Verification test calculations for the Source Term Code Package
The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs.
- Research Organization:
- Battelle Columbus Labs., OH (United States)
- OSTI ID:
- 5538258
- Report Number(s):
- NUREG/CR-4656; BMI-2140; ON: TI86901579
- Country of Publication:
- United States
- Language:
- English
Similar Records
Parametric radionuclide release calculations using the MAAP-3. 0 computer code: Final report
Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CONTAINMENT SYSTEMS
M CODES
N CODES
T CODES
V CODES
PEACH BOTTOM-1 REACTOR
POWER LOSSES
PEACH BOTTOM-2 REACTOR
PEACH BOTTOM-3 REACTOR
CALCULATION METHODS
FAILURES
HEAT TRANSFER
HYDRAULICS
MATHEMATICAL MODELS
MELTDOWN
ACCIDENTS
BWR TYPE REACTORS
COMPUTER CODES
CONTAINMENT
ENERGY LOSSES
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
ENRICHED URANIUM REACTORS
FLUID MECHANICS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
HELIUM COOLED REACTORS
HTGR TYPE REACTORS
LOSSES
MECHANICS
POWER REACTORS
REACTOR ACCIDENTS
REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled