• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of Energy Engineering
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• v.25 no.4
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• pp.124-132
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• 2016

In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

• Journal of the Korean Society of Safety
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• v.15 no.2
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• pp.70-77
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• 2000

For hazardous air pollutants(HAP) such as NO, $NO_2$ and $SO_2$ decomposition efficiency, power consumption, and applied voltage were investigated by SPCP(Surface induced discharge Plasma Chemical Processing) reactor to obtain optimum process variables and maximum decomposition efficiencies. Decomposition efficiency of HAP with various electric frequencies(5~50 kHz), flow rates(100~1,000 mL/min), initial concentrations(100~1,000 ppm) and additive($CH_4$) were measured and the products were analyzed with FT-IR. Experimental results showed that for the frequency of 10 kHz, the highest decomposition efficiency of 94.3 % for NO, 84.7 % for $NO_2$ and 99 % far $SO_2$ were observed at the power consumptions of 19.8, 20 and 19W, respectively, and that decomposition efficiency decreased with increasing frequency above 20 kHz. And decomposition efficiency per unit power were 5.21 %/W for $SO_2$, 4.76 %/W for NO and 4.24 %/W for $NO_2$ and the highest decomposition efficiency was observed with $SO_2$. Decomposition efficiency was increased with increasing residence times and with decreasing initial concentration of pollutants. When the additive of $CH_4$ was used, decomposition efficiency was increased with increasing $CH_4$ content, and NO, $NO_2$ and $SO_2$ were almost completely decomposed with the efficiency of 99 %, 98 % and 99 %, respectively and therefore $CH_4$ was a good additive material. The optimum power for the maximum decomposition efficiency were 7.5 W for $SO_2$, 9.5 W for NO and 15.5 W for $NO_2$, respectively. Optimum power with the maximum decomposition efficiency were 9.5 W at 1,000 ppm of NO, 7~8 W at 100~500 ppm of NO and 15.5 W at all concentration range of $NO_2$ and 11.5 W at 1,000 ppm, 4.9 W at 500 ppm, 3.7 W at 100~300 ppm of $SO_2$ and power efficiency was best in these case.

• Progress in Medical Physics
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• v.24 no.3
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• pp.171-175
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• 2013

Applicator of various kind of number ten kinds is used to raise from efficiency of brachytherapy to maximum. The compatibility of radiation source and applicator is very important subject for safety brachytherapy. Developed high dose rate brachytherapy source through Hanaro nuclear reactor in Korea Atomic Energy Research Institute and improve compatibility with using equipment in present. In this research, we wished to evaluate stability mechanical safety of radiation source and we developed phantom for evaluate several quality about Ir-192 sealed source that improve newly in Korea Atomic Energy Research Institute and is improved. The result for suitability of Ir-192 HDR source with brachytherapy tools that did normal operation in 2.2~2.7 cm extent about change of equal curvature and consider change of sudden curvature that did normal operation in radius 1.5~1.8 cm extent.

• Fire Science and Engineering
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• v.34 no.3
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• pp.134-140
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• 2020

Nuclear power plants (NPPs) in Korea are required to be maintained using a defense in-depth approach to prevent leakage of radioactive substances outside the plant and allow safe shutdown in the event of a fire. Periodic testing must be conducted to ensure that the fire protection facilities perform as required by the laws for various nuclear reactor types. In June 2017, for the first time in Korea, a nuclear plant, Kori Unit 1, was permanently shut down. It was prepared for decommissioning in accordance with the fire protection regulations imposed by the regulatory body. However, a standard protocol is necessary for systematically establishing the fire protection program for decommissioning of NPPs in the future. Therefore, the nuclear legal systems of countries with many operating nuclear power plants, such as the United States, Japan, Canada, and various European countries, were reviewed and guidelines for establishing a fire protection program for decommissioning NPPs was suggested; the fire protection requirements stated by Reg Guide 1.191 (Decommissioning fire protection program for NPPs during decommissioning and permanent shutdown) were used as a model. Suggestions for establishing legal regulations to optimize fire protection programs and secure basic technology for decommissioning NPPs were also made.

• Journal of Animal Environmental Science
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• v.4 no.1
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• pp.21-28
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• 1998

This study was carried out to estimate the economic impacts on operation cost and curtail the bulking agent between two kinds of plants in swine farms. Bulking agents and Plants have a variety of roles in the fields of the composting for livestock manure and also represent an economic problem in terms of plant operation costs and compost production. Two farms which have rotary(size of reactor : 10${\times}$35${\times}$1.5m) or bucket(size of reactor : 10${\times}$68${\times}$2m) plants were used for 24 weeks for bucket conveyor system, which are composed of refilling rice hull as a bulking agent every 3 weeks till decreasing volume, for 4 weeks for rotary conveyor system, which has continuous compost production system without refilling rice hull, respectively. Composts was produced in 24 weeks in the bucket conveyor system, in 4 weeks in the rotary conveyor system, respectively. The results are as follows : 1. The tissue change of Rice hull at the composts of 45 days pursuant to composting steps was more crumbling in bucket conveyor system than in rotary conveyor system. 2. Microbial counts of the final composts for safety and quality showed that total bacteria counts was 1.01${\times}$108 cfu/g in bucket conveyor system, 2.82${\times}$108 cfu/g in rotary conveyor system, Salmonella was 0.3${\times}$102 cfu/g in bucket conveyor system, 7.6${\times}$102 cfu/g in rotary conveyor system, colifom bacteria was 0.5${\times}$106 cfu/g in bucket conveyor system, 1.5${\times}$106cfu/g in rotary conveyor system, fungi 1.24${\times}$106 cfu/g in bucket conveyor system, 0.01${\times}$106 cfu/g in rotary conveyor system, respectively. However, Any system used in this trial could not be met the regulation of A grade compost of EPA and USA. 3. C:N ratio according to the composting was more rapidly changed in bucket conveyor system with 64.5 of 5 days compost to 25.4 of final products than in rotary conveyor system with 26.7 of 5 days compost to 25.9 of final products. 4. Based on the mechanical characteristics of plants used in trial and compared with Rotary conveyor system, the Bucket conveyor system in which has 0.72 ㎥/㎥ of bulking agent capacity per slurry could be curtailed 1.78 ㎥of rice hull for disposal of waste, 1㎥. It was proper facilities to produce composts quantitative in Rotary conveyor system, and to treat waste quantitative and obtain good results in compost quality in Bucket conveyor system.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• Journal of the Korean Society for Nondestructive Testing
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• v.24 no.2
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• pp.164-170
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• 2004

