Analysis of BCL transient ECC-bypass test with TRAC-PD2/MOD1 code. [PWR]
Conference
·
OSTI ID:5416668
As part of the independent assessment of the TRAC-PD2/MOD1 code, a calculation for the Battelle Columbus Laboratories (BCL) transient Emergency Core Cooling (ECC) bypass test was performed. The purpose of this calculation was to assess the code's capability to predict the lower plenum refill rate after the initiation of ECC water injection during a postulated large break LOCA in a PWR system. BCL conducted several ECC bypass tests in their 2/15th-scale model of a typical pressurized water reactor vessel.
- Research Organization:
- Brookhaven National Lab., Upton, NY (USA)
- DOE Contract Number:
- AC02-76CH00016
- OSTI ID:
- 5416668
- Report Number(s):
- BNL-NUREG-30828; CONF-820609-21; ON: DE82009190; TRN: 82-011614
- Resource Relation:
- Conference: American Nuclear Society annual meeting, Los Angeles, CA, USA, 6 Jun 1982
- Country of Publication:
- United States
- Language:
- English
Similar Records
Independent assessment of TRAC-PD2/MOD1 code with BCL ECC bypass tests
Comparison of TRAC-PD2 calculations to emergency-core-coolant bypass data
TRAC-PD2 analysis of a large-break loss-of-coolant accident in a typical US PWR
Technical Report
·
Mon Apr 01 00:00:00 EST 1985
·
OSTI ID:5416668
Comparison of TRAC-PD2 calculations to emergency-core-coolant bypass data
Conference
·
Fri Jan 01 00:00:00 EST 1982
·
OSTI ID:5416668
TRAC-PD2 analysis of a large-break loss-of-coolant accident in a typical US PWR
Technical Report
·
Tue Jun 01 00:00:00 EDT 1982
·
OSTI ID:5416668
Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ECCS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
PWR TYPE REACTORS
COMPUTER CALCULATIONS
MOCKUP
REACTOR SAFETY
ACCIDENTS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
STRUCTURAL MODELS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ECCS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
PWR TYPE REACTORS
COMPUTER CALCULATIONS
MOCKUP
REACTOR SAFETY
ACCIDENTS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID MECHANICS
MECHANICS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
STRUCTURAL MODELS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled