Text

IR 05000400-11-011, on 08/29/2011 - 09/02/2011 & 09/19-23/2011; Shearon Harris Nuclear Plant, Unit 1; Triennial Fire Protection Inspection Report & Exercise of Enforcement Discretion
	ML113420167
Person / Time
Site:	Harris
Issue date:	12/08/2011
From:	Joel Munday; Division of Reactor Safety II
To:	Jefferson W; Carolina Power & Light Co
References
	EA-11-274 IR-11-011
	Download: ML113420167 (43)
	v • d • e

December 08, 2011

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000400/2011011 AND EXERCISE

OF ENFORCEMENT DISCRETION

Dear Mr. Jefferson:

On September 23, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed a triennial fire protection inspection at your Shearon Harris Nuclear Power Plant. The enclosed inspection report documents the inspection results, which were discussed on that date with you and members of your staff. On December 8, 2011, a final exit meeting was conducted to discuss the results of this inspection with Mr. Corlett, and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

The enclosed report documents one NRC-identified finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements. However, because of the very low safety significance and because it was entered into your corrective action program, the NRC is treating the violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the NRC Enforcement Policy. If you contest this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Senior Resident Inspector at the Shearon Harris nuclear power plant. In addition, if you disagree with the crosscutting aspect assigned to this finding, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Senior Resident Inspector at the Shearon Harris nuclear power plant.

CP&L The enclosed report also documents three noncompliances for which the NRC is exercising enforcement discretion in accordance with Section 9.1 of the NRC Enforcement Policy, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48). The noncompliances are associated with your implementation of the requirements and standards of Harris Operating License Condition 2.F, Fire Protection Program. The NRC is not taking enforcement action for these noncompliances because they meet the criteria of the NRCs Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48). In this case, the NRC concluded that: (1) Carolina Power & Light entered the noncompliances into its corrective action program and implemented appropriate compensatory measures; (2) the noncompliances were not associated with a finding that the reactor oversight process significance determination process would evaluate as Red; (3) the noncompliances were not willful; and (4) Carolina Power & Light submitted a letter of intent, before December 31, 2005, stating its intent to transition to 10 CFR 50.48(c), which includes approaches in National Fire Protection Association Standard 805-2001 Edition (NFPA 805). To address these noncompliances, corrective actions and plant modifications were implemented as part of your transition to NFPA 805. Based on these corrective actions, and in accordance with NRC Inspection Manual Chapter 0305, Section 11.05, Treatment of Items Associated with Enforcement Discretion, subsections 11.05.b and 11.05.c, the NRC will refrain from including the noncompliances in the Agency Action Matrix.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at:

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Joel T. Munday, Director

Division of Reactor Safety

Docket No.: 50-400 License No.: NPF-63

Enclosure:

Inspection Report 05000400/2011011

w/Attachment: Supplemental Information

REGION II==

Docket No.:

50-400

License No.:

NPF-63

Report No.:

05000400/2011011

Licensee:

Carolina Power and Light Company

Facility:

Shearon Harris Nuclear Power Plant, Unit 1

Location:

5413 Shearon Harris Road

New Hill, NC 27562

Dates:

August 29 - September 2 (Week 1)

September 19 - 23 (Week 2)

Inspectors:

R. Rodriguez, Senior Reactor Inspector (Lead Inspector)

P. Braxton, Reactor Inspector B. Correll, Reactor Inspector (Region IV)

N. Merriweather, Senior Reactor Inspector M. Thomas, Senior Reactor Inspector

G. Wiseman, Senior Reactor Inspector

J. Montgomery, Reactor Inspector (in-office review of open items, Section 4OA5)

Accompanying H. Barrett, Sr. Fire Protection Engineer (NRR)

Personnel:

G. MacDonald, Senior Reactor Analyst D. Orr, Senior Reactor Inspector (Region I)

Approved by:

Rebecca L. Nease, Chief

Engineering Branch 2

Division of Reactor Safety

Enclosure

SUMMARY OF FINDINGS

IR 05000400/2011011: August 29 - September 2, 2011 and September 19 - 23, 2011; Shearon

Harris Nuclear Plant, Unit 1; Triennial Fire Protection Inspection.

This report covers an announced two-week Triennial Fire Protection Inspection (TFPI) by a team of six regional inspectors as well as an in-office review of open items by one regional inspector. One NRC-identified finding of very low safety significance (Green) was identified.

The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross cutting aspects are determined using IMC 0310, Components within the Cross Cutting Areas.

Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor oversight Process, Revision 4, dated December 200

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The team identified a non-cited violation of Harris Nuclear Plant Technical Specification 6.8.1.a. for inadequate guidance in fire response abnormal operating procedure AOP-036.04, Fire Areas: 1-A-BAL-C, 1-A-BAL-D, 1-A-BAL-F, 1-G, FPYARD,

Revision 17. Specifically, the procedure could not have been performed as written in that, AOP-036.04, Section 3.1, directed operators to implement a step in the procedure that did not exist. The licensee initiated Nuclear Condition Report 489092 to address this issue in the Corrective Action Program and subsequently revised the procedure.

The team determined that inadequate fire response procedure guidance was a performance deficiency. This finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and it affected the cornerstone objective of protection against external events (i.e., fire). The team assessed this finding using IMC 0609, Appendix F, Fire Protection Significance Determination Process. The team assigned a low degradation rating to this finding because the abnormal operating procedure deficiency was compensated by available emergency operating procedure guidance, operator experience/familiarity, and training.

It was likely that plant operators would have been able to assess plant parameters and would have taken the appropriate actions required to ensure post-fire safe and stable plant conditions. Therefore, this finding was of very low safety significance (Green).

The cause of this finding was determined to have a cross cutting aspect in the Human Performance Area, Resources Component, because the licensees validation and verification process did not ensure that the procedure was adequate and accurate.

(H.2 (c)). (Section 1R05.01.b)

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R05 Fire Protection

This report documents the results of a Triennial Fire Protection Inspection (TFPI) at the Shearon Harris Nuclear Power Plant (HNP), Unit 1. The inspection was conducted in accordance with NRC Inspection Procedure (IP) 71111.05XT, Fire Protection - NFPA 805 (Triennial) dated August 1, 2011. Additionally, IP 71111.17, Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications-dated October 31, 2008, was used to review 14 engineering changes (EC). These ECs were deemed necessary to complete the transition to full compliance with 10 CFR 50.48(c) by December 31, 2010, in accordance with the transition license condition. The objective of the inspection was to review a minimum sample of two risk-significant Fire Areas (FAs)to verify implementation of the HNP Fire Protection Program (FPP). An additional objective was to verify site specific implementation of at least one B.5.b mitigating strategy as well as the storage, maintenance, and testing of B.5.b mitigating equipment.

