• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.88-103
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• 1999

The purposes of this study are to understand the severe accident phenomena, to establish the simulation method for the experimental test, and to assess the current models in MELCOR for future improvement. This paper presents the results of the PHEBUS FPT1 post test analysis using MELCOR computer code, version 1.8.4. The entire PHEBUS facility has been modeled; the core, the primary circuit including the steam generator, and the containment vessel. Both the thermal hydraulic and the fission product behavior have been investigated. The code simulation results of the thermal hydraulic behavior show good agreement with the experimental data, The fission product release and transport are calculated using the CORSOR models in MELCOR code and the results will be compared with the experiment when the experimental data are available.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.283-288
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• 1996

Core support barrel (CSB) assembly is one of the most important reactor internals structures supporting and protecting the nuclear core during normal operation and faulted events. For Yonggwang 3 and 4 (YGN 3&4), the adequacy of the analytical response prediction of reactor internals for flow induced vibration was demonstrated through the comprehensive vibration assessment program (CVAP) performed during hot functional test. Besides, the vibration characteristics of the CSB of operating nuclear power plant can be examined via the excore neutron noise monitoring signal. In this paper data from YGN 4 analyses, CVAP, and neutron noise monitoring system are compared and evaluated. In general, the results are comparable each other and conservative enough to ensure sufficient design margin and structural integrity. Further investigations on the modelling and analyses procedure are recommended to utilize the experimental results to the maximum extent. And collection of the neutron noise data is desired to serve as a baseline information.

• Proceedings of the Korean Magnetic Resonance Society Conference
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• 2002.08a
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• pp.72-72
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• 2002

In most traditional sorption study in environmental conditions, experimental sorption data have been measured and interpreted by empirical ways such as partition coefficient and sorption isotherms. A mechanistic understanding of heavy metal interactions with various minerals (metal oxides, clay minerals) in aqueous medium is required to describe the behavior of radioactive metal ions in the environment. Various spectroscopic methods provide direct or indirect information on sorption mechanisms involved. We applied EPR (Electron Paramagnetic Resonance) spectroscopy to investigate the nature of metal ion sorption at water/mineral interfaces using Cu(II) as a spin probe. The major sorbed species and their motional state was identified by their EPR spectra. They showed distinct signals due to their strength of binding, local structure and motional state. The EPR results together with macroscopic sorption data show that sorption involved at least three different mechanisms depending on chemical environments (1).

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3336-3346
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• 2022

In general, a number of severe accident scenarios derived from Level 2 probabilistic safety assessment (PSA) are typically grouped into several categories to efficiently evaluate their potential impacts on the public with the assumption that scenarios within the same group have similar source term characteristics. To date, however, grouping by similar source terms has been completely reliant on qualitative methods such as logical trees or expert judgements. Recently, an exhaustive simulation approach has been developed to provide quantitative information on the source terms of a large number of severe accident scenarios. With this motivation, this paper proposes a machine learning-based categorization method based on exhaustive simulation for grouping scenarios with similar accident consequences. The proposed method employs clustering with an autoencoder for grouping unlabeled scenarios after dimensionality reductions and feature extractions from the source term data. To validate the suggested method, source term data for 658 severe accident scenarios were used. Results confirmed that the proposed method successfully characterized the severe accident scenarios with similar behavior more precisely than the conventional grouping method.

• Analytical Science and Technology
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• v.13 no.2
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• pp.179-188
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• 2000

Trace elements in air samples artificially loaded on filters with urban dust and the bulk material of urban dust as an environmental sample were determined non-destructively using instrumental neutron activation analysis. Standard reference material (Urban Dust, SRM 1648) of the National Institute of Standard and Technology was used for the analytical quality control. The relative error for 37 elements was less than 15% and the standard deviation was less than 10%. 29 elements in the urban dust and 21 elements in the loaded filter sample were determined respectively. To evaluate the proficiency and reliability of the measurement, data intercomparison was performed and 39 analytical laboratories participated in the analysis using different analytical methods; neutron activation analysis, particle induced X-ray emission analysis, X-ray fluorescence analysis and atomic absorption spectrometry. Z-scores were calculated using the standard deviation of the laboratorie's mean as target standard deviation, and a good result was obtained that the values fall between -1 and +1 except some elements.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.81-92
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• 2003

