TRAC-PF1/MOD2 analysis of downflow in a ribbed vertical annulus

Conference · Fri Jun 01 04:00:00 EDT 1990 · Transactions of the American Nuclear Society; (USA)

OSTI ID:6124658

Pasamehmetoglu, K O ^[1]

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1. Los Alamos National Lab., TN (USA)

As part of an effort to upgrade the calculation of the emergency cooling system assembly power limits at the Savannah River plant (SRP), a number of heat transfer experiments were designed and conducted to measure the local effective heat transfer coefficients for a two-phase downflow in a ribbed annulus. In the present study, the unpowered portion of these experiments is analyzed using the new version of their transient reactor analysis code, TRAC-PF1/MOD2. Only the results of base study are reported in this paper. The Los Alamos National Laboratory (LANL) TRAC series of computer codes was developed to analyze transients in pressurized light water reactors (PWRs). The design and flow configuration in PWRs are considerably different from those encountered in the fuel assemblies at SRP. What we refer to as a base study is the application of the TRAC code without modifications in the constitutive packages to account for the characteristics of a downflow in a ribbed annulus, which is typical of the SRP fuel assembly flow configuration. As outlined in this paper, TRAC performed very well in modeling the qualitative behavior of all existing data. Quantitatively, the discrepancies ranged from 0 to 100%. The details of these results are documented in this paper. Section II provides a brief description of the experimental setup and the resulting data for the three experiments mentioned above.

• OSTI ID: 6124658
• Report Number(s): CONF-900608--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (USA) Journal Volume: 61
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220700 -- Nuclear Reactor Technology-- Plutonium & Isotope Production Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CODES
ECCS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LANL
LOSS OF COOLANT
MECHANICS
NATIONAL ORGANIZATIONS
PRODUCTION REACTORS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTOR SAFETY EXPERIMENTS
REACTORS
SAFETY
SAVANNAH RIVER PLANT
SPECIAL PRODUCTION REACTORS
T CODES
TWO-PHASE FLOW
US AEC
US DOE
US ERDA
US ORGANIZATIONS