Overview of TRAC-PD2 assessment calculations
A summary of Transient Reactor Analysis Code Version PD2 (TRAC-PD2) calculations performed at the Idaho National Engineering Laboratory (INEL) is presented in this report as part of the US Nuclear Regulatory Commission's (NRCs) overall assessment program of TRAC-PD2. The calculated and measured parameters summarized in this report are break mass flow rate, primary coolant system pressure, reactor core flow rates, and fuel rod cladding temperatures. The data were obtained from seven tests that were performed at two test facilities. The tests were conducted to study the various aspects of cold leg break transients, including the effects of large and small beaks, and core reflood phenomena. User experience gained from the various calculations is also summarized. 42 figs., 10 tabs.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls, ID (United States)
- DOE Contract Number:
- AC07-76ID01570
- OSTI ID:
- 6475367
- Report Number(s):
- NUREG/CR-4195; EGG-2380; ON: TI86003980
- Resource Relation:
- Other Information: Includes 5 sheets of 24x reduction microfiche
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
LOSS OF COOLANT
HEAT TRANSFER
HYDRAULICS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
REACTOR CORES
COMPUTER CODES
COMPUTERIZED SIMULATION
FLOW RATE
FLUID FLOW
FUEL CANS
FUEL RODS
PRESSURE MEASUREMENT
STEADY-STATE CONDITIONS
T CODES
TEMPERATURE MEASUREMENT
TRANSIENTS
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
FUEL ELEMENTS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
SIMULATION
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled