• The Korean Journal of Food And Nutrition
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• v.28 no.2
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• pp.202-218
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• 2015

Many countries have introduced new imported food safety measures, following the accident at Fukushima Daiichi Nuclear Power Station. This study was conducted to evaluate the measures contents and effects on food trades values. Eight percent of members were notified the introduced measures to the World Trade Organization. The measures' contents were banning imports, enhancing inspection and adding certification requirement. The covered regions were some prefectures, entire Japan or all affected countries. European Union introduced a measure that subjecting foods originating from 12 prefectures to import at designated ports with required certification. The measures were amended 8 times until March 2014 to apply listed foods from 15 prefectures. The trade value of fishery products and miscellaneous foods were affected. Australia introduced a measure that required additional inspection of dairy, fishery and plants products from 13 prefectures with subsequent amendments. The trade value had no effect in tested foods. Chinese Taipei introduced a temporary import ban for all foods from 6 prefectures. Trade values for fruits were affected. The United States issued an import alert for detention without examination for listed prefectures and goods without introducing new measures. Although no specific products were affected, trade values for all foods were affected.

• Journal of Korea Society of Digital Industry and Information Management
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• v.16 no.2
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• pp.61-68
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• 2020

Nuclear power plants have a lower cost of power generation, and they are more eco-friendly than other power generation plants. Also, we need to prepare nuclear plant accidents because of their severe damage. In the event of a safety accident, such as a radiation leak, by applying a wireless sensor network to a nuclear power plant, many sensor nodes can be used to monitor radiation and transmit information to an external base station to appropriately respond to the accident. However, applying a wireless sensor network to nuclear power plants requires routing protocols that consider the sensor network size and bypass obstacles such as plant buildings. In general, the hierarchical-based routing protocols are efficient in energy consumption. In this study, we look into the problems that may occur if hierarchical-based routing protocols are applied to nuclear power plants and propose improved routing protocols to solve these problems. Simulation results show that the proposed routing protocol is more effective in energy consumption than the existing LEACH protocol.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2859-2865
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• 2021

One of the important issues raised from the Fukushima-Daiichi accident is the safety of multi-unit sites when simultaneous accidents occur at the site and recently a multi-unit PSA methodology is being developed worldwide. Since all operation modes of the plant should be considered in the multi-unit PSA, the accident analysis needs to be performed for shutdown operation modes, too. In this study, a station blackout during the mid-loop operation is selected as a reference scenario. The overall accident progression for the mid-loop operation is slower than that for the full-power operation because the residual heat per mass of coolant is about 6 times lower than that in the mid-loop scenario. Though the fractions of Cs released from the core to the RCS in both operation modes are almost the same, the amount of Cs delivered to the containment atmosphere is quite different due to the chemisorption in the RCS. While 45.5% of the initial inventory is chemisorbed on the RCS surfaces during the full-power operation, only 2.2% during the mid-loop operation. The containment remains intact during the mid-loop operation, though 83.9% of Cs is delivered to the containment.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.351-352
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• 2018

• New & Renewable Energy
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• v.7 no.4
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• pp.37-41
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• 2011

This paper presents a Combined solar-wave Power Generation (CPG) concept that the CPG unit is maintained as buoyant at the level of sea water and it is also supported by a submerged tunnel, with the aim of supplying emergency electric power during the station blackout events of nuclear power plants. The CPG concept has been motivated from the 2011 Fukushima-Daiichi Accidents due to the loss of both offsite AC power and emergency diesel power caused by natural hazards such as earthquake and tsunami. The CPG is conceptualized by applying different types and different sites for emergency power generation, in order to reduce common cause failures of emergency power suppliers due to natural hazards. Thus, the CPG can provide a new mean for supplying emergency electric power during station blackout events of nuclear power plants. For this application, the CPG requirements are described with a typical configuration at the ocean side of a submerged tunnel.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.157-164
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• 2015

Nuclear power plants have logical loop structures in their system configuration. This paper explains the method to solve a loop structure in reliability analysis. As examples of loop structured systems, the reactor core isolation cooling system and high-pressure core injection system of a boiling water reactor are considered and analyzed under a station blackout accident condition. The analysis results show the important role of loop structures under severe accidents. For the evaluation of the safety of nuclear power plants, it is necessary to accurately evaluate a loop structure's reliability.

• Journal of the Institute of Electronics and Information Engineers
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• v.51 no.5
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• pp.58-69
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• 2014

To cope with sever accident like Fukushima accident, the emergency response system is needed. It consist of the hardened I&C system and the mobile control station. The hardened I&C system monitors the state in the nuclear power plant and controls the emergency equipment such as valves, pumps and the mobile control station placed at 30km away from nuclear power plant receives the status information from the hardened I&C system and transmits the control data to the hardened I&C system. In this paper, we design the emergency communication system connecting the hardened I&C system to the mobile control station and analyze the performance of the system. This system consists of the terrestrial communication system and the satellite communication. The performance such as a communication link budget, throughput and delay time is analyzed for each system.

• Nuclear Engineering and Technology
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• v.28 no.6
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• pp.507-521
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• 1996

Quantification of uncertainties in the source term estimations by a large computer code, such as MELCOR and MAAP, is an essential process of the current Probabilistic safety assessment. The main objective of the present study is to investigate the applicability of a combined procedure of the response surface method (RSM) based on input determined from a statistical design and the Latin hypercube sampling (LHS) technique for the uncertainty analysis of CsI release fractions under a Hypothetical severe accident sequence of a station blackout at Younggwang nuclear power plant using MAAP3. OB code as a benchmark problem. On the basis of the results obtained in the present work, the RSM is recommended to be used as a principal tool for an overall uncertainty analysis in source term quantifications, while using the LHS in the calculations of standardized regression coefficients (SRC) and standardized rank regression coefficient (SRRC) to determine the subset of the most important input parameters in the final screening step and to check the cumulative distribution functions obtained by RSM. Verification of the response surface model for its sufficient accuracy is a prerequisite for the reliability of the final results that can be obtained by the combined procedure proposed in the present work.