• Journal of Radiation Protection and Research
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• 제21권4호
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• pp.313-327
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• 1996

원전부지내(AR) 사용후핵연료 건식저장방식으로 횡형콘크리트 모듈방식, 금속 저장용기 방식, 콘크리트 저장용기 방식, 수송저장 겸용용기 방식 및 다목적용기 방식 등이 있다. 이중다목적용기 방식을 제외한 다른 방식들은 각각 운영인허가를 받아 이미 세계 각 국에서 사용후핵연료 AR 건식저장에 사용되고 있으며 다목적용기 방식도 최근 개발을 활발히 진행하고 있는 상태이다. AR 건식저장 시설을 운영하고 있거나 추진중인 나라는 미국, 일본, 독일, 캐나다, 스페인, 체코, 스위스 등으로 AR 건식저장을 거쳐 중간저장이나 재처리시설로 수송하는 방식을 채택하고 있다. 우리나라의 경우 월성에서 콘크리트 Silo 건식저장을 이미 사용하고 있으며 일부 다른 원자로도 사용후핵연료 저장능력이 한계에 도달하고 있는 현실을 감안할 때 AR 임시 저장은 불가피한 것으로 여겨진다. 본 보고서에서는 고리를 비롯한 국내원전에 적용 가능한 외국의 AR 저장 시스템 각각에 대하여 설계특성, 설계요건, 기술기준 및 현황 등을 논의하였다. 대부분의 경우 저장용기 인허가 기간은 20년으로 제한하고 있으며 전 수명기간동안 재질의 건전성, 밀봉유지 등이 중요하게 요구되고 있다.

• Nuclear Engineering and Technology
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• 제46권4호
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.

• 전력전자학회논문지
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• 제11권3호
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• pp.201-208
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• 2006

최근 에너지 절약에 대한 관심이 고조되면서 기존 광원보다 효율이 좋은 광원개발이 계속되고 있으며 그 대표적인 것이 고효율 및 고출력(high-power) LED이다. 고출력 LED의 개발로 인해 일부에서는 일반 조명용으로 사용하려는 연구가 진행 중이며 풍력과 태양광 발전을 이용한 야간 조명용 전원으로 사용하려는 연구도 진행되고 있다. 이에 본 논문에서는 조명용 광원으로 직류 전원을 이용한 고출력 LED의 사용 가능성을 확인하고자 하며 기존 방식에 비하여 보다 안정적이고 효율이 좋은 새로운 LED 전원구동장치를 제안한다. 제안된 방식은 기존 방식보다 리액터의 크기를 줄일 수 있었으며 효율이 개선됨을 확인하였다.

• 한국정밀공학회지
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• 제19권11호
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• pp.174-182
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• 2002

The integrity evaluation of pressure tube is essential for the safety of CANDU reactor, and integrity must be assured when flaws or contacts between pressure tube and surrounding calandria tube are found. In order to complete the integrity evaluation, not only complicated and iterative calculation procedures but also a lot of data and knowledge are required. For this reason, an integrity evaluation system, which provides an efficient way of the evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec.？ and FFSG issued by the AECL, and applicable for the evaluation of blister, sharp flaw and blunt notch. Delayed hydride cracking and blister evaluation modules are included in the general flaw and notch evaluation module. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.344-363
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• 1999

A multi-dimensional thermal-hydraulic system analysis code, MARS 1.3.1, has been developed in order to have the realistic analysis capability of two-phase thermal-hydraulic transients for pressurized water reactor (PWR) plants. As the backbones for the MARS code, the RELAP5/MOD3.2.1.2 and COBRA-TF codes were adopted in order to take advantages of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the MARS code, all the functional modules of the two codes were unified into a single code first. Then, the source codes were converted into the standard Fortran 90, and then they were restructured using a modular data structure based on "derived type variables" and a new "dynamic memory allocation" scheme. In addition, the Windows features were implemented to improve user friendliness. This paper presents the developmental work of the MARS version 1.3.1 including the hydrodynamic model unification, the heat structure coupling, the code restructuring and modernization, and their verifications.their verifications.

• 산업공학
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• 제24권2호
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• pp.156-163
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• 2011

The radiopharmaceutical synthesis for PET (positron emission tomography) is composed of chemical reactions using automated synthetical equipment. Due to the radioactive material, the automated equipment is being frequently developed to replace human operators who conduct dangerous, repetitive and dexterous operations. As to operation, the manipulating program is commonly coded using the spread sheet while the whole actuators are mapped in every step. The process map (program) is changed according to such parameters as temperature of reactor, keeping time, mixture sequence and amount of reagent. In cases of customizing the automated synthetical equipment or developing the new radiopharmaceuticals, lots of experiments should be conducted and the programming mistake is not allowed as it can lead abnormal control of the equipment to leak the radioactive materials. The exact process map has depended on trial and error manner. Thus, this study developed the methodology to tabulate the synthetical process to convert the process map automatically while the synthetical module formation is represented by a network model. The proposed method is validated using the actual radiopharmaceutical synthetical procedure.

• Transactions on Control, Automation and Systems Engineering
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• 제1권2호
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• pp.141-146
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• 1999

This paper describes the communication method of an auto-startup system. The Auto-Startup system is designed to operate a nuclear power plant automatically during the startup operation . In general , the operations during startup in existing plant have only been manually controlled by the operator. The manual operation caused to the operator mistake. The Auto-Startup system consists of the Distributed Control System(DCS) and G2 (Expert System). Also, Functional Test Facility(FTF) provides the plant's real-data for an Auto-Startup system. So, it is necessary to develop the communication method between these systems. We developed two methods ; one is a network and the other is a hardwire line. To communicate between these systems (DCS-G2 and DCS-FTF) , we developed the communication program. In case of DCS-FTF, we used the TCP/IP and VXI. BUt, in case of DCS-G2 , we , what it called , developed the bridge program using the GSI(G2 Standard Interface). We test to check the function of the important parameter, in time, for analysis of the developed communication method. The results are a good performance when we check the communication time of important parameter. We conclude that Auto-startup system could save heat-up time about at least 5 hours and reduced the change of the reactor operation and trip.

• 대한기계학회논문집A
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• 제24권9호
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• pp.2167-2173
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• 2000

The HANARO (High-flux Advanced Neutron Application Reactor) has operated since 1995. The Cold Neutron(CN) hole was implanted in the reflector tank from the design stage. Before a vacuum chamber and a moderator cell for the cold neutron source are installed into the CN hole, it is necessary to measure the exact size of the inside diameter and thickness of the CN hole to prevent the interference problem. Due to inaccessibility and high radiation field in the CN hole, a mechanical measurement method is not permitted. The immersion ultrasonic technique is considered as the best method to measure the thickness and the diameter. The 4 axis manipulator of the 2 channel of a sensor module was fabricated. The transducer of 10 MHz results in 0.03 nun of resolution. The inside diameter and thickness for 550 points of the CN hole were measured using 2 channel ultrasonic sensors. The results showed that the thickness is in the range of 13-6.7 mm and inside diameter is in the range of o 156-165. These data will be a good reference in the design of a cold neutron source facility.

• 전기학회논문지
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• 제64권12호
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• pp.1776-1781
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• 2015

Electric power trains receive electric power from overhead cables via a pantograph system. Power collector system in trains increase the cross section of tunnel and require a massive coreless filter reactor in propulsion inverter because of the power disturbance by contact loss phenomenon of a train. In this paper we proposed a wireless train which can run to next station with charging energy of ultra-capacitor module block. We designed DC-DC converter to charge and discharge ultra-capacitor modules by using Next Train running test results and confirm the feasibility of the proposed system through simulation.

• Korean Membrane Journal
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• 제3권1호
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• pp.32-38
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• 2001

Using the hollow fiber membrane module in a lab-scale membrane bioreactor, the anoxic- oxic (AO) process for nitrogen removal was operated for about one year. For the influent wastewater containing 1,200-1,400 mg $1^{-1}$ of CODcr and 200-310 mg $1^{-1}$ of nitrogen, this process achieved a high quality effluent of less than 30 mgCOD $liter^{-1}$ and 50 mgN $liter^{-1}$. The removal rate of organics was above 98% at a loading rate larger than 2.5 kgCOD $m^{-3}$$d^{-1}$. When the internal recycle from the oxic to the anoxic reactor changed room 2n to 600% rout the influent flow rate, the nitrogen removal rate increased from about 70 to 90% at a loading rate of 0.4 kgT-N m-s d-1. The initial increase of transmembrane pressure (TMP) was observed after a 4-month operation while maintaining the flux and MLSS concentration at 7-9 1 $m^2$ $h^{-1}$ and 6,000-14,000 mg $1^{-1}$, respectively. The TMP could be maintained below 15 cmHg for an 8-month operation. The chemical cleaning with an acid followed by an immersion in an alkali solution gave better cleaning result with the membrane operated for 10 month rather than that only by an alkali immersion.