Information Notice 2004-15, Dual-Unit Scram at Peach Bottom Units 2 and 3
| ML041950006 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/22/2004 |
| From: | Beckner W NRC/NRR/DIPM/IROB |
| To: | |
| Hodge, CV, NRR/DIPM/IROB, 415-1861 | |
| References | |
| IN-04-015 | |
| Download: ML041950006 (6) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555-0001
July 22, 2004
NRC INFORMATION NOTICE 2004-15:
DUAL-UNIT SCRAM AT PEACH BOTTOM
UNITS 2 AND 3
Addressees
All holders of operating licenses for nuclear power reactors except those who have permanently
ceased operation and have certified that fuel has been permanently removed from the reactor
vessel.
Purpose
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert
addressees to recent experience in which a dual unit facility lost offsite power, had a dual unit
scram, and experienced other problems including the loss of a common emergency diesel
generator (EDG). It is expected that recipients will review this information for applicability to
their facilities and consider actions, as appropriate. However, suggestions contained in this
information notice are not NRC requirements; therefore, no specific action or written response
is required.
Description of Circumstances
On September 15, 2003, offsite power to the emergency buses at Peach Bottom Units 2 and 3 was lost for about 16 seconds when two of the three offsite power sources were briefly lost. All
four EDGs automatically started and supplied power to the emergency buses. The third offsite
power source remained available to two of the four plant non-emergency plant buses
throughout the event.
The offsite power grid dispatcher notified the control room that the portion of the offsite power
that was supplying the emergency buses was available half an hour after the event started.
However, because the emergency buses were powered from the EDGs and plant transient
response actions were the operational priority, operators did not transfer from the EDGs to
offsite power for several hours until they were more certain of the reliability of the offsite power
source. The licensee determined that the loss of offsite power was the result of a lightning
strike approximately 35 miles northeast of the site.
Before the event, Unit 2 was operating at full power and Unit 3 was operating at 91 percent of
full power. Both units automatically scrammed when power was lost to the reactor protection
system motor generator sets. Containment isolation signals resulted in the closure of the main
steam isolation valves and isolation of each reactor from its normal heat sink, the condenser.
All four EDGs automatically started and supplied power to the emergency buses; each EDG
supplies power for two buses (one per unit). The licensee was able to safely bring both units
into the cold shutdown condition. However, the shutdown of each unit was complicated both by
equipment challenges and by procedural problems. The NRC organized an Augmented
Inspection Team (AIT) because of the overall risk significance of the event and multiple failures
in systems used to mitigate the event. The AIT mission was to determine the causes, conditions, and circumstances relevant to issues directly related to the event and to assess the
safety significance of the event (NRC Augmented Inspection Team Report 05000277/2003013 and 05000278/2003013, ADAMS Accession No. ML033530016).
Discussion:
The most significant equipment problem during this event was the unexpected E2 EDG trip
during the cooling of the Unit 2 torus while other EDGs were supplying power to the emergency
buses. The E2 EDG shut down due to an engine protective trip initiated by low jacket water
pressure. The AIT found that combustion gases entered the jacket water coolant system
because of one or more leaking cylinder adapter gaskets, causing low jacket water pressure
and automatic shutdown of the E2 EDG. The leakage was due to deficient installation
procedures and stress relaxation of the cylinder adapter gaskets. These adapter gaskets, made of copper, provide a seal between high-pressure gases in each cylinder and the jacket
water system. The licensee concluded that the root cause was inadequate initial pre-loading
combined with the natural stress relaxation of the copper over time. The licensee has four
Fairbanks Morse 12 cylinder, opposed piston diesel engines for both units.
The AIT found that the EDG cylinder liner replacement procedure did not incorporate adequate
guidance to ensure proper sealing of the cylinder liner adapter gaskets. The gaskets relaxed
over several years, allowing combustion gases to enter the jacket coolant system. Additionally, the licensee may have missed opportunities associated with jacket water anomalies. Degraded
conditions, such as jacket water leaks and high vibration on the E2 EDG from 1996-2002, were
tolerated and a condition adverse to quality following two instances of low jacket water pressure
was not corrected.
The licensee performed a number of corrective actions to remedy the EDG gasket problem.
The licensee replaced all adapter gaskets on the tripped EDG, inspected the cylinders during
hydrostatic testing, temporarily installed a sight glass to ensure no combustion gas leakage, revised test and maintenance procedures, and sampled the expansion tank air space and
jacket coolant heat exchanger for combustion gases. The final three actions were performed
on all the EDGs.
The AIT found that the maintenance procedure for installing the cylinder liner adapter gaskets on
the EDGs was deficient and that the licensee took inadequate corrective actions for the low
jacket water pressure conditions observed on the E2 EDG in March and April 2003. Using the
reactor safety Significance Determination Process (SDP), the AIT determined this incident to be
a White finding for Unit 2 (i.e., a low-to-moderate safety-significant finding that may require
additional NRC inspection) and a Green finding for Unit 3. The difference in risk significance
between the units is due to differences in electrical bus loads. The Unit 2 transient was complicated by the following factors:
1.
Due to the momentary loss of offsite power, the controlling channel of the Unit 2 condenser hotwell level instrumentation failed low. This previously identified equipment
deficiency resulted in the draining of the Unit 2 condensate storage tank to the Unit 2 condenser hotwell. The condensate storage tank is the preferred common suction of two
Unit 2 mitigating systems, the high-pressure coolant injection system and the reactor
core isolation cooling system. As a result of the decreasing condensate storage tank
level, the suction of these mitigating systems automatically but unexpectedly changed
from the Unit 2 condensate storage tank to the Unit 2 torus.
2.
As a result of the transient, the Unit 2 torus water heated up, necessitating the use of
residual heat removal (RHR) pumps to cool the Unit 2 torus. At the Peach Bottom site, there are 4 RHR pumps per unit and 4 EDGs common to both units. Therefore, 1 RHR
pump from each unit is associated with 1 EDG. A minimum of 1 of the 4 RHR pumps per
unit is required to satisfy the containment cooling design function. The licensee needed
to use a minimum of 1 RHR pump on both units but was prohibited from simultaneously
using pumps powered by the same electrical division, that is, off the same EDG. Thus
the Unit 2 A RHR pump and the Unit 3 A RHR pump could not be used at the same time.
This is due to electrical load restrictions on the EDG that supplies the same electrical
division for both Units 2 and 3. This is a known design limitation of the Peach Bottom
station involving the significant electrical load requirements for operating the RHR pump
motors.
3.
On isolation of the Unit 2 condenser hotwell due to closure of the main steam isolation
valves, the associated E2 EDG unexpectedly tripped, stopping the Unit 2 torus cooling.
The E2 EDG tripped on low jacket water coolant pressure, which stopped the inservice
RHR pump and drained the B torus cooling loop, reducing the availability of torus cooling
on Unit 2.
4.
A number of other deficiencies complicated operator response and recovery actions.
The Unit 3 transient was complicated by several different factors:
1.
The Unit 3 D safety relief valve opened as designed on high reactor pressure but failed to
close at the appropriate decreasing reactor pressure setpoint. Over the next 15 minutes, reactor pressure decreased to 369 psig before the valve closed, which allowed injection
by condensate pumps and an increase in reactor water level to the high-level setpoint
before operators manually tripped these pumps. The valve closed with no operator
action. The cause of the initial failure of the valve to close was determined to be pilot
valve leakage.
2.
The Unit 3 G safety relief valve initially opened automatically on high reactor pressure as
designed and was subsequently remotely operated to control reactor pressure.
However, on a reactor pressure control operation much later in the event, the valve failed
to open on demand from the main control board control switch. The cause of the failure of the valve to open was determined to be steam leaking through the valve packing into
the air operator. The steam damaged the diaphragm of the air operator and prevented
the valve from manually operating.
3.
The Unit 3 D outboard main steam isolation valve failed to close upon receipt of the
Group I isolation signal, remained open for 76 minutes, and then closed with no operator
action. The redundant inboard main steam isolation valve appropriately closed as
designed.
4.
A number of other deficiencies complicated operator response and recovery actions.
This information notice requires no specific action or written response. If you have any
questions about information in this notice, please contact one of the technical contacts listed
below or the appropriate NRR project manager.
/RA/
William D. Beckner, Chief
Reactor Operations Branch
Division of Inspection Program Management
Office of Nuclear Reactor Regulation
Technical contacts:
Dr. C. Vernon Hodge, NRR
Neil Perry, Region I
(301) 415-1861
(610) 337-5225 Email: cvh@nrc.gov
Email: nsp@nrc.gov
Attachment: List of Recently Issued NRC Information Notice
ML041950006 DOCUMENT NAME: C:\\ORPCheckout\\FileNET\\ML041950006.wpd
OFFICE
NRR/DIPM/IROB
Tech Editor
RI
BC:SPLB:DSSA
(A)SC:OES:IROB:DIPM
C:IROB:DIPM
NAME
CVHodge*
CV for
NSPerry
JNHannon*
AMcMurtray
WDBeckner
DATE
05/27/2004
05/24/2004
06/03/2004
06/01/2004
07/22/2004
07/22/2004
______________________________________________________________________________________
OL = Operating License
CP = Construction Permit
Attachment LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
_____________________________________________________________________________________
Information
Date of
Notice No.
Subject
Issuance
Issued to
_____________________________________________________________________________________
2004-14
Use of less than Optimal
Bounding Assumptions in
Criticality Safety Analysis at
Fuel Cycle Facilities
07/19/2004
All licensees authorized to
possess a critical mass of special
nuclear material.
2004-13
Registration, Use, and Quality
Assurance Requirements for
NRC-Certified Transportation
Packages
06/30/2004
All materials and
decommissioning reactor
licensees.
2004-12
Spent Fuel Rod Accountability
06/25/2004
All holders of operating licenses
for nuclear power reactors, research and test reactors, decommissioned sites storing
spent fuel in a pool, and wet
spent fuel storage sites.
2004-11
Cracking in Pressurizer Safety
and Relief Nozzles and in
Surge Line Nozzle
05/06/2004
All holders of operating licenses or
construction permits for nuclear
power reactors, except those that
have permanently ceased
operations and have certified that
fuel has been permanently
removed from the reactor.
2004-10
Loose Parts in Steam
Generators
05/04/2004
All holders of operating licenses
for pressurized-water reactors
(PWRs), except those who have
permanently ceased operations
and have certified that fuel has
been permanently removed from
the reactor.
Note:
NRC generic communications may be received in electronic format shortly after they are
issued by subscribing to the NRC listserver as follows:
To subscribe send an e-mail to <listproc@nrc.gov >, no subject, and the following
command in the message portion:
subscribe gc-nrr firstname lastname