Information Notice 2010-13, Failure to Verify That Post-Fire-Shutdown Procedures Can Be Performed
| ML101120816 | |
| Person / Time | |
|---|---|
| Issue date: | 07/22/2010 |
| From: | Mcginty T, Tracy G Office of New Reactors, Office of Nuclear Reactor Regulation |
| To: | |
| Beaulieu, D P, NRR/DPR, 415-3243 | |
| References | |
| IN-10-013 | |
| Download: ML101120816 (5) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
OFFICE OF NEW REACTORS
WASHINGTON, DC 20555-0001 July 22, 2010
NRC INFORMATION NOTICE 2010-13: FAILURE TO ENSURE THAT POST-FIRE
SHUTDOWN PROCEDURES CAN BE
PERFORMED
ADDRESSEES
All holders of an operating license or construction permit for a nuclear power reactor issued
under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of
Production and Utilization Facilities, except those who have permanently ceased operations
and have certified that fuel has been permanently removed from the reactor vessel.
All holders of or applicants for an early site permit, standard design certification, standard
design approval, manufacturing license, or combined license issued under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
PURPOSE
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
addressees about recent NRC inspection findings on post-fire procedures and the need to
ensure that required steps can be implemented as directed. The NRC expects that recipients
will review the information for applicability to their facilities and consider taking action, as
appropriate, to avoid similar issues. However, no specific action or written response is required.
BACKGROUND
Operating nuclear power plants licensed to operate before January 1, 1979, are required to
implement Section III.G, Fire Protection of Safe Shutdown Capability, of Appendix R, Fire
Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, to
10 CFR Part 50. In Section III.G.1, the NRC requires that one train of systems necessary to
achieve and maintain safe shutdown from either the control room or the emergency control
stations is free of fire damage. Appendix R requires that, if a licensee cannot meet the
separation criteria specified in Section III.G.2, and if redundant trains of safe-shutdown cables
or equipment are in the same fire area, the licensee must implement the alternative shutdown
requirements of Section III.G.3. A Federal court1 has held that, if a licensee implements the
requirements of Section III.G.3, the licensee must also comply with Section III.L, Alternative
and Dedicated Shutdown Capability, of Appendix R to 10 CFR Part 50. In Section III.L.3, the
NRC requires that procedures be in effect to implement the alternate or dedicated shutdown
1 nd
Connecticut Light and Power, et al., v. NRC, 673 F.2 , 525 (D.C. Cir. 1982).
capability. Plants licensed to operate after January 1, 1979, had similar requirements
incorporated into their operating license.
In addition, many licensee technical specifications include a requirement in the administrative
controls section to establish, implement, and maintain the procedures for fire protection program
implementation and the procedures specified in Appendix A, Typical Procedures for
Pressurized Water Reactors and Boiling Water Reactors, of Regulatory Guide 1.33, Quality
Assurance Program Requirements (Operation), Revision 2, issued February 1978 (Agencywide
Documents Access and Management System (ADAMS) Accession No. ML003739995). Two
procedures specified in Section 6, Procedures for Combating Emergencies and Other
Significant Events, of Appendix A to Regulatory Guide 1.33 are Fire in Control Room or
Forced Evacuation of Control Room and Plant Fires.
DESCRIPTION OF CIRCUMSTANCES
Browns Ferry Nuclear Plant, Units 1, 2, and 3
On December 23, 2008, the licensee at Browns Ferry Nuclear Plant revised the post-fire
safe-shutdown procedure to add an entry condition that would have prevented or delayed the
operators from entering the post-fire safe-shutdown procedure during an Appendix R fire event
if reactor water level stayed at or above +2 inches. As long as operators could maintain reactor
water level during a fire event, they would continue to use the Emergency Operating Instructions
in lieu of the post-fire safe-shutdown instructions. The additional entry condition could prevent
or delay entry into the safe shutdown procedures such that the timing of operator actions, needed to ensure reactor core and containment cooling functions are met, could be inconsistent
with the timeline assumed in the safe-shutdown analysis. Additional information is available in
Browns Ferry Nuclear Plant - NRC Integrated Inspection Report 05000259/2009002,
05000260/2009002 and 05000296/2009002, and Annual Assessment Meeting Summary, dated
April 30, 2009 (ADAMS Accession No. ML091210243).
