• Proceedings of the KOSOMBE Conference
• /
• v.1997 no.05
• /
• pp.92-94
• /
• 1997

By measuring the increase of regional cerebral blood flow (rCBF) during the activation tasks, we can describe the brain regions that participate in certain specific functions. In this study, we composed the functional maps of verbal and nonverbal memory by performing the rCBF positron emission tomography (PET) activation studies and analyzing the differences between control and each activation state. Successive four tasks, which consist of one control state and three different activation tasks, were performed on 6 normal volunteers. All images were spatially normalized on standard atlas and the differences between control and activation states were statistically analyzed. The verbal memory activated predominantly left-sided structures, especially left superior temporal cortex, and the nonverbal short-term memory activated the right frontal cortex. Also, some regions ,where is thought to be related with short-term memory system, such as cingulate gyrus and hippocampus were activated. We conclude that biological validity of the brain regions for verbal and nonverbal memory could be tested using rCBF PET imaging technique and statistical analysis.

• Nuclear Engineering and Technology
• /
• v.42 no.5
• /
• pp.552-561
• /
• 2010

This study investigates the solubility of neptunium (Np) in the deep natural groundwater of the Korea Atomic Energy Research Institute Underground Research Tunnel (KURT). According to a Pourbaix diagram (pH-$E_h$ diagram) that was calculated using the geochemical modeling program PHREEQC 2.0, the redox potential and the carbonate ion concentration both control the solubility of neptunium. The carbonate effect becomes pronounced when the total carbonate concentration is higher than $1.5\;{\times}\;10^{-2}$ M at $E_h$ = -200 mV and the pH value is 10. Given the assumption that the solubility-limiting stable solid phase is $Np(OH)_4(am)$ under the reducing condition relevant to KURT, the soluble neptunium concentrations were in the range of $1\;{\times}\;10^{-9}$ M to $3\;{\times}\;10^{-9}$ M under natural groundwater conditions. However, the solubility of neptunium, which was calculated with the formation constants of neptunium complexes selected in an OECD-NEA TDB review, strongly deviates from the value measured in natural groundwater. Thus, it is highly recommended that a prediction of neptunium solubility is based on the formation constants of ternary Np(IV) hydroxo-carbonato complexes, even though the presence of those complexes is deficient in terms of the characterization of neptunium species. Based on a comparison of the measurements and calculations of geochemical modeling, the formation constants for the "upper limit" of the Np(IV) hydroxo-carbonato complexes, namely $Np(OH)_y(CO_3)_z^{4-y-2z}$, were appraised as follows: log $K^{\circ}_{122}\;=\;-3.0{\pm}0.5$ for $Np(OH)_2(CO_3)_2^{2-}$, log $K^{\circ}_{131}\;=\;-5.0{\pm}0.5$ for $Np(OH)_3(CO_3)^-$, and log $K^{\circ}_{141}\;=\;-6.0{\pm}0.5$ for $Np(OH)_4(CO_3)^{2-}$.

• Nuclear Engineering and Technology
• /
• v.41 no.4
• /
• pp.427-446
• /
• 2009

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical $CO_2$ Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.4
• /
• pp.263-269
• /
• 2004

During the last two years, Site Information and Total Environmental database management System (SITES) ver. 1.0 has been developed for the systematic SITES Database Module (SDM), which includes site information, facility information and environmental information. The SITES includes the module for site environmental monitoring system and safety assessment (M&A) system for the nuclear facility. SITES is expected to be an effective system for the radioactive waste disposal management facility. Currently, SITES ver.2.0 is under development after the SITES ver.1.0 that is focused on the M&A system. The main purpose of this paper is to introduce and try to account for the major development in the concept of SEMS sub-module of the M&A module. The SEMS is purposed of development of the program for real time environmental monitoring, prediction, and automatic alarm system using SITES Database and related information.

• Nuclear Engineering and Technology
• /
• v.36 no.1
• /
• pp.112-120
• /
• 2004

The purpose of this paper is to analyze the piping failure frequency for the main feedwater system in domestic nuclear power plants(NPPs) for the application to an in-service inspection(ISI), leak before break(LBB) concept, aging management program(AMP), and probabilistic safety analysis(PSA). First, a database was developed for piping failure events in domestic NPPs, and 23 domestic piping failure events were collected. Among the 23 events, 12 locations of wall thinning due to flow accelerated corrosion(FAC) were identified in the main feedwater system in 4 domestic WH 3-loop NPPs. Two types of the piping failure frequency such as the damage frequency and rupture frequency were considered in this study. The damage frequency was calculated from both the plant population data and damage(s) including crack, wall thinning, leak, and/or rupture, while the rupture frequency was estimated by using both the well-known Jeffreys method and a new method considering the degradation due to FAC. The results showed that the damage frequencies based on the number of the base metal piping susceptible to FAC ranged from $1.26{\times}10^{-3}/cr.yr\;to\;3.91{\times}10^{-3}/cr.yr$ for the main feedwater system of domestic WH 3-loop NPPs. The rupture frequencies obtained from the Jeffreys method for the main feedwater system were $1.01{\times}10^{-2}/cr.yr\;and\;4.54{\times}10^{-3}/cr.yr$ for the domestic WH 3-loop NPPs and all the other domestic PWR NPPs respectively, while those from the new method considering the degradation were higher than those from the Jeffreys method by about an order of one.