Section 71111.05-05 of the IP specifies a minimum sample size of two FAs. The team chose four FAs based on available risk information as analyzed onsite by a senior reactor analyst from Region II, data obtained from in-plant walkdowns regarding potential ignition sources, location, and characteristics of combustibles, and location of equipment needed to achieve and maintain the reactor in a safe and stable condition.

Other considerations for selecting the FAs were the relative complexity of the post-fire safe shutdown (SSD) procedures, information contained in FPP documents, and results of prior NRC TFPIs. In selecting the B.5.b mitigating strategy sample, the inspectors reviewed licensee submittal letters, safety evaluation reports, licensee commitments, B.5.b implementing procedures, and previous NRC inspection reports.

Detailed inspection of these four FAs fulfills the procedure completion criteria. The four areas chosen were:

1-A-SWGRB - Switchgear Room B in Reactor Auxiliary Building which is a performance based fire area with focus on fire scenarios involving the 480V Auxiliary Switchgear 1B1.

1-A-CSRB - Cable Spreading Room B, Reactor Auxiliary Building which is a performance based fire area with focus on fire scenarios involving the Auxiliary Transfer Panel SB.

1-A-BAL-C - Reactor Auxiliary Building Unit 1 - Analysis Area C which is a performance based fire area with focus on fire scenarios involving the 480V 1B21 motor control center (MCC).

1-A-BAL-G - Reactor Auxiliary Building Unit 1 - Analysis Area G, which is a deterministic based fire area.

For each of the selected FAs, the inspection team evaluated the licensees FPP against the applicable NRC requirements. The specific documents reviewed by the team are listed in the Attachment.

.01 Protection of Safe Shutdown Capabilities

a. Inspection Scope

The team examined HNP fire response Abnormal Operating Procedures (AOPs) and compared them to the NFPA 805 Nuclear Safety Capability Assessment (NSCA) and Fire Safety Analysis (FSA), systems flow diagrams, and other design basis documents to determine if equipment required to achieve post-fire safe and stable plant conditions was properly identified and adequately protected from fire damage in accordance with the requirements of 10 CFR 50.48(c) and the HNP approved FPP. Cable routing information by FA was reviewed for a selected sample of SSD components to verify that either the associated cables would not be damaged for the selected FAs/fire scenarios or the licensees analysis determined that the fire damage would not prohibit achieving safe and stable plant conditions. For the one case where the licensee credited a recovery action during non-power operations to meet the nuclear safety performance criteria, the team reviewed the recovery action to verify that it was feasible. A list of SSD components examined for cable routing is included in the Attachment. The specific fire response AOPs reviewed is also listed in the Attachment.

The team reviewed applicable sections of the AOPs for the selected FAs and fire scenarios to verify that the shutdown methodology properly identified the components and systems necessary to achieve and maintain safe and stable plant conditions. The team performed an in-plant walk-through of procedure steps to ensure the implementation and human factors adequacy of the procedures. The team verified the licensee personnel credited for procedure implementation had procedures available, were trained on implementation, and were available in the event a fire occurred. The team also reviewed one operator recovery action and selected operator defense-in-depth actions to verify that the operators could reasonably be expected to perform the specific actions within the time required to maintain plant parameters within specified limits.

b. Findings

Inadequate Fire Response Procedure AOP-036.04

Introduction:

The team identified a Green non-cited violation (NCV) of HNP Technical Specification 6.8.1.a. for inadequate guidance in fire response procedure AOP-036.04, Fire Areas: 1-A-BAL-C, 1-A-BAL-D, 1-A-BAL-F, 1-G, FPYARD, Rev. 17. Specifically, the procedure could not have been performed as written in that, AOP-036.04, Section 3.1, directed operators to implement a step in the procedure that did not exist. The licensee initiated nuclear condition report (NCR) 489092 to address this issue, and subsequently revised the procedure.

Description:

Procedure AOP-036.04, Section 3.1, would have been used concurrently and in conjunction with emergency operating procedure (EOP), EOP PATH-1 to achieve post-fire safe and stable plant conditions from the main control room (MCR) fire scenarios in Fire Area 1-A-BAL-C. The team walked down procedure AOP-036.04 with licensee operations personnel to assess the procedural guidance for post-fire safe and stable conditions. During review and walkdown of the procedure, the team noted that Section 3.1, Step 10.d (in the Response Not Obtained (RNO) column) directed operators to implement Step 10.f. However, there was no Step 10.f in Section 3.1 of AOP-036.04.

The team also noted that Section 3.1 did not have a Step 10.e either. Therefore, the RNO column of Section 3.1, Step 10.d, could not have been performed as written due to inadequate procedural guidance. The team discussed this issue with HNP operations personnel and reviewed the guidance in EOP PATH-1, which would have been used concurrently with AOP-036.04. Based on the discussions and reviews, the team determined that the underlying intent of Step 10.d in AOP-036.04 was to ensure that reactor coolant level was maintained within the indicating range in the pressurizer. The team concluded that based on the EOP guidance and operator experience/training, it was likely the plant operators would have been able to determine pressurizer level and would have taken the appropriate actions required to ensure post-fire safe and stable plant conditions. The licensee initiated NCR 489092 to address this issue.

Analysis:

The team determined that the inadequate fire response procedure guidance was a performance deficiency (PD). This finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and it affected the cornerstone objective of protection against external events (i.e., fire). The team assessed this finding using IMC 0609, Appendix F, Fire Protection Significance Determination Process. The team assigned a low degradation rating to this finding because the AOP procedure deficiencies were compensated by available EOP guidance, operator experience/familiarity, and training. It was likely that plant operators would have been able to assess plant parameters and would have taken the appropriate actions required to ensure post-fire safe and stable plant conditions. Therefore, the team determined that this finding was of very low safety significance (Green). The cause of this finding was determined to have a cross cutting aspect in the Human Performance Area, Resources Component, because the licensees validation and verification process did not ensure that procedure AOP-036.04 was adequate and accurate (H.2(c)).

Enforcement:

The HNP Technical Specification 6.8.1.a. requires that written procedures shall be established, implemented, and maintained covering the activities recommended in Appendix A, of Regulatory Guide 1.33, Rev. 2, dated February 1978. Regulatory Guide 1.33, Appendix A, Section 6.v., requires procedures for combating emergencies such as plant fires. The licensee established procedure AOP-036.04, Fire Areas: 1-A-BAL-C, 1-A-BAL-D, 1-A-BAL-F, 1-G, FPYARD, Rev. 17, for combating plant fires.

Contrary to the above, on September 23, 2011, the team identified that the licensee failed to establish, implement, and maintain procedures for combating emergencies such as plant fires. Specifically, licensee procedure AOP-036.04 provided inadequate instructions for achieving post-fire safe and stable plant conditions for certain fire scenarios in FA 1-A-BAL-C. The licensee initiated NCR 489092 to address this issue in the corrective action program and subsequently revised the procedure. Because this finding was of very low safety significance (Green) and was entered into the licensee=s corrective action program (NCR 489092), this finding is being treated as an NCV, consistent with Section 2.3.2, of the NRC Enforcement Policy. This finding is identified as NCV 05000400/2011011-01, Inadequate Procedure AOP-036.04 for Fire Area 1-A-BAL-C Post-Fire Safe Shutdown.

Minor Performance Deficiencies

Inspection Manual Chapter (IMC) 0612 provides guidance and defines the content and format for inspection reports of power reactor inspections. Section 0612-11of the IMC states in part that minor violations are not routinely documented in inspection reports.

However, as stated in the NRC Enforcement Policy and the NRC Enforcement Manual, there may be exceptions. An issue of concern, regardless of whether it involves a violation of requirements, may be documented if related directly to an issue of agency-wide concern. Because this report documents the first NFPA 805 TFPI, issues of concern and minor violations are being presented for the benefit of all stakeholders.

In addition to the above Green NCV, the team identified three examples of minor PDs that were not subject to enforcement action in accordance with the NRC's Enforcement Policy. The licensee initiated NCRs to address these issues in the corrective action program.

Fire response procedure AOP 36.05, Section 3.2, Fire Area 1-A-CSRB, Step 34, credited instrument numbers for the wrong train of equipment. Specifically, the step to monitor residual heat removal (RHR) cool down locally at temperature indicator TI-5551B should have stated to monitor RHR cool down locally at TI-5551A. The licensee initiated NCR 486083 for this issue. This PD was determined to be minor because the procedure error was quickly recognized by the operator assisting with the procedure walkdown and the correct action was demonstrated during the walkdown.

Fire response procedure AOP-36.08, Section 3.2, Fire Area 1-A-SWGRB, Step 9.b and Step 29 RNO step b. have incorrect Train designations. Specifically, the second bullet of Step 9.b referenced the Train A charging safety injection pump (CSIP) where it should have been the Train B CSIP. Step 29, RNO step b.

referenced switchgear room A (SWGRA) to de-energize Train B equipment where it should have referenced SWGRB. This PD was determined to be minor because the procedure error was quickly recognized by the operator assisting with the walkdown and the correct action was demonstrated during the walkdown. The licensee initiated NCR 489589 for this issue.

While performing a walkdown of the selected FAs to verify that the licensee had identified all credible ignition sources the team found several ignition sources that were not accounted for in the licensees ignition source calculation. Three of the ignition sources were associated with ECs and were being tracked to be added to the calculation. However, one ignition source was not accounted for in the calculation. Missing an ignition source was a PD. However, this PD was considered minor because there were no target structures, systems, and components (SSCs) within the ignition source zone of influence. The licensee initiated NCR 486087 to enter this issue into the corrective action program (CAP). In addition, the team noted that bus ducts located in Switchgear Room B (as well as in other fire areas) were not counted as credible ignition sources.

This issue was also identified by the NRC reviewers during their review of the licensees NFPA 805 license amendment request. Progress Energy performed a sensitivity analysis to determine the risk impact of the bus ducts as ignition sources. As detailed in the HNP NFPA 805 safety evaluation report dated June 28, 2010, the result of the sensitivity analysis, which was reviewed by the NRC staff reviewers, was below the threshold for acceptance. The licensee is tracking this item as NCR 489655 to be incorporated into the base case at the next probabilistic risk assessment (PRA) update.

Observation

During the walkdown of FA 1-BAL-C, the team found one non-safety MCC with doors that could not be closed because bolts were missing. This was considered to be very important because one of the key assumptions in the licensees fire PRA was that the MCCs were sealed and fires would not propagate outside of the MCCs cubicles. The licensee initiated work order 502904 to fix the doors on the MCC. Furthermore, the NRC reviewers questioned this assumption during the NFPA 805 licensing process. Progress Energy performed a sensitivity analysis to determine the impact of 10% of MCC fires propagating outside of the cubicles. This sensitivity analysis, which was reviewed by the NRC staff reviewers, showed that the increase in risk was below the threshold for acceptance, resolving the concern. After discussions with the inspection team, the licensee agreed to include the results of the sensitivity analysis at the next fire PRA scheduled update. The licensee initiated NCR 489655.

.02 Passive Fire Protection

a. Inspection Scope

The team conducted walkdown inspections and examined the material condition and as-built configuration of accessible passive fire barriers surrounding and within the FAs selected for review to evaluate the adequacy of the fire resistance in accordance with the associated NFPA 805 FSA calculations. Fire barriers in use included block walls, poured walls, ceilings, floors, mechanical and electrical penetration seals, doors, dampers, electrical raceway fire barrier systems (ERFBS) and safety system fire resistive cable. The team compared the as-built installed barrier configurations to the approved construction details, and supporting fire endurance test data, which established the ratings of the fire barriers. For example, fire doors and hatches were examined for attributes such as material condition, tightness, proper operation, Underwriters Laboratories (UL) label on door, frame, and latch, method of attachment to the wall, etc. Construction detail drawings were reviewed as necessary. Other types of fire barriers were inspected in a similar manner. The team verified that the as-built configurations met the engineering design, standard industry practices, and were properly evaluated or qualified by appropriate fire endurance tests. The team also reviewed licensing bases documentation, such as 10 CFR 50.48(a), 10 CFR 50.48(c),the NRC NFPA 805 Safety Evaluation Report (SER), and a sample of licensee change evaluations for variances from deterministic requirements (VFDRs) of non-standard fire barrier configurations in FAs 1-A-SWGRB, 1-A-CSRB, and 1-A-BAL-C, to verify that passive fire protection features met license commitments. In addition, a sample of completed surveillance and maintenance procedures for selected fire doors, fire dampers, and ERFBS were reviewed to ensure that these passive fire barriers were being properly inspected and maintained. The passive fire barriers included in the review are listed in the Attachment.

b. Findings

No findings were identified.

.03 Active Fire Protection

a. Inspection Scope

The team reviewed the licensees conventional fire detection systems, automatic fire suppression systems, and manual fire fighting fire hose and standpipe systems protecting the FAs selected for review.

The team reviewed the adequacy of the design, installation, and operation of the conventional automatic detection and alarm system to actuate in the early stage of a fire.

