, respectively, under the certain loading condition of the plant. The data (such as, voltage, current, power factor) collected from actual measurement were used to modify the existing ETAP model and then the reanalysis was conducted to simulate the testing conditions. Through these actual measurement and analysis, it ensures that the existing electrical system analysis including assumptions and methods was conducted properly. Finally, the voltage of safety related buses was not dropped below the acceptable level, and the discrepancy between two results was within the limit.

• Nuclear Engineering and Technology
• /
• v.54 no.12
• /
• pp.4514-4521
• /
• 2022

In this study, the effects of heat and radiation on the degradation behaviour of fluoroelastomer under simulated normal operation and a severe accident environment were investigated using sequential testing of gamma irradiation and thermal degradation. Tensile properties and Shore A hardness were measured, and thermogravimetric analysis was used to evaluate the degradation behaviour of fluoroelastomer. Fourier transform infrared spectroscopy and X-ray photoelectron spectroscopy were used to characterize the structural changes of the fluoroelastomer. Heat and radiation generated in nuclear power plant break and deform the chemical bonds, and fluoroelastomer exposed to these environments have decreased C-H and functional groups that contain oxygen and double bonds such as C-O, C=O and C=C were generated. These functional groups were formed by auto oxidation by reacting free radicals generated from the cleaved bond with oxygen in the atmosphere. In this auto oxidation reaction, crosslinks were generated where bonded to each other, and the mobility of molecules was decreased, and as a result, the fluoroelastomer was hardened. This hardening behaviour occurred more significantly in the severe accident environment than in the normal operation condition, and it was found that thermal stability decreased with the generation of unstable structures by crosslinking.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.30 no.1 s.244
• /
• pp.90-96
• /
• 2006

Since Risk Based Inspection(RBI) was introduced to quality the risks for jet engine and nuclear power, the technique has been expanded to the area of petrochemical plant. Recently among USA and Europe, window-based computer programme for RBI has been rapidly developed. When local companies procure such program to apply to their plants, it is difficult to build up a proper database as well as expect a good result. In this regards, K-RBI programme accustomed to Korean environments was developed. After completion of the programme in 2004, it was tested by 2 local petrochemical plants and was produced fruitful results. By using these programme, we are expecting accident prevention through risk based planning and implementing of equipment inspection, and saving dollars caused by procuring foreign expensive programme.

• Journal of Radiation Protection and Research
• /
• v.13 no.2
• /
• pp.78-87
• /
• 1988

This article suggests the radionuclides which should be considered more important to the offsite consequence assesment during a nuclear power plant accident. For this purpose, the relative importance to the early health effects of released radionuclides on the major organs during the accident is estimated under the assumption of the same release fraction. The inventories of the 25 elements, 54 nuclides selected in the Reactor Safety Study are calculated by ORIGEN 2 code. The organs of interest in the estimation are G. I. track, bone marrow, thyroid and lung. The result shows the relative potential importance of radionuclides as follows: For G.I. track, Np, Ce, Ru, Y, and Zr are of importance in sequence, Np, I, La, Sr, Ba for bone marrow, I and Te for thyroid, Cm, Ce, Ru, Pu, Zr for lung. In addition to iodine and noble gases, therefore, the potential contribution of those nuclides listed above to the offsite consequences should not be overlooked for some accidents of particular sequence.

• Journal of Radiation Protection and Research
• /
• v.41 no.3
• /
• pp.237-244
• /
• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• Journal of the Korean Society of Systems Engineering
• /
• v.18 no.2
• /
• pp.75-93
• /
• 2022

Machine learning (ML) data-driven meta-model is proposed as a surrogate model to reduce the excessive computational cost of the physics-based model and facilitate the real-time prediction of a nuclear power plant's transient response. To forecast the transient response three machine learning (ML) meta-models based on recurrent neural networks (RNNs); specifically, Long Short Term Memory (LSTM), Gated Recurrent Unit (GRU), and a sequence combination of Convolutional Neural Network (CNN) and LSTM are developed. The chosen accident scenario is a control element assembly withdrawal at power concurrent with the Loss Of Offsite Power (LOOP). The transient response was obtained using the best estimate thermal hydraulics code, MARS-KS, and cross-validated against the Design and control document (DCD). DAKOTA software is loosely coupled with MARS-KS code via a python interface to perform the Best Estimate Plus Uncertainty Quantification (BEPU) analysis and generate a time series database of the system response to train, test and validate the ML meta-models. Key uncertain parameters identified as required by the CASU methodology were propagated using the non-parametric Monte-Carlo (MC) random propagation and Latin Hypercube Sampling technique until a statistically significant database (181 samples) as required by Wilk's fifth order is achieved with 95% probability and 95% confidence level. The three ML RNN models were built and optimized with the help of the Talos tool and demonstrated excellent performance in forecasting the most probable NPP transient response. This research was guided by the Systems Engineering (SE) approach for the systematic and efficient planning and execution of the research.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2008.04a
• /
• pp.425-430
• /
• 2008

To verify the system performance of portable AE leak diagnosis system which can measure with moving conditions, AE activities such as RMS voltage level, AE signal trend, leak rate degree according to AE database, FFT spectrum were measured during operation on total 11 valves of the secondary system in nuclear power plant. AE activities were recorded and analyzed from various operating conditions including different temperature, type of valve, pressure difference, valve size and fluid. The results of this field study are utilized to select the type of sensors, the frequency band for filtering and thereby to improve the signal-to-noise ratio for diagnosis for diagnosis or monitoring of valves in operation. As the final result of application study above, portable type leak diagnosis system by AE was developed. The outcome of the study can be definitely applied as a means of the diagnosis or monitoring system for energy saving and prevention of accident for power plant valve. The purpose of this study is to verify availability of the acoustic emission in-situ monitoring method to the internal leak and operating conditions of the major valves at nuclear power plants. In this study, acoustic emission tests are performed when the pressurized temperature water and steam flowed through glove valve(main steam dump valve) and check valve(main steam outlet pump check valve) on the normal size of 12 and 18 ". The valve internal leak monitoring system for practical field was designed. The acoustic emission method was applied to the valves at the site, and the background noise was measured for the abnormal plant condition. To improve the reliability, a judgment of leak on the system was used various factors which are AE parameters, trend analysis, frequency analysis, voltage analysis and amplitude analysis of acoustic signal emitted from the valve operating condition internal leak.