IR 05000324/2012004

Errata - Brunswick Steam Electric Plant - NRC Integrated Inspection Report Nos. 05000325-12-004 and 05000324-12-004

ML12325A392
Person / Time
Site:	Brunswick
Issue date:	11/20/2012
From:	Randy Musser NRC/RGN-II/DRP/RPB4
To:	Annacone M Carolina Power & Light Co, Duke Energy Corp
Shared Package
ML12325A266	List: IR 05000324/2012004 ML12312A082
References
IR-12-004
Download: ML12325A392 (14)

Text

UNITED STATES ber 20, 2012

SUBJECT:

ERRATA - BRUNSWICK STEAM ELECTRIC PLANT - NRC INTEGRATED INSPECTION REPORT NOS.: 05000325/2012004 AND 05000324/2012004

Dear Mr. Annacone:

On November 7, 2012, the U.S. Nuclear Regulatory Commission (NRC) issued the subject inspection report for Brunswick Steam Electric Plant, ADAMS accession ML12312A082. In reviewing this report it was noted that in section 4OA2.2 the Inspection Scope was duplicated and incorrectly indicated that no findings were identified. Accordingly, we have revised pages 18-26 of Inspection Report 05000325/2012004 and 05000324/2012004, in order to document the necessary changes. Please replace pages 18-26 of the original report with the pages enclosed.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). An ADAMS package which contains both the original report and this errata is available at ML12325A266. ADAMS is accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html (The Public Electronic Reading Room).

I apologize for any inconvenience this error may have caused. If you have any questions, please contact me at (404) 997-4603.

Sincerely,

/RA/

Randall A. Musser, Chief Reactor Projects Branch 4 Division of Reactor Projects Docket Nos.: 50-325, 50-324 License Nos.: DPR-71, DPR-62 Enclosure: Errata Inspection Report 05000325, 324/2012004 Pages 18-26 cc w/encl: (See page 2)

_ML12325A392________________ G SUNSI REVIEW COMPLETE G FORM 665 ATTACHED OFFICE RII:DRP RII:DRP SIGNATURE /RA/ /RA/

NAME JWorosilo RMusser DATE 11/ /2012 11/ /2012 11/ /2012 11/ /2012 11/ /2012 11/ /2012 11/ /2012 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO cc w/encl: Randy C. Ivey Plant General Manager Manager, Nuclear Oversight Brunswick Steam Electric Plant Brunswick Steam Electric Plant Progress Energy Progress Energy Carolinas, Inc.

Electronic Mail Distribution Electronic Mail Distribution Edward L. Wills, Jr. Paul E. Dubrouillet Director Site Operations Manager, Training Brunswick Steam Electric Plant Brunswick Steam Electric Plant Electronic Mail Distribution Electronic Mail Distribution J. W. (Bill) Pitesa Joseph W. Donahue Senior Vice President Vice President Nuclear Operations Nuclear Oversight Duke Energy Corporation Progress Energy Electronic Mail Distribution Electronic Mail Distribution John A. Krakuszeski Senior Resident Inspector Plant Manager U.S. Nuclear Regulatory Commission Brunswick Steam Electric Plant Brunswick Steam Electric Plant Electronic Mail Distribution U.S. NRC 8470 River Road, SE Lara S. Nichols Southport, NC 28461 Deputy General Counsel Duke Energy Corporation John H. O'Neill, Jr.

Electronic Mail Distribution Shaw, Pittman, Potts & Trowbridge 2300 N. Street, NW M. Christopher Nolan Washington, DC 20037-1128 Director - Regulatory Affairs General Office Peggy Force Duke Energy Corporation Assistant Attorney General Electronic Mail Distribution State of North Carolina P.O. Box 629 Michael Raleigh, NC 27602 Vice President Brunswick Steam Electric Plant Chairman Electronic Mail Distribution North Carolina Utilities Commission Electronic Mail Distribution Annette H. Pope Manager-Organizational Effectiveness Robert P. Gruber Brunswick Steam Electric Plant Executive Director Electronic Mail Distribution Public Staff - NCUC 4326 Mail Service Center Lee Grzeck Raleigh, NC 27699-4326 Regulatory Affairs Manager Brunswick Steam Electric Plant (cc w/encl continued - see page 3)

Progress Energy Carolinas, Inc.

Electronic Mail Distribution cc: w/ encl contd:

Anthony Marzano Director Brunswick County Emergency Services Electronic Mail Distribution Public Service Commission State of South Carolina P.O. Box 11649 Columbia, SC 29211 W. Lee Cox, III Section Chief Radiation Protection Section N.C. Department of Environmental Commerce & Natural Resources Electronic Mail Distribution Warren Lee Emergency Management Director New Hanover County Department of Emergency Management 230 Government Center Drive Suite 115 Wilmington, NC 28403 Letter to Michael from Randall A. Musser dated November 20, 2012.

SUBJECT: ERRATA - BRUNSWICK STEAM ELECTRIC PLANT - NRC INTEGRATED INSPECTION REPORT NOS.: 05000325/2012004 AND 05000324/2012004 Distribution w/encl:

J. Baptist, RII L. Douglas, RII OE Mail RIDSNRRDIRS PUBLIC RidsNrrPMBrunswick Resource

samples, TS requirements, issue reports, and event reports for the period to validate the accuracy of the submittals. In addition to record reviews, the inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample.

• Reactor Coolant System Leakage - Unit 1

• Reactor Coolant System Leakage - Unit 2 The inspectors sampled licensee submittals for the Reactor Coolant System Leakage performance indicator for the period from the third (3rd) quarter 2011 through the second (2nd) quarter 2012. The inspectors reviewed the licensees operator logs, RCS leakage tracking data, issue reports, and event reports for the period to validate the accuracy of the submittals.

b. Findings No findings were identified.

4OA2 Identification and Resolution of Problems (71152 - 2 samples)

.1 Routine Review of Items Entered Into the Corrective Action Program a. Inspection Scope To aid in the identification of repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed frequent screenings of items entered into the licensees corrective action program. The review was accomplished by reviewing daily action request reports.

b. Findings No findings were identified.

.2 Assessments and Observations Selected Issue Follow-up Inspection: UPS-A Failure and Loss of Emergency Response Facility Information System (ERFIS), Plant Process Computer (PPC), Business Network a. Inspection Scope The inspectors selected AR 542704, UPS-A Failure and Loss of ERFIS, PPC, Business Network, for detailed review. This AR identified that a single failure caused the loss of ERFIS and Safety Parameter Display System (SPDS) on both units. The inspectors reviewed the licensees CAP for ERFIS and SPDS failures in the past. The inspectors reviewed these reports to verify that the licensee identified the full extent of the issue, performed an appropriate evaluation, and specified and prioritized appropriate corrective actions. The inspectors evaluated the reports against the requirements of the licensees CAP as delineated in corporate procedure CAP-NGGC-0200, Corrective Action Program, 10 CFR 50.47, and 10 CFR 50 Appendix E.

Enclosure

b. Findings Introduction: A self-revealing Green NCV of 10 CFR 50.54(q)(2) was identified for the licensees failure to properly evaluate or consider the impact to emergency response facilities of design change ESR98-00436 which was implemented in 1999. As a result, a number of temporary losses of ERFIS, Emergency Response Data System (ERDS),

SPDS, and all displays including radiation monitors for the emergency response facilities occurred. Specifically, the licensee failed to ensure that adequate emergency response facilities and equipment were available as required by the Brunswick Nuclear Plant Radiological Emergency Plan, Section 1.3.1.3, revision 80, and 10 CFR 50.47(b)(8).

This issue was captured in the licensees CAP as AR 542704.

Description: In 1999, the licensee implemented design change ESR98-00436 for the power supply to the ERFIS, ERDS, SPDS, and all displays including RMS for the emergency response facilities. The licensee did not properly evaluate or consider the impact to emergency response facilities and equipment prior to implementation of this design change. As a result, the ERFIS, ERDS, and SPDS systems, and all radiation monitoring system (RMS) displays were susceptible to a single point power failure mode.

The implementation of the design change introduced a single point failure mode which did not meet the design requirements specified in their Design Basis Document (DBD 60) sections 3.6.7.2 and 3.6.7.3. Prior to the licensees implementation of design change ESR98-00436 in 1999, this single point vulnerability did not exist as the power supply system had automatic switching capability on loss of one power source. When the design change was implemented, the ERFIS, ERDS, and SPDS systems and RMS displays were degraded as demonstrated by the resulting failures of those systems on multiple occasions including July 17, 2004 and June 12, 2012. Additionally, all displays for those systems were lost in all of the emergency facilities including the radiation monitoring system.

On June 13, 2012, the licensee made an event notification to the NRC Operations Center, 50.72(b)(3)(xiii) Loss of Emergency Assessment Capability, Offsite Response Capability, or Offsite Communications Capability for the emergency response facilities.

The report delineated that at 5:57 p.m. EDT on June 12, 2012, Brunswick Nuclear Plant experienced a fault on the Emergency Response Facility Information System (ERFIS)

uninterruptible power supply (UPS) electrical bus A. This resulted in a loss of site Safety Parameter Display System (SPDS), Emergency Response Data System (ERDS)

and Plant Process Computer (PPC) for both Unit 1 and Unit 2.

During the loss of SPDS, the emergency response capability of that system was lost to the site. During the loss of ERDS, the automatic data transfer feature of that system was lost for transmissions to the NRC, however manual data transfer was still available.

During the loss of the PPC, automatic core thermal power averaging and automatic core thermal limit monitoring was lost. Manual calculations were available for these functions.

Unit 1 SPDS was restored to the Emergency Operations Facility (EOF) at 7:49 p.m. on June 12, 2012. Unit 2 SPDS was restored to the EOF at 8:30 p.m. on June 12, 2012.

The inverter was restored to service on June 17, 2012 at 12:00 noon.

Enclosure

Inspectors determined that the licensee did not properly evaluate or consider the impact to all emergency response facilities and equipment prior to implementation of the ESR98-00436 design change. The inspectors concluded that the ERFIS, ERDS, and SPDS systems required by the Brunswick Nuclear Plant Radiological Emergency Plan were degraded from 1999 when the design change was installed to present.

Compensatory measures were put in place during the June 2012 event to manually obtain and log the required data from the instrumentation in the control room and transmit to the emergency response facilities, and after the June 2012 event, the licensee initiated a design change to restore the power configuration to those systems back to the original design which would remove this failure mechanism.

Analysis: The licensees failure to properly evaluate or consider the impact to emergency response facilities of design change ESR98-00436 which was implemented in 1999 was a performance deficiency. Specifically, the licensee introduced a single point failure mode which did not meet the design requirements specified in their Design Basis Document (DBD 60) sections 3.6.7.2 and 3.6.7.3. This resulted in the licensees failure to ensure that adequate emergency response facilities and equipment were available as delineated in the Updated Final Safety Analysis Report (UFSAR) Section 7.7.1.9, and required by the Brunswick Nuclear Plant Radiological Emergency Plan, Section 1.3.1.3, revision 80, and 10 CFR 50.47(b)(8).

The finding was more than minor because it adversely affected the Emergency Preparedness Cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the Facilities and Equipment attribute was affected during the time when the ERFIS, ERDS, SPDS, and all displays including radiation monitors for the emergency response facilities were degraded, and as a result did not meet 10 CFR 50.47(b)(8) Planning Standard program element, adequate emergency facilities and equipment to support the emergency response are provided and maintained. The finding was assessed for significance in accordance with NRC IMC 0609, Appendix B Emergency Preparedness Significance Determination Process.

Attachment 2 of Appendix B, Failure to Comply Significance Logic is as follows: Failure to comply; Loss of Risk Significant Planning Standard Function (RSPS), No; RSPS Degraded Function, No; Loss of Planning Standard Function, No; the result is a Green finding. The inspectors determined that this resulted in a low safety significance finding (Green). No cross-cutting aspect was assigned to this finding because the performance deficiency occurred more than three years ago and is not reflective of current plant performance.

Enforcement: 10 CFR 50.54(q)(2) requires, in part, a licensee to follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to this part and, for nuclear power reactor licensee, the planning standards of 10 CFR 50.47(b).

The Brunswick Nuclear Plant Radiological Emergency Plan, Section 1.3.1.3, revision 80, states in part that special provisions have been made to assure that ample space and proper equipment are available to effectively respond to a full range of possible emergencies. Contrary to the above, from 1999, when design change ESR98-00436 was installed, until the compensatory measures were put in place in June 2012, the Enclosure

licensee failed to maintain adequate emergency facilities and equipment to support emergency response when the ERFIS, ERDS, SPDS, and all displays including radiation monitors for the emergency response facilities were degraded due to the implementation of the design change. This resulted in failures of those systems on July 17, 2004 and June 12, 2012. The licensee has compensatory measures in place, entered this issue their CAP as AR 542704, and initiated a design change to restore the power configuration back to the original design. Because the licensee entered the issue into its CAP and the finding is of very low safety significance (Green), this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRCs Enforcement Policy: NCV 05000325; 324/2012004-02, Failure to Maintain Reliability and Availability of Emergency Response Equipment for Emergency Response Facilities.