The review included walkdowns of the as-built configurations and an examination of the types of detectors, detector spacing, the licensee=s technical evaluation of the detector locations, and the ceiling steel beam reinforcing plans to assess whether the areas were protected by fire detectors in accordance with the Code of Record requirements (NFPA 72E, 1978 Edition). The inspectors also reviewed the licensees alarm response procedures, FSA calculations, NFPA 805 License Amendment Request (LAR) submittals and associated NRC NFPA 805 SER to verify that the conventional fire detection and alarm systems for the selected FAs were installed in accordance with the design and licensing bases of the plant.

The team inspected the material condition, operational configuration, design, and testing of the sprinkler system in the cable spreading room (FA 1-A-CSRB). The inspection consisted of reviewing the system layout drawings against the field installation. During field walkdowns, piping configurations and the locations of sprinkler heads were observed to check for proper positions and obstructions. The team also reviewed the HNP NFPA 13 Code Compliance Evaluation to determine if there were any outstanding code deviations for the selected area. Fire Areas 1-A-SWGRB, 1-A-BAL-C and 1-A-BAL-G did not have fixed fire suppression systems.

The team also reviewed fire fighting pre-fire plans and fire response procedures for the selected FAs to determine if appropriate information was provided to fire brigade members to facilitate suppression of an exposure fire that could impact NSCA strategies.

Additionally, the inspectors evaluated whether the fire response procedures and pre-plans for the selected FAs contained provisions for containment and monitoring of smoke and fire suppression agent runoff to minimize radioactive releases generated from the combustion of radioactive materials or from fire suppression activities. The team walked down the selected FAs to compare the associated pre-fire plans and drawings with as-built plant conditions and fire response procedures. This was done to verify that fire fighting pre-fire plan instructions and drawings were consistent with the fire protection features and potential fire conditions described in the NFPA 805 FSA calculations to demonstrate the achievement of the nuclear safety and radioactive release performance criteria of NFPA 805. The inspectors also evaluated whether the fire response procedures and pre-fire plans for the selected FAs could be implemented as intended. Additionally, fire brigade response time-to-drill scenarios that transpired since 2008 were reviewed to verify that fire brigade response time performance basis monitoring criteria were met. During plant tours, the team observed placement of the fire hoses and extinguishers to verify they were not blocked and were consistent with the fire fighting pre-fire plans and FPP documents. Specific documents reviewed by the team are listed in the Attachment.

b. Findings and Observations

No findings were identified. However, the team observed that the licensee was calculating the probability of non-suppression (PNS) utilizing newly installed incipient detection systems. The risk reduction credited for use of this method of detection differed from the guidance in NUREG/CR-6850, Supplement 1. This issue was also noted by the NRC reviewers during their review of the licensees NFPA 805 license amendment request. This guidance was not available at the time that HNP staff was evaluating PNS for fire scenario incorporating incipient detection. This guidance was still in the frequently asked questions (FAQ) process and was not finalized when HNP submitted their license amendment request. However, HNP staff performed a sensitivity analysis to determine the impact the incipient detection PNS guidance in NUREG/CR-6850, Supplement 1. The result of the sensitivity was below the threshold for acceptance. Therefore, the NRC reviewers determined that the licensees approach was acceptable for transition. However, the licensee is required, as part of their license amendment, to utilize the revised methodology contained in NUREG/CR-6850, Supplement 1 prior to making changes (unless the change has no more than a minimal risk impact) to FAs with incipient detection.

.04 Protection From Damage From Fire Suppression Activities

a. Inspection Scope

The team walked down the selected FAs to evaluate whether the ability to achieve the nuclear safety performance criteria could be adversely affected due to damage from fire suppression activities or from the rupture, or inadvertent operation of, fire suppression systems. The team addressed the possibility that a fire in one FA could lead to the migration of smoke or hot gases to other plant areas. Airflow paths out of the selected FAs identified on heating ventilation and air conditioning drawings were reviewed to verify that inter-area migration of smoke or hot gases would not inhibit necessary local operator recovery actions required for the selected FAs. The team also evaluated whether the manual fire fighting activities could adversely affect the credited nuclear safety equipment and/or adversely affect the local operator recovery actions required for the selected FAs. Additionally, the team checked that fire fighting water would either be contained in the fire affected area or be safely drained off. This portion of the inspection was carried out through a combination of walkdowns, drawings, and records review.

Documents reviewed by the team are listed in the Attachment.

b. Findings

No findings were identified.

.05 Shutdown from Primary Control Station

a. Inspection Scope

Methodology

For postulated fire scenarios in FAs 12-A-CR, 12-A-CRC1, and 12-A-HVIR, which may impair MCR functions, the licensee credited shutdown from primary control stations to achieve safe and stable plant conditions. This would involve transferring plant controls from the MCR to the auxiliary control panel (ACP). Electrical schematics were reviewed to verify that circuits for SSD equipment, which could be damaged due to fire, were isolated by disconnect switches and by swapping power supplies for selected MCCs.

The team reviewed the ACP transfer switch testing methodology to assess the functionality of the isolation feature of the transfer switches. The team also reviewed the licensees FPP as described in UFSAR Section 9.5.1, the NSCA, DSPs, piping and instrumentation drawings, electrical drawings, and other supporting documents. The reviews focused on ensuring that the required functions for post-fire SSD and the corresponding equipment necessary to perform those functions were included in the fire response AOPs. The review included assessing whether safe and stable plant conditions from the primary control stations outside the MCR could be implemented and that transfer of control from the MCR to the ACP could be accomplished in accordance with procedure AOP-004, Remote Shutdown. This review also included verification that safe and stable conditions could be achieved and maintained from the primary control station both with and without the availability of offsite power.

b. Findings and Observations

No findings were identified. However, the team observed that the licensee did not ensure that all circuits had been electrically isolated from the ACP.

The licensee performed functional transfer verification to ensure the equipment could be controlled from the ACP, and verified control of the equipment was removed from the MCR by attempting to operate the equipment from the MCR. The inability to control equipment from the MCR provided indication that portions of the MCR circuit were isolated, but did not ensure all circuits had been electrically isolated from the ACP.

The licensee demonstrated with a bounding analysis that the risk associated with not ensuring complete isolation of these circuits would be acceptably low. Therefore, no additional testing was required to ensure that all circuits would have been isolated from the MCR.

.06 Circuit Analyses

a. Inspection Scope

The team reviewed the NSCA, licensee circuit analysis documents, flow diagrams, and electrical schematics to verify that the licensee properly identified circuits that could impact the ability to achieve and maintain safe and stable conditions. The team verified, on a sample basis, that the licensee properly identified cables and equipment required to achieve and maintain safe and stable conditions, i.e. hot standby, for the selected fire scenarios in the selected fire areas. The team verified that cables associated with safe shutdown-related equipment were protected from the adverse effects of fire damage or were analyzed to show that fire induced cable faults would not prevent shutdown to safe and stable conditions.

The team reviewed flow diagrams for safe shutdown systems to assess the licensees review of potential flow diversions or maloperations that may impact nuclear safety performance criteria.

The team also reviewed, on a sample basis, coordination study documents for the 480V bus 1B1 and the 480V MCC 1B24 to ensure proper coordination existed between load and incoming supply breakers. The team also verified adequate margin was available for the Alternate Seal Injection diesel generator to supply the required loading.

For the sample of components selected, the team reviewed piping and instrumentation drawings, electrical schematics, and wiring diagrams to identify power, control, and instrumentation cables necessary to support safe shutdown equipment operation. In addition, the team reviewed cable routing information to verify that fire protection features were in place to satisfy the requirements specified in the fire protection license basis. Specific components reviewed by the team are listed in the Attachment.

b. Findings

No findings were identified.

.07 Communications

a. Inspection Scope

The team walked down sections of the emergency control station shutdown procedures to verify that portable radio communications and fixed emergency communication systems remained available, operable, and adequate for the performance of designated activities. The team verified, through inspection, the contents of designated emergency storage lockers. Additionally, the team verified the capability of the communication systems to support the operators in the conduct and coordination of their required actions. The team discussed system design, testing, and maintenance with engineering personnel.

The team reviewed the adequacy of the communication system to support plant personnel in the performance of alternative post-fire safe-shutdown functions and fire brigade duties. Selected fire brigade drill evaluation/critique reports were reviewed to assess proper operation and effectiveness of the fire brigade command post portable radio communications during fire drills and to identify any history of operational or performance problems with radio communications during fire drills. In addition, the team verified the radio battery usage ratings for the fire brigade radios stored and maintained on charging stations.

b. Findings

No findings were identified.

u.08 Emergency Lighting

a. Inspection Scope

The team reviewed the adequacy of the emergency lighting units (ELUs) used to support plant personnel during post-fire safe shutdown for the selected FAs. The team performed plant walkdowns and observed the placement and coverage area of fixed 8-hour battery pack emergency lights throughout the selected FAs to evaluate their adequacy for illuminating access and egress pathways and any equipment requiring local operation and/or instrumentation monitoring for post-fire SSD.

a. Inspection Scope

One of the requirements in NFPA 805 is for licensees to implement FPP controls during non-power operational modes. Although HNP did not enter an outage during the inspection period, the team reviewed plant calculations, procedures, and analyses that defined the key safety functions required to maintain the plant in a safe and stable condition during non-power operational modes. The team also verified that where the licensee had identified specific areas or pinch points where one or more key safety functions could be lost, additional actions had been established which would be taken during high-risk evolutions in the locations of the pinch points where key safety functions could be lost.

b. Findings and Observations

No findings were identified. However, the team identified the following minor PD not subject to enforcement action in accordance with the NRC's Enforcement Policy:

The team reviewed several procedures that were revised to incorporate actions in response to the licensees non-power operations evaluation. AOP-20, AOP-36.04, AOP-36.05, and AOP-36.08 all have a common step to locally control RHR temperature and flow if indication is lost in the MCR. The step directs the operator to control RHR flow rate between 2000 and 3750 gallons per minute (gpm) for the applicable Train while maintaining desired RCS temperature. The note just before the step states that RHR pump flow can be monitored locally at flow indicators FIS-602A and FIS-602B. While performing a walkdown of the procedure steps, the team found that these flow indicators were not sufficient for ensuring that the desired RHR flow could be maintained in that these FIS-602A and FIS-602B flow indicators had a range of 0 - 1500gpm. The failure to direct the operator to an instrument with the appropriate range is a PD. This PD was determined to be minor because the underlying intent of the step was to maintain the desired RCS temperature, which could be adequately monitored locally. The licensee initiated NCRs 489107 and 489387 to resolve this issue and implemented hourly fire watches as compensatory measures in accordance with the FPP.

.14 Monitoring Program

1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

a. Inspection Scope

The team reviewed the plant modifications, as described in the Shearon Harris NFPA 805 Safety Evaluation, dated June 28, 2010, to verify that they were completed in order for HNP to finalize their transition to NFPA 805. The following modifications were listed in the Safety Evaluation and reviewed by the team:

Engineering Change(EC) #62343, Fire Protection for Steam Generator PORV-C Cables for Auxiliary Control Panel Operation (ACP) - NFPA 805

EC# 62820, Upgrade Reliability of Operator SSD Communications at the ACP for a Postulated Fire in the MCR

EC# 68645, Protect 1AF-74 From Fire Damage for a Postulated Fire in Fire Zone 1-A-SWGRB ( B Switchgear Room)

EC# 68646, Add Thermal Shields over Bus 1B-SB, 1B-NNS, 1E-NNS, 1B1-NNS in SWGR room 1B

EC# 68648, Protect Cable 10988B-SR4 to Prevent Spurious AFW Isolation to Steam Generators

EC# 68658, Prevent Spurious Opening of Either 1CC-147 or 1CC-167 During a Postulated Fire in 12-A-CR

EC# 68769, HNP Evaluation/Modification of SSD MOVs Subject to IN 92-18

EC# 69501, Design and Installation of Incipient Fire Detection For NFPA 805

EC# 69764, Upgrade Existing Credited Hemyc Wrap Applications

EC# 69765, Upgrade Existing Credited MT Wrap Applications

EC# 70027, Add 480 VAC Power Outlets to Supply Compensatory Fans for Cooling at Charging Safety Injection Pump (CSIP) and B RHR Pump Rooms

EC# 70350, Alternate Seal Injection and Back-Up Diesel Generator System

EC#70895, Protect TD AFW MOVs Free from Fire Damage in 12-A-CRC-1

EC# 71147, Prevent Multiple Spurious Operation Due to Fire in MCC Vertical Stacks-NFPA 805

The team reviewed procedures, engineering calculations, modification design and implementation packages, work orders, site drawings, corrective action documents, and supporting analysis. The team review verified that changes resulting from the modifications were adequately incorporated in licensing and design basis documents and associated plant procedures. The team performed plant walkdowns of the modifications and selected aspects were discussed with engineering personnel.

b. Findings and Observations

No findings of significance were identified. However, the team identified the following examples of minor PDs not subject to enforcement action in accordance with the NRC's Enforcement Policy:

The team reviewed training and qualification for instrumentation and controls technicians that performed weekly and quarterly maintenance on the incipient fire detection system. As described in NFPA Standard 72 (2007 edition) Chapter 10.2.2.5, maintenance and servicing personnel of the fire detection system shall be factory trained and certified to service the specific equipment in the building.

