• Nuclear Engineering and Technology
• /
• v.41 no.8
• /
• pp.1079-1086
• /
• 2009

A great volume of radioactive concrete is generated during the operation and the decommissioning of nuclear facilities. The washing-electrokinetic technology in this study, which combined an electrokinetic method and a washing method, was developed to decontaminate the concrete generated in nuclear facilities. The results of only an electrokinetic decontamination for the concrete showed that cobalt was removed to below 1% from the concrete due to its high pH. Therefore, the washing-electrokinetic technology was applied to lower the pH of the concrete. Namely, when the concrete was washed with 3 M of hydrochloric acid for 4 hours (0.17 day), the $CaCO_3$ in the concrete was decomposed into $CO_2$ and the pH of the concrete was reduced to 3.7, and the cobalt and cesium in the concrete were removed by up to 85.0% and 76.3% respectively. Next, when the washed concrete was decontaminated by the electrokinetic method with 0.01M of acetic acid in the 1L electrokinetic equipment for 14.83 days, the cobalt and the cesium in the concrete were both removed by up to 99.7% and 99.6% respectively. The removal efficiencies of the cobalt and cesium by 0.01M of acetic acid were increased more than those by 0.05M of acetic acid due to the increase of the concrete zeta potential. The total effluent volume generated from the washing-electrokinetic decontamination was 11.55L (7.2ml/g).

• Journal of the Korea Institute of Information and Communication Engineering
• /
• v.22 no.10
• /
• pp.1392-1397
• /
• 2018

Nuclear power plants(NPPs) are protected as core facilities managed by major countries. Applying general IT technology to facilities of NPPs, the proportion of utilizing the digitized resources for the rest of the assets except for the existing installed analog type operating resources is increasing. Using the network to control the IT assets of NPPs can provide significant benefits, but the potential vulnerability of existing IT resources can lead to significant cyber security breaches that threaten the entire NPPs. In this paper, we analyze the nuclear cyber security vulnerability regulatory requirements, characteristics of existing vulnerability scanners and their requirements and investigate commercial and free vulnerability scanners. Based on the proposed application method, we can improve the efficiency of checking the network security vulnerability of NPPs when applying vulnerability scanner to NPPs.

• Convergence Security Journal
• /
• v.22 no.1
• /
• pp.11-17
• /
• 2022

Cyber attacks on national infrastructure, including large-scale power outages in Ukraine, have continued in recent years. As a result, ICS-CERT vulnerabilities have doubled compared to last year, and vulnerabilities to industrial control systems are increasing day by day. Most control system operators develop vulnerability countermeasures based on the vulnerability information sources provided by ICS-CERT in the United States. However, it is not applicable to the security of domestic control systems because it does not provide weaknesses in Korean manufacturers' products. Therefore, this study presents a vulnerability analysis framework that integrates CVE, CWE, CAPE, and CPE information related to the vulnerability based on ICS-CERT information (1843 cases). It also identifies assets of nuclear facilities by using CPE information and analyzes vulnerabilities using CVE and ICS-CERT. In the past, only 8% of ICS-CERT's vulnerability information was searched for information on any domestic nuclear facility during vulnerability analysis, but more than 70% of the vulnerability information could be searched using the proposed methodology.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11a
• /
• pp.211-216
• /
• 1996

Since the middle of the eighties of this century a research program both for heating reactor and investigation of heating reactor thermal-hydraulics has been carried out in Institute of Nuclear Energy Technology(INET) of Tsinghua university in China. This kind of heating reactor is a light water cooled and integrated natural circulation reactor with low system pressure and low quality at the exit of core. Because of relatively long riser and low system pressure. a little change of the quality at the exit of the core will result in a relatively large variation of void fraction in the riser. Two full scale test loops. HRTL-5 and HRTL-200 simulating the HR-5 and HR-200 heating reactors in geometry and operation parameters respectively, and some test results from the HRTL-200 test facility are shown in this paper. The range of studied system pressure is from 1.0MPa to 4.0MPa, the largest heat flux is about 50 W/cm2, and the quality at the exit of test section is less than 5%.

• Korean Institute of Interior Design Journal
• /
• v.16 no.5
• /
• pp.81-88
• /
• 2007

Energy consumption has been increased world widely, and the energy retain is one of the most important economic alternatives. These tendencies expand the nuclear power plants not only quantitatively but also qualitatively. Despite of the increasing construction of nuclear power plants and related facilities, every system in main control room(MCR) has been designed and administered solely based on the safety-first principles because of the specificity of nuclear industry. However, recent main control rooms started with the concept that the operators' performance could be optimized though the organic interrelation between human, machine, and environments. Now, it has been recognised in the scope of Ergonomics and Space Design which acknowledge our living spaces as Man-Environment Interface and this change connotes the MCR spaces should be special spaces rather than ordinary spaces. This research investigated domestic and foreign nuclear power plants' MCRs to suggest basic alternatives which can be applied to future MCR. With the review of characteristics of MCR, an integration of interior design, lighting and Ergonomics was explored and classified as types. Futhermore, the classification of environmental characteristics within the relationships between human, machine, and environments was developed through the case analysis of nuclear power plants. The results of this study will provide a basis of space design for system environments that the high level of safety and function are extremely important.

