High Temperature of Finned-tube Sodium-to-Air Heat Exchanger in the SELFA Test Loop
Creators
- 1. Korea Atomic Energy Research Institute Daedeok-daero 989-111, Yuseong-gu, Daejeon (Korea, Republic of)
Description
A medium-scale sodium test loop named as the 'SELFA'(Sodium Thermal‐hydraulic Experiment Loop for Finned‐tube Sodium‐to‐Air heat exchanger) for simulating thermal hydraulic behavior of the FHX (Finned-tube Sodium-to-Air heat Exchanger) in Korean prototype sodium-cooled fast reactor is planned to be constructed at KAERI (Korea Atomic Energy Research Institute). In this study, elevated temperature design for the FHX and creep-fatigue damage evaluation have been conducted according to the design codes of ASME section III subsection NH and RCC-MRx based on full 3D finite element analyses. Design optimization for the finned-tubes and tube arrangements in the scaled-down FHX mock-up has been performed. The materials of the FHX and piping systems are austenitic stainless steel 316, the design temperature and of the SELFA test loop is 600°C and design pressure is 1MPa. The damage evaluation results have shown that no creep-fatigue damage occurrs in the present design of the FHX under the intended test conditions. (author)
Additional details
Identifiers
Publishing Information
- Publisher
- IAEA
- Imprint Place
- Vienna (International Atomic Energy Agency (IAEA))
- ISBN
- 978-92-0-104114-2
- Imprint Title
- Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios (FR13). COMPANION CD-ROM. Proceedings of an International Conference
- Imprint Pagination
- [1 CD-ROM]
- Series
- Proceedings Series
- Journal Page Range
- 10 p.
- ISSN
- 0074-1884
Conference
- Title
- International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios
- Acronym
- FR13
- Dates
- 4-7 Mar 2013
- Place
- Paris (France)
INIS
- Country of Publication
- International Atomic Energy Agency (IAEA)
- Country of Input or Organization
- International Atomic Energy Agency (IAEA)
- INIS RN
- 46091223
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- AUSTENITIC STEELS; CREEP; DAMAGE; DESIGN; FATIGUE; FINITE ELEMENT METHOD; HEAT EXCHANGERS; KAERI; SODIUM; SODIUM COOLED REACTORS; THERMAL HYDRAULICS; TUBES
- Descriptors DEC
- ALKALI METALS; ALLOYS; CALCULATION METHODS; CARBON ADDITIONS; ELEMENTS; FLUID MECHANICS; HYDRAULICS; IRON ALLOYS; IRON BASE ALLOYS; KOREAN ORGANIZATIONS; LIQUID METAL COOLED REACTORS; MATHEMATICAL SOLUTIONS; MECHANICAL PROPERTIES; MECHANICS; METALS; NATIONAL ORGANIZATIONS; NUMERICAL SOLUTION; REACTORS; STEELS; TRANSITION ELEMENT ALLOYS
Optional Information
- Notes
- 14 refs., 11 figs., 3 tabs.
- Secondary number(s)
- IAEA-CN--199/301