• Proceedings of the KIEE Conference
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• 2007.07a
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• pp.1630-1631
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• 2007

This paper deals with the design of control cabinet based on safety PLC for reactor power control system(PCS). The PCS controls the operation of the CEDMs(Control Element Drive Mechanisms). The CEDM moves the CEAs(Control Element Assemblies) which regulates the reactor power, vertically in the reactor core. The Control Cabinet in PCS makes and conveys control signals to the power cabinet which provides power to the CEDM. We designed the Control Cabinet, based on POSAFE-Q, safety PLC. The application programs working in PLC can be programmed by pSET, Identified Development Environment.

• Journal of the Korean Society of Safety
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• v.36 no.2
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• pp.101-110
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• 2021

This paper presents the results of an initiating event analysis as part of a Level 1 probabilistic safety assessment (PSA) for at-power internal events for the Korea Research Reactor (KRR). The PSA methodology is widely used to quantitatively assess the safety of research reactors (RRs) in the domestic nuclear industry. Initiating event frequencies are required to conduct a PSA, and they considerably affect the PSA results. Because there is no domestic database for domestic trip events, the safety of RRs is usually assessed using foreign databases. In this paper, operating experience data from the KRR for trip events were collected and analyzed in order to determine the frequency of specific initiating events. These frequencies were calculated using two approaches according to the event characteristics and data availability: (1) based on KRR operating experience or (2) using generic data.

• Journal of the Korea Organic Resources Recycling Association
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• v.8 no.4
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• pp.160-167
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• 2000

This study was performed to evaluate efficiency of soil microbial inoculator for active composting of food waste. In addition the number of microorganisms in roil microbial inoculator and the effect of seeding in the process of composting were investigated. food waste samples collected from a refectory were analyzed for physical-chemical properties. The samples were adjusted to moisture content of 65% by saw dust and seeded with soil microbial inoculator of 10% by the weight in case of reactor B. The number of microorganisms, aerobic bacteria, actinomyces, yeast, and fungi in soil microbial inoculator were over $2.98{\times}10^9/g$, $3.93{\times}10^7/g$, $1.21{\times} 10^5/g$, and $5.79{\times}10^7/g$, respectively. During the process of composting, the highest temperatures were $63.4^{\circ}C$ at reactor A(unseeded control)after 10 days and $66.8^{\circ}C$ at reactor B(seeded compost) after 4 days. The pH values of reactor A and B rapidly increased after 3 days and after first few days during composting period, respectively. The highest $CO_2$ concentrations were 6.1%(after 10 days) and 10.8%(afer 4 days) in reactor A and B, respectively. The degradation rates of organic matter(rd) between reactor A and B increased by 17.1% and 64.5%, respectively Consequently, the effects of Inoculation on comporting parameter such as temperature increasing, pH change, chemical properties, and the degradation rates of organic matter(rd) were higher in seeded compost than in unseeded control.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.1244-1248
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• 2003

This paper describes visual tracking procedure of the underwater mobile robot for nuclear reactor vessel inspection, which is required to find the foreign objects such as loose parts. The yellowish underwater robot body tends to present a big contrast to boron solute cold water of nuclear reactor vessel, tinged with indigo by Cerenkov effect. In this paper, we have found and tracked the positions of underwater mobile robot using the two color information, yellow and indigo. The center coordinates extraction procedures are as follows. The first step is to segment the underwater robot body to cold water with indigo background. From the RGB color components of the entire monitoring image taken with the color CCD camera, we have selected the red color component. In the selected red image, we extracted the positions of the underwater mobile robot using the following process sequences; binarization, labelling, and centroid extraction techniques. In the experiment carried out at the Youngkwang unit 5 nuclear reactor vessel, we have tracked the center positions of the underwater robot submerged near the cold leg and the hot leg way, which is fathomed to 10m deep in depth.

