Published 1984 | Version v1
Report Restricted

Simulation of LOFT anticipated-transient experiments L6-1, L6-2, and L6-3 using TRAC-PF1/MOD1

Creators

Description

Anticipated-transient experiments L6-1, L6-2, and L6-3, performed at the Loss-of-fluid Test (LOFT) facility, are analyzed using the latest released version of the Transient Reactor Analysis Code (TRAC-PF1/MOD1). The results are used to assess TRAC-PF1/MOD1 trip and control capabilities, and predictions of thermal-hydraulic phenomena during slow transients. Test L6-1 simulated a loss-of-stream load in a large pressurized-water reactor (PWR), and was initiated by closing the main steam-flow control valve (MSFCV) at its maximum rate, which reduced the heat removal from the secondary-coolant system and increased the primary-coolant system pressure that initiated a reactor scram. Test L6-2 simulated a loss-of-primary coolant flow in a large PWR, and was initiated by tripping the power to the primary-coolant pumps (PCPs) allowing the pumps to coast down. The reduced primary-coolant flow caused a reactor scram. Test L6-3 simulated an excessive-load increase incident in a large PWR, and was initiated by opening the MSFCV at its maximum rate, which increased the heat removal from the secondary-coolant system and decreased the primary-coolant system pressure that initiated a reactor scram. The TRAC calculations accurately predict most test events. The test data and the calculated results for most parameters of interest also agree well

Availability note (English)

MF available from INIS under the Report Number; Available from NTIS, PC A03/MF A01 as DE84004416.

Files

Restricted

The record is publicly accessible, but files are restricted to users with access.

Additional details

Publishing Information

Imprint Pagination
27 p.
Report number
LA-UR--83-3649

Conference

Title
Design, construction and operation of nuclear power plants conference.
Dates
5-8 Aug 1984.
Place
Portland, OR (USA).

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
16035078
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference, Numerical Data
Descriptors DEI
COMPUTER CALCULATIONS; EXPERIMENTAL DATA; FLOW RATE; HEAT TRANSFER; HYDRAULICS; LOFT REACTOR; PRESSURE GRADIENTS; PWR TYPE REACTORS; REACTOR ACCIDENTS; REACTOR SAFETY; THEORETICAL DATA
Descriptors DEC
ACCIDENTS; DATA; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; INFORMATION; NUMERICAL DATA; POWER REACTORS; REACTORS; RESEARCH AND TEST REACTORS; SAFETY; TANK TYPE REACTORS; TEST REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS

Optional Information

Secondary number(s)
CONF-840813--2.