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. 

ML041950006
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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.
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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 
IN 2004-15

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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
IN 2004-15

Page 4 of 4

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

DISTRIBUTION:

ADAMS

IN File

ADAMS ACCESSION NUMBER: 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

OFFICIAL RECORD COPY
______________________________________________________________________________________

OL = Operating License

CP = Construction Permit

Attachment

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Page 1 of 1

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.

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