ML101120816 

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

 Connecticut Light and Power, et al., v. NRC, 673 F.2nd, 525 (D.C. Cir. 1982). 
IN 2010-13 

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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 safeshutdown (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 Plant—NRC Special Inspection Report 05000325/2008010 and 

05000324/2008010, dated December 12, 2008 (ADAMS Accession No. ML083470550). 
IN 2010-13 

Page 3 of 4 

Cooper Nuclear Station

During an NRC inspection at Cooper Nuclear Station completed on June 15, 2007, NRC 

inspectors found that the licensee’s 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 valve’s 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 valve’s 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 licensee’s 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. 
IN 2010-13 

Page 4 of 4 

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. 
IN 2010-13 

Page 4 of 4 

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