Inherent safety evaluation of a long-life fast reactor (SALUS-100) under unprotected loss of heat sink conditions

Inherent safety evaluation of a long-life fast reactor (SALUS-100) under unprotected loss of heat sink conditions

The Korea Atomic Energy Research Institute (KAERI) is developing the long-life sodium-cooled fast reactor SALUS-100, designed for continuous 20-year operation without refueling. The concept adapts the proven technology of the Prototype Generation-IV SFR (PGSFR) to a non-light-water small modular rea...

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Bibliographic Details
Main Authors: Junkyu Han, Nam-il Tak, Sun Rock Choi, Hyun-Sik Park, Jonggan Hong, Ji-woong Han, In Sub Jun, Huee-Youl Ye, Jeong Ik Lee
Format: Article
Language:English
Published: Elsevier 2025-12-01
Series:Nuclear Engineering and Technology
Subjects:
Unprotected loss of heat sink
Sodium-cooled fast reactors
Safety analysis
GAMMA+
Design extension condition
Direct reactor auxiliary cooling system
Online Access:http://www.sciencedirect.com/science/article/pii/S1738573325004164
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Summary:The Korea Atomic Energy Research Institute (KAERI) is developing the long-life sodium-cooled fast reactor SALUS-100, designed for continuous 20-year operation without refueling. The concept adapts the proven technology of the Prototype Generation-IV SFR (PGSFR) to a non-light-water small modular reactor platform. To verify that these inherited technologies still provide the intrinsic sodium-cooled fast reactors (SFR) safety features—negative reactivity feedback and passive decay-heat removal via the diverse residual heat-removal system (DRHRS)—an unprotected loss-of-heat-sink (ULOHS) analysis was performed.The transient calculations employed GAMMA+ 2.0, a system code validated against data from JOYO, PFBR, Monju, EBR-II, FFTF and other SFR facilities. A concurrent station blackout was assumed to challenge the passive cooling path, and separate cases examined the loss of one versus two intermediate-loop pumps.During the ULOHS event, a rise in core-inlet temperature triggered dominant negative reactivity through the Doppler effect and radial core expansion, stabilizing the reactor. With a single IHTS pump trip, −0.0207 $ of reactivity was inserted at 213 s, fixing power at 73 % of nominal. When both IHTS pumps tripped, −0.0501 $ occurred at 216 s, stabilizing power at 42.6 %. The magnitude of the initial power excursion scaled with the inlet-temperature spike, highlighting the influence of rapid temperature changes on early transients. Sensitivity studies also quantified the impact of intermediate heat-transport system (IHTS) pump coast-down characteristics and timing of operator actions.
ISSN:1738-5733

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