Liquid Injection Nozzle(LIN) tube and Calandria tube(CT) in pressurized Heavy Water Reactor (PHWR) are .ross-aligned horizontally. These neighboring tubes can contact each other due to the sag of the calandria tube resulting from the irradiation creep and thermal creep, and fuel load, etc. In order to judge the contact which might be the safety concern, the remote field eddy current (RFEC) technology is applied for the gap measurement in this paper. LIN can be detected by inserting the RFEC probe into pressure tube (PT) at the crossing point directly. To obtain the optimal conditions of the RFEC inspection, the sensitivity, penetration and noise signals are considered simultaneously. The optimal frequency and coil spacing are 1kHz and 200mm respectively. Possible noises during LIN signal acquisition are caused by lift-off, PT thickness variation, and gap variation between PT and CT. The simulated noise signals were investigated by the Volume Integral Method(VIM). Signal analysis on the voltage plane describes the amplitude and shape of LIN and possible defects at several frequencies. All the RFEC measurements in the laboratory were done in variance with the CT/LIN gap and showed the relationship between the LIN gap and the signal parameters by analyzing the voltage plane signals.

• Proceedings of the KSRS Conference
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• 2006.03a
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• pp.178-186
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• 2006

기상 수치예보는 (Numerical Weather Pridiction, NWP)는 바람, 기온, 등과 같은 기상요소의 시간 변화를 나타내는 물리방정식을 컴퓨터로 풀어 미래의 대기 상태를 예상하는 과학적인 방법으로 지구를 상세한 격자 2진부호(GRIdded Binary, 이하 GRIB)로 나누어 그 격자점에서의 값으로 대기 상태를 나타낸다. 지구 각지에서의 각종 관측자료를 기초로 격자점상의 현재값을 구한다. 대용량의 격자데이터는 이진형태이어서 컴퓨터, 서버 저장장치에서 동일형태 데이터로 존재한다. 우리나라 최초의 저궤도 관측 위성인 다목적 실용위성 KOMPSAT-1호(이하, 아리랑 위성1호)는 전자광학카메라(Electro Optical Camera, EOC)를 탑재하여 1999년 12월 21일에 발사된 이후 2006년 1월 현재까지 6여년간 성공적으로 임무를 수행, 7049여회의 영상을 획득하여 국가적으로 귀중한 자료로 활용하고 있다. 아리랑 위성1호는 일일 2-3회 EOC영상을 획득하고 있으며, 임무계획(Mission Planning)은 MP(Mission Planner)가 사용자로부터 자료를 수집하여 임무분석 및 계획 서브시스템(MAPS)에 의해 계산되어진 위성의 제도예측 데이터에 촬영하고자하는 목표지점 좌표를 입력하여 자동명령생성기(KSCG)에 의해 계산된 촬영 경사각도(Tilt)값을 위성에 전송하여 목표지역의 영상을 획득하게 된다. 위성영상 획득에 있어 고가의 위성을 운영하면서 기상의 상태를 정확히 예측하여 실패없이 유효한 영상을 획득하는 것이 무엇보다 중요하다. 본 논문에서는 효율적인 위성임무계획을 위한 기상수치예보 자료를 분석하여 앞으로 발사하게 될 고해상 카메라 탑제위성인 아리랑 위성2호와 3호에 적용하고자 한다. the sufficient excess reactivity to override this poisoning must be inserted, or its concentration is decreased sufficiently when its temporary shutdown is required. As ratter of fact, these have an important influence not only on reactor safety but also on economic aspect in operation. Considering these points in this study, the shutdown process was cptimized using the Pontryagin's maximum principle so that the shutdown mirth[d was improved as to restart the reactor to its fulpower at any time, but the xenon concentration did not excess the constrained allowable value during and after shutdown, at the same time all the control actions were completed within minimum time from beginning of the shutdown.및 12.36%, $101{\sim}200$일의 경우 12.78% 및 12.44%, 201일 이상의 경우 13.17% 및 11.30%로 201일 이상의 유기의 경우에만 대조구와 삭제 구간에 유의적인(p<0.05) 차이를 나타내었다.는 담수(淡水)에서 10%o의 해수(海水)

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.465-471
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• 2013

A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.