.3 Assessments and Observations Selected Issue Follow-up Inspection: EDG 2 wiring associated with Alternate Safe Shutdown (ASSD) Switch 2-DG-SS-A1 a. Inspection Scope The inspectors performed a detailed review of AR 557897 associated with the wiring for the EDG 2 Alternate Safe Shutdown (ASSD) Switch 2-DG-SS-A1. The issue was discovered during a planned system outage for EDG2 during the week of August 26.

The inspectors verified that the issue was captured completely and accurately in the CAP. The inspectors evaluated the licensees operability determinations and performed walk-downs with licensee staff of applicable fire areas as needed. The inspectors followed the licensees actions to restore the wiring to its proper configuration and also verified the extent of condition inspections for the remaining EDGs 1, 3 and 4 were completed in a timely manner. The inspectors reviewed the licensees reportability evaluation and subsequent 8-hour report made to the NRC in accordance with 10 CFR 50.72(b)(3)(ii)(B). Additional documents reviewed are listed in the Attachment.

b. Findings Introduction: The inspectors opened an unresolved item (URI) for this issue of concern to determine if a performance deficiency existed.

Description: A wiring discrepancy was identified during inspection of the EDG 2 ASSD switch 2-DG-SS-A1. A contact in the circuit was determined to be bypassed that would have the potential to prevent proper isolation of the EDG2 control circuits from the Main Control Room (MCR) during an Appendix R fire event. The inspectors plan to review the licensees cause evaluation for this event and determine if a performance deficiency existed. This issue is being tracked as URI 05000325; 324/2012004-03, EDG2 wiring on ASSD switch.

Enclosure

4OA3 Follow-up of Events (71153 - 2 samples)

.1 Notice of Unusual Event for Fire in the Protected Area a. Inspection Scope For the plant event listed below, the inspectors reviewed plant parameters, reviewed personnel performance, and evaluated performance of mitigating systems. The inspectors communicated the plant events to appropriate regional NRC personnel, and compared the event details with criteria contained in IMC 0309, Reactive Inspection Decision Basis for Reactors, for consideration of potential reactive inspection activities.

As applicable, the inspectors verified that the licensee made appropriate emergency classification assessments and properly reported the event in accordance with 10 CFR 50.72. The inspectors reviewed the licensees follow-up actions related to the events to assure that the licensee implemented appropriate corrective actions commensurate with their safety significance.

• On August 2, 2012, a fire existed in the protected area on the Units 1 and 2 turbine building roof for approximately two hours, meeting the criteria for a Notice of Unusual Event declaration.

b. Findings One licensee identified violation is documented in Section 4OA7 of this report.

.2 (Closed) LER 05000325/2012-004-00, High Pressure Coolant Injection (HPCI)

Inoperable Due to Erratic Governor Operation a. Inspection Scope On May 2, 2012, Unit 1 HPCI was declared inoperable due to erratic governor operation during Surveillance Test 0PT-09.2, HPCI System Operability Test. The erratic governor operation was due to the failure of the Ramp Generator Signal Convertor (RGSC). The licensee determined that the root cause of the RGSC failure was due to a lack of a replacement preventative maintenance (PM) for the RGSC, which had been installed for at least 22 years. The corrective actions included replacing the RGSC and creating a PM task to replace the RGSCs. The licensee documented the root cause evaluation in NCR 534364. The inspectors reviewed the LER, the NCR, and corrective actions to determine whether the station adequately evaluated the condition.

b. Findings One licensee identified violation is documented in Section 4OA7 of this report. This LER is closed.

Enclosure

4OA5 Other Activities

.1 (Discussed) NRC Temporary Instruction (TI) 2515/187, Inspection of Near-Term Task Force Recommendation 2.3 Flooding Walk-downs, and NRC TI 2515/188, Inspection of Near-Term Task Force Recommendation 2.3 Seismic Walk-downs a. Inspection Scope Inspectors accompanied the licensee on a sampling basis, during their flooding and seismic walk-downs, to verify that the licensees walk-down activities were conducted using the methodology endorsed by the NRC. These walk-downs are being performed at all sites in response to a letter from the NRC to licensees, entitled Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, dated March 12, 2012 (ADAMS Accession No.

ML12053A340).

Enclosure 3 of the March 12, 2012, letter requested licensees to perform seismic walk-downs using an NRC-endorsed walk-down methodology. Electric Power Research Institute (EPRI) document 1025286 titled, Seismic Walk-down Guidance, (ADAMS Accession No. ML12188A031) provided the NRC-endorsed methodology for performing seismic walk-downs to verify that plant features, credited in the current licensing basis (CLB) for seismic events, are available, functional, and properly maintained.

