Analysis of Reverse Flow Restriction Device to Prevent Dryout Fuel Damage during BWR Instability

Journal Article · Wed Jun 15 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:22992019

Radaideh, Majdi Ibrahim; Kozlowski, Tomasz ^[1]; Farawila, Yousef M. ^[2]

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1. University of Illinois at Urbana-Champaign, 104 South Wright Street, Urbana, Illinois, 61801 (United States)
2. Farawila et al., Inc., 106 Edgewood Drive, Richland, Washington, 99352 (United States)

Boiling Water Reactors (BWRs) are subjected to instabilities due to thermal hydraulics feedback or coupled Neutronics-thermal hydraulics feedback. Feedback is affected by many reactor parameters that could destabilize the core such as distribution of pressure drop across the core, axial power shape, subcooling level at the core inlet, radial power peaking factor, reactivity coefficients, and others. There are three common modes of instability in BWRs: single channel instability, core-wide instability, and regional instability. Each instability mode could damage the fuel if the power oscillations are left without an intervention like reactor scram. However, the instability problem becomes more dangerous if the automatic reactor scram is lost, like in Anticipated Transient without Scram (ATWS) accidents. A reverse flow restriction device (RFRD) would mitigate this problem. Core simulation using TRACE/PARCS code has been used to examine the potential effect of restricting the flow in the downward direction on mitigation of the power and flow oscillations to increase the safety of BWRs. The presented results demonstrate the ability of the device on limiting flow oscillation amplitude in the one bundle where it is used, on an adjacent bundle without it and reducing the power peak when the flow is low and power is high during BWR power oscillations. This exercise demonstrates the potential of previously described hardware device on preventing clad temperature excursion to keep the fuel safe from dryout damage and provide encouraging results ahead of the planned more detailed investigation where the device is to be used in more bundles up to full core. In the future work it is planned to examine the effect of device installed in a fraction of the bundles on the flow rate in other bundles. Also, the effect on the characteristics of dryout and rewetting and the resulting temperature excursions with and without the device will be given special attention. (authors)

• OSTI ID: 22992019
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Journal Issue: 1 Vol. 114; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
42 ENGINEERING
97 MATHEMATICS AND COMPUTING
ATWS
BWR TYPE REACTORS
DAMAGE
DRYOUT
EQUIPMENT
EXCURSIONS
FEEDBACK
FLOW RATE
NUCLEAR FUELS
PRESSURE DROP
REACTIVITY COEFFICIENTS
SCRAM
SUBCOOLING
THERMAL HYDRAULICS