Experiment data report for semiscale Mod-1 test S-02-3 (blowdown heat transfer test)

Technical Report · Mon Sep 01 00:00:00 EDT 1975

DOI:https://doi.org/10.2172/4170346· OSTI ID:4170346

Crapo, H. S.; Jensen, M. F.; Sackett, K. E.

Recorded test data are presented for Test S-02-3 of the Semiscale Mod-1 blowdown heat transfer test series. Test S-02-3 was conducted from an initial cold leg fluid temperature of 544$sup 0$F and an initial pressure of 2,263 psig. A simulated double-ended offset shear cold leg break was used to investigate the system response to a depressurization transient with a moderately heated core (75 percent design power level). An electrically heated core was used in the pressure vessel to simulate the effects of a nuclear core. System flow was also set at the 75 percent design level to achieve full core temperature differential. The flow resistance of the intact loop was based on core area scaling. During system depressurization, core power was reduced from the initial level of 1.2 MW in such a manner as to simulate the surface heat flux response of the LOFT nuclear fuel rods until such time that departure from nucleate boiling (DNB) occurs. Blowdown to the pressure suppression system was accomplished without simulated emergency core coolant injection or pressure suppression system coolant spray. (auth)

• Research Organization: SEE CODE- 9502158 Aerojet Nuclear Co., Idaho Falls, Idaho (USA). Idaho National Engineering Lab.
• DOE Contract Number: E(10-1)-1375
• NSA Number: NSA-33-006801
• OSTI ID: 4170346
• Report Number(s): ANCR--1233
• Country of Publication: United States
• Language: English

Related Subjects

*BLOWDOWN-- HYDRAULICS
*PRIMARY COOLANT CIRCUITS-- MOCKUP
*PWR TYPE REACTORS-- REACTOR SAFETY
220900* --Nuclear Reactor Technology--Reactor Safety
FLOW RATE
LOSS OF COOLANT
N77200 --Reactors--Power Reactors
Non-breeding
Light-water Moderated
Non-boiling Water-cooled
N77900* --Reactors--Reactor Safety & Environmental Aspects
PRESSURE GRADIENTS
SPECIFICATIONS
TEMPERATURE GRADIENTS