MIST final report

Technical Report · Thu Aug 01 00:00:00 EDT 1991

OSTI ID:5340930

Gloudemans, J R ^[1]

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1. Babcock and Wilcox Co., Alliance, OH (United States)

The multiloop integral system test (MIST) was part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock Wilcox-designed plants. MIST was sponsored by the US Nuclear Regulatory Commission, the Babcock Wilcox Owners Group, the Electric Power Research Institute, and Babcock Wilcox. The unique features of the Babcock Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral system facilities to addresss the thermal-hydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility -- the once-through integral system (OTIS) -- was also used. Data from MIST and OTIS are used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST program is reported in eleven volumes; Volumes 2 through 8 pertain to groups of Phase 3 tests by type, Volume 9 presents inter-group comparisons. Volume 10 provides comparisons between the RELAP5 MOD2 calculations and MIST observations, and Volume 11 (with addendum) presents the later, Phase 4 tests. This is Volume 1 of the MIST final report, a summary of the entire MIST program. Major topics include: test advisory grop (TAG) issues; facility scaling and design; test matrix; observations; comparisons of RELAP5 calculations to MIST observations; and MIST versus the TAG issues. 11 refs., 29 figs., 9 tabs.

• Research Organization: Electric Power Research Inst., Palo Alto, CA (United States); Babcock and Wilcox Co., Alliance, OH (United States)
• Sponsoring Organization: EPRI; Electric Power Research Inst., Palo Alto, CA (United States)
• OSTI ID: 5340930
• Report Number(s): EPRI-NP-6480-Vol.1
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
AFTER-HEAT REMOVAL
LOSS OF COOLANT
PWR TYPE REACTORS
EXPERIMENTAL DATA
HEAT TRANSFER
HIGH PRESSURE COOLANT INJECTION
HYDRAULICS
PRESSURE DROP
PUMPS
R CODES
REACTOR SAFETY
SCALE MODELS
SCALING LAWS
STEAM GENERATORS
T CODES
TEST FACILITIES
TRANSIENTS
TUBES
ACCIDENTS
BOILERS
COMPUTER CODES
DATA
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
INFORMATION
MECHANICS
NUMERICAL DATA
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
REMOVAL
SAFETY
STRUCTURAL MODELS
VAPOR GENERATORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled