Analyzing the rod drop accident in a BWR with high burnup fuel. Revised

Conference · Sat Feb 01 04:00:00 EST 1997

OSTI ID:465206

Diamond, D J; Neymotin, L

The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. In this study, a fit of RAMONA-4B bundle powers was used to estimate the local power peaking. It was determined that the peaking factor could be 25% higher than the factor usually assumed for RDA analysis. Combining this error with the 2 sigma random error means that for this analysis the actual fuel rod enthalpy could be 100% larger than calculated by RAMONA-4B. This is much larger than the uncertainty in most parameters that are calculated with best-estimate methods for other design-basis events.

• Research Organization: Brookhaven National Lab., Upton, NY (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC02-76CH00016
• OSTI ID: 465206
• Report Number(s): BNL-NUREG--63663-Rev.; CONF-970315--1-Rev.; ON: DE97004576
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
BURNUP
BWR TYPE REACTORS
CONTROL ELEMENTS
CONTROL ROD WORTHS
DESIGN BASIS ACCIDENTS
ENTHALPY
FUEL PELLETS
HEAT TRANSFER
HYDRAULICS
R CODES
REACTIVITY
REACTOR COOLING SYSTEMS
REACTOR SAFETY
ROD DROP ACCIDENTS
SENSITIVITY ANALYSIS