• Journal of the Korean Society of Safety
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• v.35 no.3
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• pp.86-95
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• 2020

This paper describes the work and the results of the final Probabilistic Safety Assessment (PSA) for the Jordan Research and Training Reactor (JRTR). This final PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA, which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, nine typical initiating events were selected regarding internal events during the normal operation of the reactor. AIMS-PSA (Version 1.2c) was used for the accident quantification, and FTREX was used as the quantification engine. 1.0E-15/yr of the cutoff value was used to deliminate the non-effective Minimal Cut Sets (MCSs) when quantifying the JRTR PSA model. As a result, the final result indicates a point estimate of 2.02E-07/yr for the overall Core Damage Frequency (CDF) attributable to internal initiating events in the core damage state for the JRTR. A Loss of Primary Cooling System Flow (LOPCS) is the dominant contributor to the total CDF by a single initiating event (9.96E-08/yr), and provides 49.4% of the CDF. General Transients (GTRNs) are the second largest contributor, and provide 32.9% (6.65E-08/yr) of the CDF.

• International Journal of Safety
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• v.3 no.1
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• pp.53-58
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• 2004

SMART-P is a promising advanced small and medium category nuclear power reactor. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. The enhancement of the safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, and component modularization. Preliminary safety analyses on selected limiting accidents confirm that the inherent safety improving design characteristics and the safety system of SMART-P ensure the reactor's safety. SMART-P is an advanced integral pressurized water reactor. The purpose of this study is for the safety analysis of the steam generator module pipe break for the SMART-P. The integrity of the fuel rod is the major criteria of this analysis. As a result of this analysis, the safety of the RCS and the secondary system is guaranteed against the module pipe break of a steam generator of the SMART-P.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Proceedings of the Korean Vacuum Society Conference
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• 2010.02a
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• pp.314-314
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• 2010

The design of the blanket and shield play a key role in determining the size of a reactor since it has an impact on the various reactor components. The blanket should produce enough tritium for tritium self-sufficiency and the shield should provide sufficient protection for the superconducting TF coil. Neutronic optimization of the blanket and the shield is necessary, and we coupled the system analysis with a neutronic calculation to account for the interrelation of the blanket and shield with the plasma performance of a reactor system in a self-consistent manner. By using the coupled system analysis code, the operational space for a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil is investigated with an spect ratio in the range of 1.5 - 2.5. The minimum major radius which satisfies all the physics and engineering requirements increases with the magnetic field at the magnetic axis. A required inboard shield thickness is mainly determined by the requirement on the protection of the TF coil against radiation damage. It is shown that to have a fusion power bigger than 3,000 MW in the LAR tokamak with a superconducting TF coil, a major radius bigger than 4.0 m is required.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.2
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• pp.171-179
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• 2004

The design concepts of in-service inspection and maintenance are very important for the reactor system design of the nuclear power plant. The strategy of in-service inspection and maintenance should be reflected in the mechanical system design for the verification of the operability of liquid metal reactor KALIMER. In this paper the fundamental approaches of the in-service inspection and maintenance of the KALIMER are established to ensure the safety and reliability of the reactor system. The general method and requirement of the in-service inspection and maintenance for the reactor system and components are proposed and described to satisfy the intents of the ASME Section XI Division 3 and the design characteristics of KALIMER.

• Environmental Engineering Research
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• v.23 no.2
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• pp.195-204
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• 2018

A pilot scale study was conducted on a down-flow hanging sponge (DHS) reactor installed at a sewage treatment plant in Banda Aceh, Indonesia for treatment of desludging septic tank wastewater. Raw wastewater with an average biochemical oxygen demand (BOD) and total suspended solids of 139 mg/L and 191 mg/L, respectively, was pumped into the reactor. Two different hydraulic retention times (HRTs, 3 h and 4 h) were investigated, equivalent to organic loadings of 1.11 and $0.78kg\;BOD/m^3/d$, respectively. The average BOD concentration in the final effluent was 46 and 26 mg/L at HRTs of 3 and 4 h, respectively. The concentration of retained sludge along the reactor height was 10.2-18.7 g VSS/L-sponge, and the sludge activities were 0.24-0.32 and 0.04-0.40 mg/g VSS/h for heterotrophs and nitrification, respectively. Values of water hold-up volume, dispersion coefficient, and number of tank in-series found from tracer studies of clean sponge and biomass-loaded sponge confirmed that growth of retained sludge on the sponge module improved hydraulic performance of the reactor. Adoption of the DHS reactor by this Indonesian sewage treatment plant would enhance the role of the current desludging septic tank wastewater treatment system.