• Nuclear Engineering and Technology
• /
• v.49 no.8
• /
• pp.1589-1599
• /
• 2017

Field programmable gate arrays (FPGAs) are getting more attention in safety-related and safety-critical application development of nuclear power plant instrumentation and control systems. The high logic density and advancements in architectural features make static random access memory (SRAM)-based FPGAs suitable for complex design implementations. Devices deployed in the nuclear environment face radiation particle strike that causes transient and permanent failures. The major reasons for failures are total ionization dose effects, displacement damage dose effects, and single event effects. Different from the case of space applications, soft errors are the major concern in terrestrial applications. In this article, a review of radiation effects on FPGAs is presented, especially soft errors in SRAM-based FPGAs. Single event upset (SEU) shows a high probability of error in the dependable application development in FPGAs. This survey covers the main sources of radiation and its effects on FPGAs, with emphasis on SEUs as well as on the measurement of radiation upset sensitivity and irradiation experimental results at various facilities. This article also presents a comparison between the major SEU mitigation techniques in the configuration memory and user logics of SRAM-based FPGAs.

• Nuclear Engineering and Technology
• /
• v.34 no.3
• /
• pp.259-267
• /
• 2002

NFPA-803 has been referred as the Fire Protection Standard at the Nuclear Power Plants of Pressurized Water Reactor. This Standard has been used as the fire protection regulation, containing prescriptive requirements with deterministic methodology. Recently, with cumulative efforts by the U.S. Nuclear Regulatory Commission and Utilities in America to establish a new Standard, including a quantitative evaluation methodology, NFPA-805, the Performance-Based Standard for FIRE Protection for Light Water Reactor Electric Generating Plants was issued and approved by the American National Standards Institute as an American National Standard with an effective date of February 9, 2001. This paper presents an analysis result from the computer modeling for the fire simulation In addition, it proposes the idea that this kind of analytic method can be available for the facilities design of fire prevention and protection fields, as well as an evaluation for the fire suppression system with a quantitative analysis for the thermal phenomena in fire compartments in Nuclear Power Plants.

• Nuclear Engineering and Technology
• /
• v.47 no.7
• /
• pp.895-906
• /
• 2015

Background: Mitigation of primary water stress corrosion cracking (PWSCC) is a significant issue in the nuclear industry. Advanced nickel-based alloys with lower susceptibility have been adopted, although they do not seem to be entirely immune from PWSCC during normal operation. With regard to structural integrity assessments of the relevant components, an accurate evaluation of crack growth rate (CGR) is important. Methods: For the present study, the extended finite element method was adopted from among diverse meshless methods because of its advantages in arbitrary crack analysis. A user-subroutine based on the strain rate damage model was developed and incorporated into the crack growth evaluation. Results: The proposed method was verified by using the well-known Alloy 600 material with a reference CGR curve. The analyzed CGR curve of the alternative Alloy 690 material was then newly estimated by applying the proven method over a practical range of stress intensity factors. Conclusion: Reliable CGR curves were obtained without complex environmental facilities or a high degree of experimental effort. The proposed method may be used to assess the PWSCC resistance of nuclear components subjected to high residual stresses such as those resulting from dissimilar metal welding parts.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.14 no.2
• /
• pp.32-40
• /
• 2018

Since 2000, the frequency of earthquakes beyond the 5.0 magnitude quake has been increasing in the Korean peninsula. For instance, the 5.0-magnitude earthquake in Baekryong-do in 2003 has occurred, and recent earthquake with Gyeongju(2016) and Pohang(2017) measured respectively magnitude of 5.2 and 5.8 on the Richter scale. As results, the public concern and anxiety about earthquakes are increasing, and therefore it is necessarily required for social infrastructure to reinforce seismic design and energy production facilities directly related to the national economy and security. This study represents the analysis of seismic performance evaluation methodology such as Seismic Margin Assessment (SMA), Seismic Probabilistic Risk Assessment (SPRA), High Confidence Low Probability Failure (HCLPF) in nuclear power plants in order to develop optimal seismic performance improvement. Current methodologies to evaluate nuclear power plants are also addressed. Through review of the nuclear structure evaluation past and current trend, it contributes to be the basis for the improvement of evaluation techniques on the next generation of nuclear power plants.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.19 no.3
• /
• pp.279-287
• /
• 2021

For application in nuclear decommissioning, underwater laser cutting studies were conducted on thick stainless-steel plates for various cutting directions using a 6 kW fiber laser. For cutting along the horizontal direction with horizontal laser irradiation, the maximum cutting speed was 110 mm·min^-1 for a 48 mm thick stainless-steel plate. For cutting along the vertical direction with horizontal laser irradiation, a maximum speed of 120 mm·min^-1 was obtained for the same thickness, which confirmed that the cutting performance was similar but slightly better. Moreover, when cutting with vertically downward laser irradiation, the maximum cutting speed was 120 mm·min^-1 for a plate of the same thickness. Thus, the cutting performance for vertical irradiation was nearly identical to that for horizontal irradiation. In conclusion, it was possible to cut thick stainless-steel plates regardless of the laser irradiation and cutting directions, although the assist gas rose up due to buoyancy. These observations are expected to benefit laser cutting procedures during the actual dismantling of nuclear facilities.