Experiment data report for semiscale Mod-3 integral blowdown and reflood heat transfer test S-07-6 (Baseline Test Series)
Description
Recorded test data are presented for Test S-07-6 of the Semiscale Mod-3 baseline test series. This test is one of several Semiscale Mod-3 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-07-6 was conducted from initial conditions of 15.21 MPa and 559 K to investigate the response of the Semiscale Mod-3 system to a blowdown transient following a simulated double-ended offset shear of the broken loop cold leg piping. The specific objective of this test was to provide reference data to evaluate integral blowdown and reflood behavior during a 200% cold leg break with emergency core coolant (ECC) injection into the intact loop cold leg of the Mod-3 system. The purpose of this report is to make available the uninterpreted data from Test S-07-6 for future data analysis. The data, presented in the form of graphs in engineering units, have been analyzed only to the extent necessary to ensure that they are reasonable and consistent
Availability note (English)
MF available from INIS under the Report Number; Available from NTIS., PC A13/MF A01.
Files
Additional details
Additional titles
- Augmented title (English)
- PWR
Publishing Information
- Imprint Pagination
- 295 p.
- Report number
- NUREG/CR--0467
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 10461203
- Subject category
- S61: RADIATION PROTECTION AND DOSIMETRY; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
- Descriptors DEI
- BLOWDOWN; ECCS; FLUID FLOW; HEAT TRANSFER; HYDRODYNAMICS; LOSS OF COOLANT; MOCKUP; PERFORMANCE TESTING; PWR TYPE REACTORS; SIMULATION
- Descriptors DEC
- ACCIDENTS; ENERGY TRANSFER; FLUID MECHANICS; MECHANICS; REACTOR ACCIDENTS; REACTOR PROTECTION SYSTEMS; REACTORS; STRUCTURAL MODELS; TESTING; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- TREE--1226.