Analyzing the rod drop accident in a boiling water reactor

Journal Article · Fri Jan 01 04:00:00 EST 1982 · Nucl. Technol.; (United States)

OSTI ID:7068678

Cheng, H S; Diamond, D J

The center rod drop accident was calculated for a boiling water reactor using the two-dimensional (R,Z) core dynamics code BNL-TWIGL. Analysts frequently neglect moderator feedback under the assumption that it leads to conservative results. The present study shows that the peak of the power burst and peak fuel enthalpy can indeed be reduced by a factor of 2 or more by including this effect. The magnitude of the effect depends on reactor conditions. Moderator feedback is particularly important when there are voids in the core initially (i.e., at power conditions) or when the core is near saturation condition. When the reactor is initially at zero power and considerably subcooled, moderator feedback will influence the power peak by < 10% but will have a much larger effect on the peak fuel enthalpy, which occurs later in time. The moderator feedback is the result of heat conducted from the fuel rod and direct energy deposition. At power conditions, the time constant for heat conduction is small and this is the primary mechanism for changing the steam void content during the accident. At zero power, the initial thermal constant is very large and, hence, any generation of voids at short times is due to direct energy deposition in the moderator. The effect of a different initial power level, flow rate, and inlet subcooling, as well as the effect of delayed neutron fraction, rod drop speed, and accident rod worth, was calculated. In all cases, with moderator feedback accounted for, the maximum fuel enthalpy during the accident is well below presently established limits. Accident consequences are insensitive to the delayed neutron fraction and rod drop velocity. The parameters of most significance are inlet subcooling and accident rod worth.

• Research Organization: Brookhaven National Laboratory, Department of Nuclear Energy Upton, New York 11973
• OSTI ID: 7068678
• Journal Information: Nucl. Technol.; (United States), Journal Name: Nucl. Technol.; (United States) Vol. 56:1; ISSN NUTYB
• Country of Publication: United States
• Language: English

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ABSORPTION
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B CODES
BARYONS
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MODERATORS
NEUTRONS
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PHYSICAL PROPERTIES
POWER COEFFICIENT
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REACTIVITY INSERTIONS
REACTIVITY WORTHS
REACTOR ACCIDENTS
REACTOR COMPONENTS
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