• Journal of Energy Engineering
• /
• v.23 no.2
• /
• pp.183-190
• /
• 2014

A Commercial Grade Item Dedication(CGID) for Korean operating nuclear power plants has been implemented since 2012. The CGID implementation and strategies for Korea are established as follows: CGID policy establishment, R&D of a specific methodologies of CGID, enrollment of third party organizations for CGID work, CGID program establishment for enrolled suppliers, establishment of training courses for certification, and CGID process development for quality class Q and A. Consequently, it is expected that these activities are enable to enhance the reliability and the safety of components and/or parts in nuclear power plants.

• Journal of Radiation Protection and Research
• /
• v.44 no.1
• /
• pp.8-14
• /
• 2019

Background: Management of an agricultural food product system following a nuclear accident is indispensable for reducing radiation exposure due to ingestion of contaminated food. The present study analyzes the effect of agricultural countermeasures on ingestion dose following a nuclear accident. Materials and Methods: Agricultural countermeasures suitable for domestic farming environments were selected by referring to the countermeasures applied after the Fukushima accident in Japan. The avertable ingestion doses that could be obtained by implementing the selected countermeasures were calculated using the Korean Agricultural Countermeasure Analysis Program (K-ACAP) to investigate the efficiency of each countermeasure. Results and Discussion: Of the selected countermeasures, the management of crops was effective when radionuclide deposition occurred during the growing season of plants. Treatment by soil additive and topsoil removal was effective when deposition occurred during the nongrowing season of plants. The disposal of milk was not effective owing to the small contribution of milk to the overall ingestion dose. Clean feeding of livestock was effective when deposition occurred during the growing season of fodder plants such as pasture and rice-straw. Finally, the effect of food restriction increased with the soil deposition density of radionuclide. The practical effect of countermeasures was very small when the avertable ingestion dose was absolutely low. Conclusion: The agricultural countermeasures selected to reduce the radionuclide ingestion dose after a nuclear accident must be made appropriate by considering the accident situation, such as the soil deposition density of the radionuclide and the deposition date in relation to farming cycles.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.34 no.6
• /
• pp.385-392
• /
• 2021

Interest in the seismic performance of nuclear facilities under strong earthquakes has increased because their nonlinear response is important. In this paper, we proposed appropriate parameters for the nonlinear finite element analysis of a concrete material model, for a reinforced concrete (RC) shear wall in nuclear facilities: maximum tensile strength, dilation angle, and damage parameter. The study of the effects of the important parameters, on the nonlinear behavior and shear failure mode of the RC shear wall having low aspect ratio, was conducted using ABAQUS finite element analysis program. Based on the study results the nonlinear response of a nuclear reactor containment building (RCB) subjected to a strong earthquake was evaluated using nonlinear time-history analysis.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.993-1001
• /
• 2024

The TANDEM project is a European initiative funded under the EURATOM program. The project started on September 2022 and has a duration of 36 months. TANDEM stands for Small Modular ReacTor for a European sAfe aNd Decarbonized Energy Mix. Small Modular Reactors (SMRs) can be hybridized with other energy sources, storage systems and energy conversion applications to provide electricity, heat and hydrogen. Hybrid energy systems have the potential to strongly contribute to the energy decarbonization targeting carbon-neutrality in Europe by 2050. However, the integration of nuclear reactors, particularly SMRs, in hybrid energy systems, is a new R&D topic to be investigated. In this context, the TANDEM project aims to develop assessments and tools to facilitate the safe and efficient integration of SMRs into low-carbon hybrid energy systems. An open-source "TANDEM" model library of hybrid system components will be developed in Modelica language which, by coupling, will extend the capabilities of existing tools implemented in the project. The project proposes to specifically address the safety issues of SMRs related to their integration into hybrid energy systems, involving specific interactions between SMRs and the rest of the hybrid systems; new initiating events may have to be considered in the safety approach. TANDEM will study two hybrid systems covering the main trends of the European energy policy and market evolution at 2035's horizon: a district heating network and power supply in a large urban area, and an energy hub serving energy conversion systems, including hydrogen production; the energy hub is inspired from a harbor-like infrastructure. TANDEM will provide assessments on SMR safety, hybrid system operationality and techno-economics. Societal considerations will also be encased by analyzing European citizen engagement in SMR technology safety.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.21 no.3
• /
• pp.457-469
• /
• 1997

A nuclear spent fuel shipping cask is required by IAEA and domestic regulations to withstand a 9m free drop condition. In this paper, the structural analysis under the 9m side drop condition was performed to understand the dynamic impact behavior and to evaluate the safety of the cask for 7 PWR nuclear spent fuel assemblies. The analysis result was compared with the measured value of the 9m side drop test for the 1/3 scaled-down model and the accuracy of the 3D analysis was confirmed. Analysis in accordance with the diameter of impact limiters for the proto-type cask were performed. Through the analysis, the impact behaviors due to the side drop and the effects dependent on the diameter of impact limiters were grasped. Maximum stress intensities on each part of the cask were respectively calculated by using the stress evaluation program and the structural safety of the cask was finally evaluated in accordance with the regulations.