The team found that several of the technicians did not have a valid certificate showing they were qualified to maintain the system. Upon further investigation by the licensee, the team was informed the employees were factory trained but did not receive the required certification from the vendor. The team confirmed that the licensee contacted the vendor and the vendor would send the certifications. The failure to have adequate documentation that technicians were certified to service the incipient detection system was a PD. The licensee entered the issue into the corrective action program under NCR 489038, and maintenance on the system was re-performed by employees with valid certifications.

Modification Package EC 6974, which was issued to upgrade Hemyc fire wrap installations in the plant, specified that a termination collar was to be installed on the Hemyc wrap for specific conduits at the walls separating different FAs. While performing a walkdown inspection, the team noted that the Hemyc termination collar had not been installed for conduit 10310A between FAs 1-A-CSRB and 1-A-CSRA. The licensee reviewed the nuclear capability assessment report and the fire PRA and determined that the Hemyc wrap for this conduit was not credited. The team determined that the absence of the collar was properly evaluated in calculation HNP-M/MECH-1103, Hemyc Electrical Raceway Fire Barrier System Rating Calculation; however, the EC had not been revised to reflect this fact. NCR 485422 was initiated for this issue.

OTHER ACTIVITIES

4OA2 Identification and Resolution of Problems

a. Inspection Scope

The team reviewed recent independent licensee self-assessments for thoroughness, completeness, and conformance to requirements. The team reviewed other corrective action program documents, including completed corrective actions documented in selected NCRs, and operating experience program documents to verify that industry-identified fire protection problems potentially or actually affecting the plant were appropriately entered into and resolved by the licensees CAP. The team included NRC Regulatory Issue Summaries, Information Notices, and vendor information letters in their review of operating experience program effectiveness. The team reviewed the timeliness, proposed corrective action, and other attributes associated with a sample of NCRs. The NCRs reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA5 Other Activities

.01 (Closed) Licensee Event Report (LER) 05000400/2002-004-09: Unanalyzed Condition

Due to Inadequate Separation of Associated Circuits

(Closed) Licensee Event Report (LER) 05000400/2004-004-00: Unanalyzed Condition Due to Inadequate Separation of Associated Circuits

(Closed) Unresolved Item (URI)05000400/2005007-01, Fire Response Procedures May Not Be Adequate To Prevent RCP Seal Failure and Subsequent Seal Loss of Coolant Accident for a Fire in Certain Fire Areas

a. Inspection Scope

Starting in December 2002, and continuing over the course of approximately 2 years, the licensee submitted a series of LERs documenting the discovery of unanalyzed conditions related to their FPP. The conditions discovered included failure to protect redundant trains of SSD equipment, failure to ensure the capability for alternative shutdown for fires in rooms where SSD could not be achieved from the MCR, and failure to provide adequate procedural guidance to ensure SSD. Additionally, during a 2005 TFPI, inspectors discovered another example of inadequate procedural guidance for post-fire SSD and opened a separate URI for the issue.

In 2005, the licensee committed to transition their HNP fire protection licensing basis to comply with 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805. During the transition period, these LERs and URI remained open while inspectors monitored the transition process. The licensee has completed the transition of its FPP to meet the requirements stated in 10 CFR 50.48(c). The commissions approval of this licensing basis change is documented in a safety evaluation report dated June 28, 2010.

In closing these open items, which describe noncompliances with the licensees previous FPP licensing basis, the inspectors performed a detailed review of information contained in, and associated with the LERs and URI. The inspectors also assessed the adequacy of the licensees corrective actions.

b. Findings

Problem Associated with the Failure to Implement Requirements and Standards of the Fire Protection Program

Introduction.

The inspectors determined that the two LERs (licensee identified) and one URI (NRC identified) involved multiple, closely related issues of concern associated with implementation of the FPP as described in detail below. In accordance with Section 2.13.8 of the NRC Enforcement Manual, the NRC evaluated these issues and determined they would be treated as a group of related programmatic noncompliances associated with the licensees implementation of the requirements and standards of Harris Operating License Condition 2.F, Fire Protection Program.

The licensee has completed the transition of its fire protection program, to meet the requirements stated in 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805. The commissions approval of this licensing basis change is documented in a safety evaluation report dated June 28, 2010. As part of this transition, the licensee completed several modifications, listed in Section 1R17 of this report, that have resulted in a significant decrease in risk. During this TFPI, inspectors reviewed these and other modifications, and concluded that the transition to NFPA 805 adequately addressed the noncompliances, associated with the previous FPP, described below.

Issues associated with the problem

1. Noncompliance for Failure to Protect Redundant Trains of Safe Shutdown Equipment

Introduction.

The licensee identified a noncompliance with Harris Operating License Condition 2.F, Fire Protection Program, for the failure to protect redundant trains of SSCs, located in the same FA, needed to achieve post-fire SSD. This noncompliance affected redundant SSD trains located in multiple FAs. Specifically, the licensee failed to protect the SSCs by one of the methods specified in Fire Protection Evaluation and Comparison to NUREG-0800, BTP, CMEB 9.5-1, Section C.5.b (2), which was incorporated, by reference, into the licensees FPP.

Description.

On February 18, 2003, the licensee submitted LER 2002-004-00, which described the lack of separation of redundant trains of SSCs credited by the SSA for achieving post-fire SSD. Over the course of approximately 2 years, the licensee submitted multiple revisions to the original LER, detailing additional instances of SSCs, credited by the SSA, that were not protected by one of the methods specified in BTP CMEB 9.5-1 Section C.5.b(2). These protection methods included the use of spatial separation, passive fire barriers, fire detection, and automatic fire suppression. The licensee also submitted LER 2004-004-00, which details an additional example of inadequate separation of associated circuits. The following FAs were affected:

1-A-BAL-A - Reactor Auxiliary Building

1-A-BAL-B - Reactor Auxiliary Building

1-A-BAL-C - Reactor Auxiliary Building

12-A-BAL - Reactor Auxiliary Building Balance

1-A-EPA - Electrical Penetration Area A

1-A-EPB - Electrical Penetration Area B

1-A-CSRA - Cable Spreading Room A

1-A-CSRB - Cable Spreading Room B

1-A-SWGRA - Switchgear Room A

1-A-SWGRB - Switchgear Room B

1-A-ACP - Auxiliary Control Panel Room

1-C - Containment Building

The licensee determined that fires in certain areas could potentially lead to conditions that adversely impact the ability to achieve and maintain SSD, due to the lack of separation of components and associated circuits. These conditions include loss of charging, loss of reactor coolant pump (RCP) seal cooling, and loss of auxiliary feedwater (AFW) flow to steam generators (SGs). The licensee determined that these conditions were caused by an inadequate original SSA and failure to issue a drawing change during the plants construction phase that would have specified the use of fire barrier material to protect certain cables. Upon discovery, the licensee implemented compensatory measures, including de-energizing certain valves to prevent spurious mal-operation of components, and posting roving fire watches in FAs of concern.

