Semiscale program summary: a review of Mod-3 results
Journal Article
·
· Nucl. Saf.; (United States)
OSTI ID:5747588
The objectives of the Semiscale Program are briefly defined, and accomplishments during the Mod-3 portion of the program are summarized. Significant results from several series of experiments are presented: (1) results from a series of baseline experiments that included large-break blowdown, reflood, and integral blowdown-reflood tests; (2) results from simulations of the transient at the Three Mile Island 2 nuclear power station; (3) experimental findings from a small-break blowdown test series; and (4) miscellaneous experimental results including plant blackout simulations (complete loss of electrical power), upper-head injection drain tests, and natural circulation investigations.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls
- OSTI ID:
- 5747588
- Journal Information:
- Nucl. Saf.; (United States), Journal Name: Nucl. Saf.; (United States) Vol. 22:3; ISSN NUSAA
- Country of Publication:
- United States
- Language:
- English
Similar Records
Semiscale program summary: a review of Mod-1 results
Test prediction of the Semiscale Mod-3 series Semiscale Mod-3 baseline test series, Test S-07-5
Semiscale Mod-2A Intermediate Break Test series: test-results comparison. [PWR]
Journal Article
·
Thu May 01 00:00:00 EDT 1980
· Nucl. Saf.; (United States)
·
OSTI ID:7013940
Test prediction of the Semiscale Mod-3 series Semiscale Mod-3 baseline test series, Test S-07-5
Technical Report
·
Tue Aug 01 00:00:00 EDT 1978
·
OSTI ID:6541292
Semiscale Mod-2A Intermediate Break Test series: test-results comparison. [PWR]
Technical Report
·
Fri Dec 31 23:00:00 EST 1982
·
OSTI ID:6533963
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BLOWDOWN
CONVECTION
CORE FLOODING SYSTEMS
ECCS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
LOSS OF COOLANT
NATURAL CONVECTION
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY EXPERIMENTS
REACTORS
SIMULATION
THERMAL REACTORS
THREE MILE ISLAND-2 REACTOR
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BLOWDOWN
CONVECTION
CORE FLOODING SYSTEMS
ECCS
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
LOSS OF COOLANT
NATURAL CONVECTION
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY EXPERIMENTS
REACTORS
SIMULATION
THERMAL REACTORS
THREE MILE ISLAND-2 REACTOR
WATER COOLED REACTORS
WATER MODERATED REACTORS