• Nuclear industry
• /
• v.4 no.4 s.20
• /
• pp.16-20
• /
• 1984

다음 글은 지난 5월 24일 한국전력공사 회의실에서 있은 한국원자력산업회의 주최 $\ulcorner$원자력발전의 경제성에 관한 워크숍$\lrcorner$에서 고교 굉 일본에너지경제연구소 선임고문이 행한 특별강연의 요지이다.

• Nuclear industry
• /
• v.35 no.4
• /
• pp.63-64
• /
• 2015

국제원자력안전연구소(WINS : World Institute for Nuclear Security)가 내놓은 최신 보고서는 원전 분야에서도 빅 데이터 기반의 통합적 데이터 관리와 분석의 도입을 진지하게 고려할 것을 권고하고 있다.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2016.10a
• /
• pp.29-30
• /
• 2016

• Nuclear Engineering and Technology
• /
• v.25 no.2
• /
• pp.248-254
• /
• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Nuclear Engineering and Technology
• /
• v.27 no.5
• /
• pp.701-709
• /
• 1995

The realistic discharge coefficient for the critical How model of RELAP5/AOD3/KAERI are determined for the subcooled and too-phase critical flow by assessments of nine MARVIKEN Critical flew Test(CFT). The selected test runs include a high initial subcooling and large nozzle aspect rat-io(L/D). The code assessment results show that RELAP5/MOD3/KAERI over-predicts the subcooled critical flow and under-predicts the two-phase critical flow. Using these result, the realistic discharge coefficients of critical flow models are quantified by an iterative method. The realistic discharge coefficients are determined to be 0.89 for the subcooled critical How and 1.07 for the two-phase critical flow, and the associated standard deviations are 0.0349 and 0.1189, respectively. The results obtained from this study can be applied to calculate the realistic system response of Large Break Loss of Coolant Accident and to evaluate the realistic Emergency Core Cooling System performance.

• Nuclear Engineering and Technology
• /
• v.27 no.2
• /
• pp.216-225
• /
• 1995

Simulation for a dynamic analysis of the electrolytic preparation of U(IV) in two-phases system, which consisted of mass transfer of U(VI) from TBP phase into HNO$_3$ solution and electrolytic re-duction of U(VI) to U(IV) at a cathode in aqueous phase, was carried out in order to establish the most suitable operating condition and best electrode area as basic design data for the system. It was found that maintaining an appropriate mass transfer rate was more significant rather than enlarging the surface area of the cathode for more effective production yield of U(IV). The electrode area and the operation time affected deeply the production composition of U(IV) in the resulting aqueous phase. And optimal electrode areas ore evaluated to meet production criteria of U(IV) of resulting solution in several system conditions. Though about 0.37M HNO$_3$ was preferable to prepare the solution of U(IV), nitric acid concentration should be higher than 0.5M to prevent a hydrolysis of U(IV) in the aqueous phase.

• Analytical Science and Technology
• /
• v.19 no.5
• /
• pp.389-393
• /
• 2006

The detector consisted of ZnS(Ag) scintillator and photomultiplier tube (PMT) is widely used as contamination monitor in the nuclear facilities. Such detectors are mainly manufactured by adhering the ZnS(Ag) powder onto the transparent plastic. In this study the preparation condition for ZnS(Ag) scintillator sheet using a simple method was established. The scintillator sheet was composed with a support polymer sheet and ZnS(Ag) scintillator layer. The base sheet was prepared by casting the polymer solution after solving the polymer with solvent and the scintillator layer was manufactured by printing the mixture solution with ZnS(Ag) and paste. It was found that the polysulfone(PSf) as a polymer for the base sheet and a cyano resin as a paste for adhering the ZnS(Ag) scintillator was suitable. Also, the prepared thin scintillator sheet had a sufficient mechanical strength, a optical transparency and an alpha-ray detection performance.

• Proceedings of the Korean Society for Rock Mechanics Conference
• /
• 2006.09a
• /
• pp.122-134
• /
• 2006

현재 한국원자력연구소에서는 고준위폐기물 처분시스템의 다양한 현장 실증연구를 위해 원자력연구소내 지하처분연구시설(KAERI Underground Research Tunnel, KURT)가 건설되었다. 터널 크기 $6m{\times}6m$, 총길이 255m(진입터널 180m, 연구모듈 75m) 인 KURT는 결정질 화강암반에 위치하고 있다. KURT 에서는 개념설계, 부지조사, 시설설계, 건설 과정에서 다양한 암석 역학 관련 연구들이 수행되었다. 물리탐사, 시추공조사, 암석물성시험, 현장 물성 시험 등을 통해 KURT 의 구조적 안정성 평가에 필요한 암석 및 암반의 물성이 얻어졌으며 이들 물성은 해석 모델의 입력자료로 활용되었다. 본 연구에서는 KURT 에서 수행되었던 암석역학 관련 시험과 주요들을 소개하고 시험을 통해 얻어진 주요 결과와 이를 활용한 3차원 구조해석에 대해 논의하고자 한다.

• Nuclear Engineering and Technology
• /
• v.27 no.5
• /
• pp.784-791
• /
• 1995

This paper describes the counting-loss problem for radiation measurement Multi-channel analyzers and spectrometers adopt various techniques for compensation for counting-losses in process-ing the radiation pulses from a detector. Researchers hate tried to seek the best solution for the problem. However, any absolute solution has not been reached and vendors of radiation instruments use their own algorithms individually. This survey explains the various compensation algorithms with electronic implementation approach. Shortcomings and merits of each algorithm are also reviewed and a direction is suggested of the recommendable development strategy for counting-loss compensation.

• Nuclear industry
• /
• v.7 no.7 s.53
• /
• pp.39-55
• /
• 1987

다음 내용은 지난 3월 16일에서 4월 24일까지 6주간 미국 알곤원자력 연구소와 카나다 원자력공사에서 공동 주최한 ${\ulcorner}$IAEA원전 안전운전과정${\lrcorner}$에서 특별연사로 초청된 Wisconsin Public Serice사의 S. Burstein부회장이 강의한 것이다. 이 과정에 참가했던 과학기술처 허남 과장이 제공한 자료를 원자력안전센터 은영수 박사가 완역했다.