Morning Report for November 2, 1999

                       Headquarters Daily Report

                         NOVEMBER 02, 1999

                    REPORT             NEGATIVE            NO INPUT
                    ATTACHED           INPUT RECEIVED      RECEIVED

REGION I                               X
REGION II                              X
REGION III                             X
REGION IV                              X

MR Number: H-99-0096

                           NRR DAILY REPORT ITEM
                            SIGNIFICANT EVENTS

Subject: Scram and Partial Loss of Vital Power at Indian Point, Unit 2,
         Classified As A Significant Event

The Indian Point, Unit 2, August 31, 1999 event is classified as a
Significant Event for the NRC Performance Indicator Program. The bases
for this classification were the number of complications that resulted
and produced unnecessary burdens on licensees operational personnel
coupled with the lapses in configuration control and management oversight
that the event demonstrated.

On August 31, 1999, at 2:31 p.m., the Indian Point Unit 2 reactor tripped
from 99 percent power, the trip indication was Over-Temperature
Delta-Temperature. About three minutes after the reactor trip, the normal
offsite power breakers to all four 480 volt vital buses tripped; all
three emergency diesel generators (EDGs) started and began to load. A
short time later, the 23 EDG output breaker tripped, leaving the 6A vital
bus de-energized. This resulted in a loss of power to one of the two
motor-driven auxiliary feedwater pumps, to battery charger 24 and to some
emergency core cooling components. The bus remained de-energized while
technicians prepared tagouts and checked for a suspected bus fault which
could have caused the loss of power. Battery 24 discharged over the next
seven hours causing a loss of power to the loads on dc panel 24 and on ac
instrument bus 24. The loss of the dc bus rendered half of the bleed and
feed capability of the unit inoperable if it had been required and
complicated the emergency feedwater flow control process. The loss of the
ac instrument bus disabled most control room annunciators for safety
related systems. The next day, about 1:00 p.m., vital bus 6A was
re-energized using the EDG; by 9:00 p.m., normal offsite power had been
restored to all buses and the three EDGs secured.

On September 1, 1999, the NRC initiated an Augmented Team Inspection to
examine the circumstances surrounding the event principally because the
event was complicated by significant, unexpected system interactions
involving safety related equipment; also, the team was to examine the
adequacy of the licensee's response to the event particularly the vital
bus power restoration efforts--the delays in which led to significant
additional complication of the operators' efforts to stabilize the plant.
The team report is provided in Inspection Report 50-247/99-08 dated
October 19, 1999.

The team found that a primary cause of the event was inadequacy in
configuration control management; the loss of bus 6A and subsequent
degradation of plant conditions were caused by two equipment
configuration control problems: the station auxiliary transformer load
tap changer having been left in the "Manual" position; and the improper
overcurrent trip setting for emergency diesel generator 23 output
breaker. In both cases, station personnel failed to ensure the equipment
configuration was controlled as specified in the licensing and design
bases. The team also concluded that lapses in management oversight
HEADQUARTERS      MORNING REPORT     PAGE  2          NOV. 02, 1999
MR Number: H-99-0096 (cont.)

significantly contributed to the event. Station management missed
significant opportunities to recognize and fully assess degrading plant
conditions and failed to establish viable plans and contingencies for
plant restoration.

The results of the licensee and NRC risk evaluations of this event were
similar. The risk estimates were conservative in that no credit was given
for "bleed and feed" cooling, the #23 auxiliary feedwater pump was
considered unrecoverable, and a low probability of success was assigned
for the operators using the feedwater system to provide make-up to the
steam generators. Based on these conservative assumptions, the calculated
conditional core damage probability (CCDP) was 2x10-4. The CCDP is used
to estimate the risk significance of conditions or events.

Contact:    Ed Goodwin, NRR

Page Last Reviewed/Updated Wednesday, March 24, 2021