Scaling criteria and an assessment of Semiscale Mod-3 scaling for small-break loss-of-coolant transients
Conference
·
OSTI ID:6733795
Various methods of scaling fluid thermal-hydraulic test facilities and their relative merits and disadvantages are examined in light of nuclear reactor safety considerations. Particular emphasis is placed on examination of the scaling of the Semiscale Mod-3 system and determination of thermal-hydraulic phenomena thought to be important during a small break loss-of-coolant accident in a pressurized water nuclear reactor. The influence of geometric and dynamic scaling concerns in the Mod-3 system on small break behavior are addressed from an engineering viewpoint and corrective measures contemplated or required to make results from Semiscale tests more meaningful relative to expected PWR response are discussed.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6733795
- Report Number(s):
- EGG-M-21082; CONF-820962-1; ON: DE83000687
- Country of Publication:
- United States
- Language:
- English
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