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Title: RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

Abstract

The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation inmore » the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.« less

Authors:
Publication Date:
Research Org.:
Idaho National Engineering and Environmental Laboratory, Idaho Falls, ID (US); Pacific Northwest National Lab., Richland, WA (US)
Sponsoring Org.:
USDOE Office of Nonproliferation and National Security (NN) (US)
OSTI Identifier:
808522
Report Number(s):
INEEL/EXT-03-00058
TRN: US0302195
DOE Contract Number:  
AC07-99ID13727
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: 17 Jan 2003
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 99 GENERAL AND MISCELLANEOUS//MATHEMATICS, COMPUTING, AND INFORMATION SCIENCE; CLEARANCE; COMPUTER CODES; COOLANTS; ENTRAINMENT; HEAT TRANSFER; MIXTURES; NATURAL CONVECTION; PRESSURIZERS; RADIATION PROTECTION; SIMULATION; STEAM GENERATORS; STRATIFICATION; TRANSIENTS; TWO-PHASE FLOW; RELAP/MOD3.2 CODE; LEAK FLOW

Citation Formats

Bayless, P.D. RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility. United States: N. p., 2003. Web. doi:10.2172/808522.
Bayless, P.D. RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility. United States. doi:10.2172/808522.
Bayless, P.D. Fri . "RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility". United States. doi:10.2172/808522. https://www.osti.gov/servlets/purl/808522.
@article{osti_808522,
title = {RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility},
author = {Bayless, P.D.},
abstractNote = {The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.},
doi = {10.2172/808522},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Jan 17 00:00:00 EST 2003},
month = {Fri Jan 17 00:00:00 EST 2003}
}

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