1 "PRISM preliminary safety information document, GEFR- 00793"
2 Dohee Hahn, "KALIMER-600 conceptual design report" 2007
3 "Fast reactor database 2006 update, IAEA-TECDOC-1531"
4 Seok-Hoon Kim, "Elevated temperature design evaluations for the ABTR internal structure" 대한기계학회 26 (26): 389-400, 2012
5 G.-H. Koo, "Development of an ASME-NH program for nuclear component design at elevated temperatures" 85 : 385-393, 2008
6 C.-G. Park, "Design and structural evaluation of the ABTR IHTS piping for representative duty events of a level a service" 2008
7 CHANG-GYU PARK, "DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR" 한국원자력학회 41 (41): 1323-1332, 2009
8 구경회, "Creep-fatigue design studies for a sodium-cooled fast reactor with tube sheet-toshell structure subjected to elevated temperature service" 대한기계학회 24 (24): 711-719, 2010
9 G. H. Koo, "Computer program of SIE ASME-NH code (Revision 1)" Korea Atomic Energy Research Institute 2008
10 Y. I. Chang, "Advanced burner test reactor preconceptual design report" ANL 2006
1 "PRISM preliminary safety information document, GEFR- 00793"
2 Dohee Hahn, "KALIMER-600 conceptual design report" 2007
3 "Fast reactor database 2006 update, IAEA-TECDOC-1531"
4 Seok-Hoon Kim, "Elevated temperature design evaluations for the ABTR internal structure" 대한기계학회 26 (26): 389-400, 2012
5 G.-H. Koo, "Development of an ASME-NH program for nuclear component design at elevated temperatures" 85 : 385-393, 2008
6 C.-G. Park, "Design and structural evaluation of the ABTR IHTS piping for representative duty events of a level a service" 2008
7 CHANG-GYU PARK, "DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR" 한국원자력학회 41 (41): 1323-1332, 2009
8 구경회, "Creep-fatigue design studies for a sodium-cooled fast reactor with tube sheet-toshell structure subjected to elevated temperature service" 대한기계학회 24 (24): 711-719, 2010
9 G. H. Koo, "Computer program of SIE ASME-NH code (Revision 1)" Korea Atomic Energy Research Institute 2008
10 Y. I. Chang, "Advanced burner test reactor preconceptual design report" ANL 2006
11 "ASME boiler and pressure vessel code, Section III Rules for Construction of Nuclear Facility Components, Division 1 – Subsection NH, Class 1 Components in Elevated Temperature Service"
12 "ASME boiler and pressure vessel code, Section III Rules for Construction of Nuclear Facility Components, Division 1 – Subsection NB, Class 1 Components"
13 "ANSYS user’s manual for revision 11.0"
14 C.-G. Park, "A comparison study of creep-fatigue defect growth evaluations for a SFR IHTS piping" 2 (2): 20-28, 2008