Analyzing the BWR rod drop accident in high-burnup cores

Conference · Tue Aug 01 04:00:00 EDT 1995

OSTI ID:104426

Diamond, D J; Neymotin, L; Kohut, P

This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ``rim`` effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions.

• 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: 104426
• Report Number(s): BNL-NUREG--62068; CONF-9509208--1; ON: DE95016503
• 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
COMPUTER CALCULATIONS
CONTROL ELEMENTS
CONTROL ROD WORTHS
ENTHALPY
EXCURSIONS
FEEDBACK
FUEL ELEMENTS
HEAT TRANSFER
HYDRAULICS
POWER DISTRIBUTION
R CODES
REACTOR ACCIDENTS
REACTOR CONTROL SYSTEMS
REACTOR KINETICS
REACTOR SAFETY
VOID FRACTION