2. Noncompliance for Failure to Provide Alternative Shutdown Capability for Certain Fire

Areas

Introduction.

The licensee identified a noncompliance with Harris Operating License Condition 2.F, Fire Protection Program, for the failure to provide alternative shutdown capability for fires in certain areas where protection of SSCs did not satisfy the guidelines of Fire Protection Evaluation and Comparison to NUREG-0800, BTP, CMEB 9.5-1, Sections C.5.b(1) and C.5.b(2). This document was incorporated, by reference, into the licensees FPP. Specifically, the licensee failed to consider the possibility of occurrence of certain fire-induced hot shorts that could adversely impact the ability to achieve and maintain SSD.

Description.

On February 18, 2003, the licensee submitted LER 2002-004-00, which described the lack of separation of redundant trains of SSCs credited by the SSA for achieving post-fire SSD. On February 13, 2004, the licensee discovered additional FAs affected by this condition and subsequently issued a revision to the LER. These areas included FA 12-A-CR (Main Control Room) & FA 12-A-CRC1 (Control Room Complex).

Per the licensees analyses, SSD for fires in these areas would be achieved by alternative shutdown methods.

The licensee discovered that cables for two volume control tank (VCT) outlet motor operated valves (MOVs) (1CS-165 & 1CS-166) were not protected from fire-induced spurious actuation. These valves were required to remain open to provide CSIP suction from the VCT for a postulated fire in the areas affected. A fire in FA 12-A-CR or FA 12-A-CRC1 could result in spurious closure of one of the VCT outlet valves. Closure of one of these valves would cause a loss of suction flow to all 3 CSIPs. This scenario could result in subsequent damage to the running CSIP, and unavailability of charging flow and RCP seal cooling, credited in the SSA for achieving and maintaining SSD.

Additionally, the licensee discovered that cables for two containment spray MOVs (1CT-102 & 1CT-105) were not protected from fire-induced spurious actuation. These valves are required to remain closed to prevent transfer of inventory from the refueling water storage tank (RWST) to the containment recirculation sump. The RWST is the suction source credited by the SSA to provide makeup inventory to the RCS. A fire in FA 12-A-CR or FA 12-A-CRC1 could result in inadvertently transferring inventory from the RWST to the containment recirculation sump, making the credited suction source unavailable to provide inventory to the RCS. The licensee determined that these conditions were caused by an inadequate original SSA. Upon discovery, the licensee implemented compensatory measures, including posting roving fire watches in FAs of concern, and committed to restoring compliance by implementing design changes.

3. Noncompliance for Providing Inadequate Procedural Guidance for Post-Fire Safe

Shutdown

Introduction.

The inspectors identified a noncompliance with 10 CFR 50 Appendix B, Criterion V for the licensees failure to provide adequate procedural guidance in post-fire SSD procedure AOP-036, Safe Shutdown Following a Fire. Specifically, AOP-036 did not include guidance from Westinghouse regarding time limitations for restoring cooling to the RCP seals. Additionally, AOP-036 also directed operators to perform actions that may not have been feasible, due to the presence of smoke, under certain conditions.

Description.

In 2005, inspectors identified a condition where the SSA credits a potentially non-feasible operator manual action, which could adversely affect the ability to achieve and maintain SSD. This issue was documented in TFPI inspection report 05000400/2005007, and was identified as URI 2005-007-01. Procedure AOP-036, Safe Shutdown Following a Fire, would have been used by operators to safely shutdown the unit from the MCR in the event of a significant fire in certain FAs. Procedure AOP-036 included local operator actions to restore component cooling water (CCW) to the RCP thermal barrier cooler if fires in certain areas resulted in the loss of all cooling to the RCP seals. Inspectors noted that the licensee had identified eight FAs where RCP seal cooling could be lost due to a significant fire in the applicable FA. Westinghouse technical bulletin TB-04-22, dated 11/17/2004 and 8/9/2005, stated that cooling to the RCP seals should be restored within 13 minutes in the event of a loss of all RCP seal cooling. The Westinghouse guidance further stated that RCP seal cooling should not be restored if restoration would take more than 13 minutes. Restoration of seal cooling after 13 minutes could result in cold thermal shock to the RCP seals, causing increased seal leakage leading to a seal loss-of-coolant accident (LOCA), or water hammer that could compromise the integrity of the CCW system.

Inspectors noted that AOP-036 did not include procedural guidance to alert operators about the Westinghouse time constraints for restoring cooling to the RCP seals contained in the Westinghouse technical manual. The licensee performed a Manual Action Feasibility Study to validate the operator manual actions incorporated into the SSD AOPs. Based on this study, the licensee found that, for a fire in FA 1-A-BAL-B (B2), the time to restore RCP seal cooling was 17 minutes and 39 seconds. Similarly, the study found that, for a fire in FA 1-A-CSRB, the time to restore RCP seal cooling was 18 minutes and 30 seconds.

Additionally, on September 20, 2005, the licensee submitted a revision to LER 2002-004-00, which described the discovery of a condition where the SSA credits a potentially non-feasible operator manual action. For a fire in FA 1-A-BAL-A (236 elevation) that disables control of valve 1CS-291 from the MCR due to spurious actuation, AOP-036 directs operators to manually open the valve. Valve 1CS-291 provides suction from the RWST to the CSIP as well as support for normal charging operation for RCS inventory control. Access to manually operate valve 1CS-291 may not be feasible due to the presence of postulated smoke, under certain conditions. Not performing this operator manual action, when needed, could have adversely impacted the ability to achieve and maintain SSD.

The licensee determined that these conditions were caused by an inadequate original SSA. The licensee also determined that the appropriate interim compensatory measure, in accordance with their FPP, was to post roving fire watches in the affected areas.

These fire watches had already been previously posted, as described above.

Analysis of the Problem

Failure to adequately implement the requirements contained in Harris License Condition 2.F, Fire Protection Program, was a PD. This finding is more than minor because it is associated with the Reactor Safety Mitigating System Cornerstone attribute of protection against external events (i.e. fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

The inspectors considered the use of IMC 0609, Appendix F, "Fire Protection Significance Determination Process" to evaluate the significance of this problem.

