Information Notice 2010-13, Failure to Verify That Post-Fire-Shutdown Procedures Can Be Performed

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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

OFFICIAL RECORD COPY