In order to predict the dynamic behaviors of uranium and cobalt in a fixed bed at various influent pH values of liquid waste, the adsorption system is regarded as a multi-component adsorption between each ionic species in the solution. Langmuir isotherm parameters of each species were extracted by incorporating equilibrium data with the solution chemistry of the uranium and cobalt using IAST. Prediction results were in good agreement with the experimental data, except for a high concentration and pH. Although there was some limitations in predicting the cobalt adsorption, this method may be useful in analyzing a complex adsorption system where various kinds of ionic species exist in a solution.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.33-46
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• 2015

During the past 10 years, the Korea Atomic Energy Research Institute (KAERI) has performed a study to control hydrogen gas in the containment of the nuclear power plants. Before the Fukushima accident, analytical activities for gas distribution analysis in experiments and plants were primarily conducted using a multidimensional code: the GASFLOW. After the Fukushima accident, the COM3D code, which can simulate a multidimensional hydrogen explosion, was introduced in 2013 to complete the multidimensional hydrogen analysis system. The code validation efforts of the multidimensional codes of the GASFLOW and the COM3D have continued to increase confidence in the use of codes using several international experimental data. The OpenFOAM has been preliminarily evaluated for APR1400 containment, based on experience from coded validation and the analysis of hydrogen distribution and explosion using the multidimensional codes, the GASFLOW and the COM3D. Hydrogen safety in nuclear power has become a much more important issue after the Fukushima event in which hydrogen explosions occurred. The KAERI is preparing a large-scale test that can be used to validate the performance of domestic passive autocatalytic recombiners (PARs) and can provide data for the validation of the severe accident code being developed in Korea.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.424-435
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• 2016

Background: The International Commission on Radiological Protection (ICRP) recommendations and the Federal Guidance Report (FGR) published by the U.S. Environmental Protection Agency (EPA) have been widely applied worldwide in the fields of radiation protection and dose assessment. The dose conversion coefficients of the ICRP and FGR are widely used for assessing exposure doses. However, before the coefficients are used, the user must thoroughly understand the derivation process of the coefficients to ensure that they are used appropriately in the evaluation. Materials and Methods: The ICRP provides recommendations to regulatory and advisory agencies, mainly in the form of guidance on the fundamental principles on which appropriate radiological protection can be based. The FGR provides federal and state agencies with technical information to assist their implementation of radiation protection programs for the U.S. population. The system of radiation dose assessment and dose conversion coefficients in the ICRP and FGR is reviewed in this study. Results and Discussion: A thorough understanding of their background is essential for the proper use of dose conversion coefficients. The FGR dose assessment system was strongly influenced by the ICRP and the U.S. National Council on Radiation Protection and Measurements (NCRP), and is hence consistent with those recommendations. Moreover, the ICRP and FGR both used the scientific data reported by Biological Effects of Ionizing Radiation (BEIR) and United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) as their primary source of information. The difference between the ICRP and FGR lies in the fact that the ICRP utilized information regarding a population of diverse races, whereas the FGR utilized data on the American population, as its goal was to provide guidelines for radiological protection in the US. Conclusion: The contents of this study are expected to be utilized as basic research material in the areas of radiation protection and dose assessment.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.760-768
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• 2017

The transition temperature shift (TTS) of the reactor pressure vessel materials is an important factor that determines the lifetime of a nuclear power plant. The prediction of the TTS at the end of a plant's lifespan is calculated based on the equation of Regulatory Guide 1.99 revision 2 (RG1.99/2) from the US. The fluence factor in the equation was expressed as a power function, and the exponent value was determined by the early surveillance data in the US. Recently, an advanced approach to estimate the TTS was proposed in various countries for nuclear power plants, and Korea is considering the development of a new TTS model. In this study, the TTS trend of the Korean surveillance test results was analyzed using a nonlinear regression model and a mixed-effect model based on the power function. The nonlinear regression model yielded a similar exponent as the power function in the fluence compared with RG1.99/2. The mixed-effect model had a higher value of the exponent and showed superior goodness of fit compared with the nonlinear regression model. Compared with RG1.99/2 and RG1.99/3, the mixed-effect model provided a more accurate prediction of the TTS.