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http://dx.doi.org/10.5516/NET.2009.41.10.1323

DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR  

Park, Chang-Gyu (Korea Atomic Energy Research Institute)
Kim, Jong-Bum (Korea Atomic Energy Research Institute)
Lee, Jae-Han (Korea Atomic Energy Research Institute)
Publication Information
Nuclear Engineering and Technology / v.41, no.10, 2009 , pp. 1323-1332 More about this Journal
Abstract
The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.
Keywords
SFR; IHX Support; Co-axial Structure; Elevated Temperature Structure; ASME Subsection NH; SIE ASME-NH;
Citations & Related Records
Times Cited By KSCI : 1  (Citation Analysis)
Times Cited By Web Of Science : 1  (Related Records In Web of Science)
Times Cited By SCOPUS : 2
연도 인용수 순위
1 RCC-MR Design and Construction Rules for Mechanical Components of FBR Nuclear Islands, Section I-Subsection Z : Technical Appendix A3, AFCEN (2002)
2 ASME Boiler and Pressure Vessel Code, Section III, Division 1 – Subsection NH, 2004 Edition, ASME (2004)
3 SIE ASME-NH, Structural Integrity Evaluation Program by the ASME Subsection NH code, Korea Atomic Energy Research Institute (2008)
4 Dohee Hahn, et al., KALIMER-600 Conceptual Design Report, KAERI/TR-3381/2007, Korea Atomic Energy Research Institute (2007)
5 ASME Boiler and Pressure Vessel Code, Section III, Division 1 – Subsection NB, 2004 Edition, ASME (2004)
6 ASME Boiler and Pressure Vessel Code, Section II, Part D, Properties, 2004 Edition, ASME (2004)
7 Modular Liquid Metal Reactor(LMR) Design Technology, GE, USA (1992)
8 Gyeong-Hoi Koo and Jae-Han Lee, “Development of an ASME-NH program for nuclear component design at elevated temperatures,” International Journal of Pressure Vessels and Piping, 85, pp385-393 (2008)   DOI   ScienceOn
9 Dohee Hahn, Y. Kim, C. Lee, S. Kim, J. Lee, Y. Lee, B. Kim and H. Jeong, “Conceptual design of the Sodium-Cooled Fast Reactor KALIMER-600,” Nuclear Engineering and Technology, 39, 3 (2007)
10 ANSYS Release 10.0, ANSYS Inc (2005)
11 G. H. Koo, Computer Program of SIE ASME-NH Code (Revision 1), KAERI/TR-3526/2008, Korea Atomic Energy Research Institute (2008)