Development of Heat Transfer Model of Helical Tube Steam Generator for Thermal Hydraulic System Analysis Code, MARS (열수력 계통분석코드MARS의 헬리컬 튜브형 증기발생기 열전달 모델 개발)

• 정법동;이영진;조종철
• • Proceedings of the Korean Nuclear Society Conference
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  • 2003.05a
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  • pp.180-180
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  • 2003
• PDF

Design of Lead-Bismuth Spallation Target with Hemi-Spherical Beam Window Based on Thermal-Hydraulic Analyses (열수력 해석에 의한 반구형빔창을 갖는 납-비스무스 핵파쇄 표적 설계)

• 조충호;송태영
• • Proceedings of the Korean Nuclear Society Conference
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  • 2003.05a
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  • pp.127.2-127.2
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  • 2003
• PDF

Critical Heat Flux Experiments on a $5\times5$ Rod Bundle Using a Freon Thermal Hydraulic Experimental Loop (프레온열수력실험장치를 이용한 $5\times5$ 봉다발 임계열유속 실험 연구)

• 홍성덕;문상기;천세영;백원필
• • Proceedings of the Korean Nuclear Society Conference
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  • 2003.05a
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  • pp.137-137
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  • 2003
• PDF

RETRAN-3D Thermal Hydraulic Analysis of PWR Spent Fuel Pool (RETRAN-3D를 이용한 사용후연료 저장조의 열수력 분석)

• 고덕준;김병태;최용준;정성엽;김상녕
• • Proceedings of the Korean Nuclear Society Conference
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  • 1998.10a
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  • pp.304-304
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  • 1998
• PDF

Initial Condition Effect for the Thermal-hydraulic Characteristics of SMART at Main Steam Isolation Valve Closure (증기관 격리밸브 오작동시 초기조건의 변화가 계통에 미치는 영향)

• 정영종;이규형;배규환;김희철
• • Proceedings of the Korean Nuclear Society Conference
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  • 2001.10a
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  • pp.175.1-175
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  • 2001
• PDF

Development of a Three-Dimensional Thermal-Hydraulic Analysis Code for Thermally Stratified Flow in a Curved Piping System

• Jo, Jong-Chull;Park, Seok-Ki
• • Proceedings of the Korean Nuclear Society Conference
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  • 1999.05a
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  • pp.151-151
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  • 1999
• PDF

Evaluation of Core Thermal-hydraulic Model for Vantage 5H Fuel Core Using KEPRI Version of VIPRE-01 Code (VIPRE 코드 KEPRI Version물 사용한 Vantage 5H 연료 노심에 대한 노심열수력 모델평가)

• 이택상;변충섭;이재용
• • Proceedings of the Korean Nuclear Society Conference
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  • 2002.05a
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  • pp.113.1-113
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  • 2002
• PDF

Thermal Hydraulic Analysis of Thorium Fuel Assemblies Loaded with Annular Seed Pins

• Han, Kyu-Hyun;Chang, Soon-Heung
• • Proceedings of the Korean Nuclear Society Conference
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  • 2002.05a
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  • pp.113.2-113
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  • 2002
• PDF

Multi-dimensional flow calculation of Integral Reactor using MULTID Component of Thermal Hydraulic System Analysis Code, MARS (열수력 계통분석코드MARS의 MULTID 컴포넌트를 이용한 일체형원자로의 다차원 유동 계산)

• 이석민;이은철;배성원;정재준;정법동
• • Proceedings of the Korean Nuclear Society Conference
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  • 2004.05a
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  • pp.122-122
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  • 2004
• PDF

Development of Scaling Laws on Thermal-Hydraulic Effect Test Facility for CANDU-6 Moderator (CANDU-6 원자로 감속재 열수력 개별영향실험설비 제작을 위한 축소화 기법 개발)

• 이재영;정윤식;김남석;김만웅;김효정
• • Proceedings of the Korean Nuclear Society Conference
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  • 2004.05a
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  • pp.165-165
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  • 2004
• PDF

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