A modeling study of the PMK-NVH integral test facility
Journal Article
·
· Nuclear Technology; (United States)
OSTI ID:5128410
- Jozef Stefan Inst., Ljubljana (Slovenia)
A way of modeling the PMK-NVH integral test facility with RELAP5 thermal-hydraulic code is presented. Two code versions, MOD2/36.05 and MOD3 5m5, are compared and assessed. Modeling is demonstrated for the International Atomic Energy Agency standard problem exercise no. 2, a small-break loss-of-coolant accident, performed on the PMK-NVH integral test facility. Three parametric studies of the break vicinity modeling are outlined, testing different ways of connecting the cold leg and hydroaccumulator to the downcomer and determining proper energy loss discharge coefficients at the break. Further, the nodalization study compared four different RELAP5 models, varying from a detailed one with more than 100 nodes, down to the miniature one, with only [approximately] 30 nodes. Modeling of some VVER-440 features, such as horizontal steam generators and hot-leg loop seal, is discussed.
- OSTI ID:
- 5128410
- Journal Information:
- Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 105:2; ISSN 0029-5450; ISSN NUTYBB
- Country of Publication:
- United States
- Language:
- English
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Conference
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Sat Dec 31 23:00:00 EST 1988
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OSTI ID:6851637
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Journal Article
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OSTI ID:7129108
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
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
COMPUTER CODES
ENRICHED URANIUM REACTORS
LOSS OF COOLANT
MATHEMATICAL MODELS
PARAMETRIC ANALYSIS
POWER REACTORS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEST FACILITIES
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
WWER TYPE REACTORS
WWER-3 REACTOR
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
99 GENERAL AND MISCELLANEOUS
990200 -- Mathematics & Computers
ACCIDENTS
COMPUTER CODES
ENRICHED URANIUM REACTORS
LOSS OF COOLANT
MATHEMATICAL MODELS
PARAMETRIC ANALYSIS
POWER REACTORS
PWR TYPE REACTORS
R CODES
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEST FACILITIES
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
WWER TYPE REACTORS
WWER-3 REACTOR