Probability and Consequences of a Rapid Boron Dilution Sequence in a PWR

Conference · Wed Oct 18 04:00:00 EDT 1995

OSTI ID:117794

Diamond, D. J. ^[1]; Kohut, P. ^[1]; Nourbakhsh, H. ^[1]; Valtonen, K. ^[2]; Secker, P. ^[3]

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1. Brookhaven National Lab. (BNL), Upton, NY (United States)
2. Finnish Centre for Radiation and Nuclear Safety, Helsinki, (Finland)
3. Univ. of Arizona, Tucson, AZ (United States). Dept. of Nuclear & Energy Engineering

The reactor restart scenario is one of several beyond-design-basis events in a pressurized water reactor (PWR) which can lead to rapid boron dilution in the core. This in turn can lead to a power excursion and the potential for fuel damage. A probabilistic analysis had been done for this event for a European PWR. The estimated core damage frequency was found to be high partially because of a high frequency for a LOOP and assumptions regarding operator actions. As a result, a program of analysis and experiment was initiated, and corrective actions were taken. A system was installed so that the suction of the charging pumps would switch to the highly borated refueling water storage tank (RWST) when there was a trip of the RCPs. This was felt to reduce the estimated core damage frequency to an acceptable level. In the US, this original study prompted the Nuclear Regulatory Commission to issue an information notice to follow work being done in this area and to initiate studies such as the work at BNL reported herein. In order to see if the core damage frequency might be as high in US plants, a probabilistic assessment of this scenario was done for three plants. Two important conservative assumptions in this analysis were that (1) the mixing of the injectant was insignificant and (2) fuel damage occurs when the slug passes through the core. In order to study the first assumption, analysis was carried out for two of the plants using a mixing model. The second assumption was studied by calculating the neutronic response of the core to a slug of deborated water for one of the plants. All three types of analyses are summarized below. More information is available in the original report.

• Research Organization: Brookhaven National Laboratory (BNL), Upton, NY (United States)
• Sponsoring Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States); USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP)
• DOE Contract Number: AC02-76CH00016
• OSTI ID: 117794
• Report Number(s): BNL-NUREG--61713; CONF-9510225--2; ON: TI96002010
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
BORON
CRITICALITY
Core Damage Frequency (CDF)
DILUTION
EXCURSIONS
HEAT TRANSFER
HYDRAULICS
Loss of Onsite Power (LOOP)
Nuclear Criticality Safety Program (NCSP)
OUTAGES
PROBABILITY
PWR TYPE REACTORS
Pressurized Water Reactor (PWR)
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTOR SAFETY
REACTOR START-UP
Reactor Coolant Pumps (RCPs)
Reactor Coolant System (RCS)
Refueling Water Storage Tank (RWST)
Thermal-Hydraulic Analysis
Volume Control Tank (VCT)