Browse "Dept. of Nuclear and Quantum Engineering(원자력및양자공학과)" by Title

Showing results 6256 to 6275 of 8456

6256
S-CO2 Compressor Performance Test Plan Considering Measurement Accuracy

조성국; 정용주; 이정익researcher, 2019 한국원자력학회 추계학술발표회, 한국원자력학회, 2019-05-23

6257
S-CO2 cycle design and control strategy for the SFR application

Ahn, Yoonhan; Kim, Minseok; Lee, Jeong Ikresearcher, The 5th International Symposium-Supercritical CO2 Power Cycles, Southwest Research Institute, 2016-03

6258
S-CO2 TURBINE DESIGN FOR DECAY HEAT REMOVAL SYSTEM OF SODIUM COOLED FAST REACTOR

Cho, Seong Kuk; Lee, Jekyoung; Lee, Jeong Ikresearcher; Cha, Jae Eun, ASME Turbo Expo 2016, ASME, 2016-06-13

6259
SA106Gr.C 주증기 배관재의 피로균열성장에 미치는 후열처리의 영향

김인섭, 국내원전기기의 건전성평가기술 워크샵, 1996

6260
SA508 Gr.3 Cl.2 저합금강과 용접부의 290℃ 수화학 환경에서 피로균열거동 분석

조평연; 김정현; 장창희researcher; 조현철, CORROSION SCIENCE AND TECHNOLOGY, v.11, no.4, pp.120 - 128, 2012-08

6261
SA508 Gr.3 Cl2 저합금강 모재와 용접부의 290oC 순수 환경에서 피로균열거동 분석

장창희researcher; 김정현; 홍종대; 나경환; 이재곤, 2012 한국부식방식학회 춘계학술대회, 사단법인 한국부식방식학회, 2012-05-18

6262
SA508 Gr.3 강의 천이영역 파괴특성에 미치는 응력상태의 영향 = Effect of stress state on the fracture behavior of SA508 Gr.3 steel in the transition regionlink

김석훈; Kim, Seok-Hun; 황용석; 김인섭; et al, 한국과학기술원, 1998

6263
SACS2: A dynamic and formal approach to safety analysis for complex safety critical system

Koh, Kwang Yong; Seong, Poong-Hyunresearcher, 6th American Nuclear Society International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies 2009, v.1, pp.298 - 311, 2009-04-05

6264
SACS2: Dynamic and Formal Safety Analysis Method for Complex Safety Critical System

고광용; 성풍현researcher, 한국원자력학회 2009 춘계학술발표회, 한국원자력학회, 2009-05-21

6265
Safety activities on safety-critical software for reactor protection system

Park, G.-Y.; Kwon, K.C.; Jee, E.; Koh, K.Y.; Seong, Poong-Hyunresearcher, TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, v.96, no.0, pp.237 - 238, 2007

6266
Safety analysis and development of control logic of KAIST Micro Modular Reactor with GAMMA+ code = GAMMA+ 코드를 이용한 초소형 모듈원전의 안전해석 및 제어논리 개발link

Oh, Bong Seong; Lee, Jeong Ik; et al, 한국과학기술원, 2017

6267
Safety Analysis of Compacted Spent Fuel with Natural Convection Flow

Lee, Kun Jairesearcher, ENS/ANS Transactions (ENC 86), pp.741 - 746, 1986

6268
Safety analysis of compacted spent fuel with natural convection flow = 자연 대류시 밀집된 사용후 핵연료 저장조에서의 안전성 분석link

Lee, Chang-Ju; 이창주; et al, 한국과학기술원, 1986

6269
Safety analysis of LOCA, transients and tritium permeation in hydrogen production HTGRlink

, 한국과학기술원, 2012

6270
Safety analysis of safety-critical software for nuclear digital protection system

Park G.-Y.; Lee J.-S.; Cheon S.-W.; Kwon K.-C.; Jee, Eunk Young; Koh K.Y., 26th International Conference on Computer Safety, Reliability, and Security, SAFECOMP 2007, v.4680, pp.148 - 161, 2007-09-18

6271
Safety Analysis of VHTRs with MED Desalination Plants

No Hee Cheonresearcher; Kim, Ho Sik; Jin, Hyung Gon, American Nuclear Society, 2010-06

6272
Safety Analysis of VHTRs with MED Desalination Plants

No, Hee Cheonresearcher; Kim, Ho Sik; Jin, Hyung Gon, American Nuclear Society, v.104, 2011-06

6273
Safety Analysis Using Coloured Petri Nets

Cho, Seung Mo; Hong, Hyoung Seok; Cha, Sungdeok, Asia-Pacific Software Engineering Conference, pp.176 - 183, IEEE, 1996-12-04

6274
Safety Assessment and Public Acceptance of Nuclear in Korea

Kang, Hyun Gookresearcher, IYNC 2002, 2002

6275
Safety Assessment on Vitrification Facility of Low- and Intermediate-Level Radioactive Wastes

Heo, GY; Lee, SJ; Ji, PG; Maeng, SJ; Park, JG; Shin, SU; Chang, Soon-Heungresearcher, 한국원자력학회 2002년도 춘계공동학술발표회, 한국원자력학회, 2002

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