Brunswick Steam Electric Plant, Units 1 and 2
On August 18, 2008, a surveillance test failure at the Brunswick Steam Electric Plant revealed
that the emergency diesel generators (EDGs) would not operate with the alternate safe- shutdown (ASSD) key switch in the local position. The cause of the failure was a wiring error
that occurred during the installation of a circuit modification to the diesel control system in June
2007. The wire segment number on either side of the ASSD local-normal key switch was the
same. The individual selecting the point for the wiring change chose the correct wire number
but the wrong locations on the ASSD local-normal key switches. This had the unintended effect
of removing control power from the circuitry whenever control of the associated emergency
diesel was transferred from the control room to the local panel. An extent of condition review
determined that the remaining EDGs were also affected. Licensee corrective actions included
rewiring and testing each affected EDG. This event illustrates the importance of required quality
assurance design control measures for verifying or checking the adequacy of design changes, particularly for design changes on risk-significant systems. Additional information is available in
Brunswick Steam Electric PlantNRC Special Inspection Report 05000325/2008010 and
05000324/2008010, dated December 12, 2008 (ADAMS Accession No. ML083470550). Cooper Nuclear Station
During an NRC inspection at Cooper Nuclear Station completed on June 15, 2007, NRC
inspectors found that the licensees post-fire safe-shutdown procedures were inadequate.
Licensee post-fire safe-shutdown procedures required operators to stroke many motor-operated
valves to the required positions from each motor-operated valves motor starter. The procedure
directed operators to open the motor-operated valve motor starter cabinet, remove the control
power fuses, then press an open or closed contactor for a specified amount of time to stroke the
valve to the required position. The NRC inspectors found that the operator at the motor starter
had no indications to confirm that the valve had stroked to the desired position, and the
procedure did not direct the operator to verify the valves position locally. With the control power
fuses removed, valve position indication in the control room was also not available. The NRC
inspectors evaluated four 125-volt direct current motor-operated valves that had motor starter
cubicles that were atypical in that the motor starters were designed without separate control
power fuses. The NRC inspectors found that the valves would not have stroked using the
procedure instructions because removing the fuses would remove motive power. With no
indications and no procedure step to verify position locally, the operator would be unaware
these valves had not stroked.
The licensees extent of condition review identified that the procedure was inadequate for
operating six additional motor-operated valves that had motor starters with separate control
power fuses and three or four contactors. The procedure directed operators to depress the
open or closed contactor. However, the valves would not have actually stroked to the required
positions because additional contactors needed to be operated for the valves to stroke.
Additional information is available in NRC Triennial Fire Protection Inspection Report 05000298/2007008, dated February 1, 2008, and NRC Triennial Fire Protection Followup
Inspection Report 05000298/2008007, dated March 19, 2008 (ADAMS Accession Nos.
ML080350425 and ML080790476, respectively).
DISCUSSION
As discussed in the Background section above, fire protection regulations require that an
operating nuclear power plant be able to achieve safe-shutdown conditions following a fire. This
IN gives examples of post-fire safe-shutdown procedures that were inadequate to ensure that
operators could implement the procedure steps as directed within the assumed timeframe and
with the expected plant equipment response in order to achieve and maintain safe shutdown.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or to the appropriate Office of Nuclear Reactor
Regulation project manager.
/RA by TQuay for/ /RA by JTappert for/
Timothy McGinty, Director Glenn Tracy, Director
Division of Policy and Rulemaking Division of Construction Inspection and
Office of Nuclear Reactor Regulation Operational Programs
Office of New Reactors
Technical Contacts: Phil M. Qualls, NRR John M. Mateychick, RIV
817-276-6550 817-275-6560
E-mail: phil.qualls@nrc.gov E-mail: john.mateychick@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
CONTACT
This IN requires no specific action or written response. Please direct any questions about this
matter to the technical contacts listed below or to the appropriate Office of Nuclear Reactor
Regulation project manager.
/RA by TQuay for/ /RA by JTappert for/
Timothy McGinty, Director Glenn Tracy, Director
Division of Policy and Rulemaking Division of Construction Inspection and
Office of Nuclear Reactor Regulation Operational Programs
Office of New Reactors
Technical Contacts: Phil M. Qualls, NRR John M. Mateychick, RIV
817-276-6550 817-275-6560
E-mail: phil.qualls@nrc.gov E-mail: john.mateychick@nrc.gov
Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under Electronic Reading Room/Document Collections.
ADAMS Accession Number: ML101120816 ME1663 OFFICE NRR/DRA/AFPB TECH EDITOR NRR/DRA/AFPB NRR/DRA/AFPB NRR/DRA
NAME PQualls KAKribbs DFrumkin AKlein MCunningham
DATE 5/7/2010 e-mail 5/17/2010 e-mail 06/23/2010 06/28/2010 06/28/2010
OFFICE NRR/DPR/PGCB NRR/DPR/PGCB NRR/DPR/PGCB NRO/DCIP NRR/DPR
NAME CHawes DBeaulieu SRosenberg GTracy JTappert for TMcGinty TQuay for
DATE 6/30/2010 6/30/2010 7/6/2010 7/7/2010 7/22/2010
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