Some assessment utilizing Appendix F was initially conducted; however, because detailed modeling for each applicable fire scenario would be exceedingly difficult and time consuming, the decision was made to utilize IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria in order to provide reasonable estimates of the finding significance within the established SDP timeliness goal of 90 days. Additionally, a limited bounding analysis was conducted to aid in determining the risk of this problem. The inspectors utilized the insights gained from the limited bounding analysis and the following qualitative factors to determine the risk of this problem:

The limited bounding analysis grouped several PDs (regardless of root causes)by fire areas to estimate the risk. Therefore, this bounding analysis did not constitute an SDP for specific PDs. Per IMC 0609, Appendix A, if the concurrent multiple equipment or functional degradations resulted from a common cause, then a single inspection finding is written and assessed.

The limited bounding analysis did not take into account the fact that the performance deficiencies accounted for did not exist at the same time. The limited bounding analysis assumed that the performance deficiencies existed at the same time, rather than determining individual exposure periods as would normally be done per IMC 0609, Appendix A. Several performance deficiencies were corrected by the licensee throughout the years.

Considering the qualitative and quantitative factors described above, the significance of this finding was determined to be less than Red. The team determined that this finding did not present an immediate safety concern because in transitioning its FPP licensing basis to 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, the licensee addressed all noncompliances associated with this Problem.

The inspectors determined that these noncompliances did not have a crosscutting aspect because they did not represent current licensee performance.

Enforcement of the Problem

Shearon Harris License Condition 2.F, Fire Protection Program states, in part, that Carolina Power & Light Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report (FSAR) for the facility as amended.

Section 9.5.1, Fire Protection System, of the FSAR incorporates, by reference, Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1, Revision 3, dated May 7, 1986.

o Section C.5.b

(2) of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1 requires one train of systems necessary to achieve and maintain hot standby conditions from either the control room or emergency control station(s) be free of fire damage by providing one of the means described in Section C.5.b (2) (i.e., use of spatial separation, passive fire barriers, and fire detection and an automatic fire suppression system).

o Section C.5.b (3), of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1 requires that alternative or dedicated shutdown capability be provided where the guidelines of Section C.5.b

(1) and C.5.b (2)cannot be met.

Section 9.5.1.5.4, Quality Assurance Program, of the FSAR states that the fire protection quality assurance program elements are included in Section 17.3 of the FSAR. Section 17.3.1.1, Methodology, of the FSAR states, in part, that the HNP quality assurance program prescribes measures for the control and accomplishment of activities for the operation of safety related and fire protection SSCs. Section 17.3.1.1 also commits to the requirements of 10 CFR 50, Appendix B. 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances.

Contrary to the above, the licensee failed to meet the requirements of its documented fire protection program, in that:

The licensee failed to protect redundant systems necessary to achieve and/or maintain hot shutdown conditions from the MCR or emergency control station(s)from fire damage by one of the means described in Section C.5.b(2) of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1.

The licensee failed to ensure alternative shutdown capability was available for two fire areas where the guidelines for ensuring one redundant train for safe shutdown remain free of fire damage, detailed in Section C.5.b

(1) and C.5.b (2)of Fire Protection Evaluation and Comparison to NUREG-0800, BTP CMEB 9.5-1 could not be met.

The licensee failed to provide adequate procedural guidance, in that the licensees fire safe shutdown procedure failed to incorporate instructions to alert operators concerning time constraints for restoring cooling to the RCP seals.

Additionally, the licensees fire safe shutdown procedure included steps that were not appropriate to the circumstances in that a required procedural step may not have been feasible due to the presence of postulated smoke, under certain conditions.

Because this issue relates to fire protection, and the associated noncompliances were resolved by compliance with 10 CFR 50.48(c), the NRC is exercising enforcement and reactor oversight process discretion for this issue in accordance with the NRC Enforcement Policy, Section 9.1, Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48) and Inspection Manual Chapter 0305. Specifically, the licensee entered the noncompliances into their corrective action program and implemented appropriate compensatory measures. The noncompliances were not associated with a finding of high safety significance (Red), the noncompliances were not willful, and the licensee submitted a letter of intent stating its intention to transition to 10 CFR 50.48(c)by December 31, 2005.

LER 05000400/2002-004-09, Unanalyzed Condition Due to Inadequate Separation of Associated Circuits; LER 05000400/2004-004-00, Unanalyzed Condition Due to Inadequate Separation of Associated Circuits; and URI 05000400/2005007-01, Fire Response Procedures May Not Be Adequate To Prevent RCP Seal Failure and Subsequent Seal Loss of Coolant Accident For a Fire in Certain Fire Areas, are closed.

4OA6 Meetings, Including Exit

On September 23, 2011, the lead inspector presented the inspection results to W.

Jefferson and other members of his staff. The licensee acknowledged the findings.

Proprietary information is not included in this report. On December 8, 2011, the lead inspector conducted a final exit meeting with Mr. Corlett and other members of the licensees staff.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

J. MacIntyre, Fire protection Program Manager

J. DeJoseph, Fire Protection System Engineer

J. Nobles, Fire Protection Coordinator

R. Rhodes, Safe Shutdown Program Manager

R. Dukes, Safe Shutdown Engineer

J. Dorrhy, Sr. Licensing Specialist

D. Corlett, Licensing Supervisor

D. Miskewicz, PSA Principal Engineer

R. Davis, PSA Senior Engineer

J. Ertman, Fleet Fire Protection Supervisor

A. Holder, Sr. Fire Protection Engineer

V. DSouza, Fleet Sr. Fire Protection Engineer

F. Womack, Operations Manager

E. Kapapoulos, Plant Manager

W. Jefferson, Site Vice President

NRC personnel

P. Lessard, Resident Inspector

J. Austin, Senior Resident Inspector

A. Klein, Chief, Fire Protection Branch, NRR

R. Nease, Chief, Engineering Branch 2, Division of Reactor Safety, Region II

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Opened and Closed

05000400/2011011-01 NCV

Inadequate Procedure AOP-036.04 for Fire Area 1-A-BAL-C Post-Fire Safe Shutdown (Section 1R05.01.b)

Closed

05000400/2002-004-09 LER Unanalyzed Condition Due to Inadequate Separation of Associated Circuits (Section 4OA5)

05000400/2004-004-00 LER

Unanalyzed Condition Due to Inadequate Separation of Associated Circuits (Section 4OA5)

05000400/2005007-01 URI

Fire Response Procedures May Not Be Adequate To Prevent RCP Seal Failure and Subsequent Seal Loss of Coolant Accident for a Fire in Certain Fire Areas (Section 4OA5)

Discussed

None

LIST OF DOCUMENTS REVIEWED