Enclosure 4 of the letter requested licensees to perform external flooding walk-downs using an NRC-endorsed walk-down methodology (ADAMS Accession No.

ML12056A050). Nuclear Energy Industry (NEI) document 12-07 titled, Guidelines for Performing Verification Walk-downs of Plant Protection Features, (ADAMS Accession No. ML12173A215) provided the NRC-endorsed methodology for assessing external flood protection and mitigation capabilities to verify that plant features, credited in the CLB for protection and mitigation from external flood events, are available, functional, and properly maintained.

b. Findings Findings or violations associated with the flooding and seismic walk-downs, if any, will be documented in future reports.

.2 (Discussed) Temporary Instruction (TI) 2515/182 - Review of the Implementation of the Industry Initiative to Control Degradation of Underground Piping and Tanks, Phase 1 a. Inspection Scope Leakage from buried and underground pipes has resulted in ground water contamination incidents with associated heightened NRC and public interest. The industry issued a guidance document, Nuclear Energy Institute (NEI) 09-14, Guideline for the Management of Buried Piping Integrity, (ADAMS Accession No. ML1030901420), to describe the goals and required actions (commitments made by the licensee) resulting from this underground piping and tank initiative. On December 31, 2010, NEI issued Enclosure

Revision 1 to NEI 09-14, Guidance for the Management of Underground Piping and Tank Integrity, (ADAMS Accession No. ML110700122), with an expanded scope of components which included underground piping that was not in direct contact with the soil and underground tanks. On November 17, 2011, the NRC issued TI-2515/182, Review of the Industry Initiative to Control Degradation of Underground Piping and Tanks, to gather information related to the industrys implementation of this initiative.

The instructors reviewed the licensees programs for buried pipe and underground piping and tanks in accordance with TI-2515/182 to determine if the program attributes and completion dates identified in Section 3.3 A and 3.3 B of NEI 09-14, Revision 1, were contained in the licensees program and implementing procedures. For the buried pipe and underground piping program attributes, with completion dates that had passed, the inspectors reviewed records to determine if the attribute was in fact complete and to determine if the attribute was accomplished in a manner which reflected good or poor practices in management.

b. Observations The licensees buried piping and underground piping and tanks program was inspected in accordance with paragraphs 03.01.a through 03.01.c of TI-2515/182 and was found to meet all applicable aspects of NEI 09-14 Revision 1, as set forth in Table 1 of the TI.

Based upon the scope of the review described above, Phase I of TI-2515/182 was completed.

c. Findings No findings were identified.

4OA6 Management Meetings Exit Meeting Summary On July 19, 2012, the inspectors presented inspection results of the triennial heat sink inspection to Mr. Michael Annacone and other members of the licensee staff. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

On September 18, 2012, the inspector presented inspection results of the TI-182, Phase 1 of the Underground Piping and Tanks Inspection by conference call to Mr. James Burke, Site Director of Engineering, and other members of the licensee staff. The inspector verified that all proprietary information was returned to the licensee.

On October 11, 2012, the inspectors presented inspection results from the quarterly inspection to M and other members of the licensee staff. The inspectors confirmed that any proprietary information received during the inspection period were properly controlled or returned to licensee staff.

Enclosure

4OA7 Licensee-Identified Violations The following violations of very low significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as NCVs.

• 10 CFR 50.54(q) requires, in part, a licensee authorized to possess and operate a nuclear power reactor shall follow and maintain in effect emergency plans which meet the standards of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4)

requires, in part, a standard emergency classification and action level scheme be used by the licensee. Procedure 0PEP-02.1.1, Emergency Control - Notification of Unusual Event, Alert, Site Area Emergency, and General Emergency, Step 5.7.2 states, that the emergency declaration will be made within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded. Procedure 0PEP-02.1, Initial Emergency Actions, HU2.1, requires the declaration of an Unusual Event when a fire is not extinguished within 15 minutes of control room notification or verification of a control room fire alarm in any Table H-1 or Table H-3 areas. Table H-1 includes the turbine building. Contrary to the above, on August 2, 2012, a Notice of Unusual Event (NOUE) was not classified within 15 minutes of a fire within the protected area not being extinguished within 15 minutes of detection. Specifically, when a fire was reported on the Turbine Building roof to the Control Room and was not extinguished within 15 minutes, conditions were met for classification of EAL HU2.1 in accordance with Procedure 0PEP-02.1; however, the EAL was not classified until approximately eight hours after the fire started. This issue was entered into the licensees CAP as NCR 552984 and the licensee is performing a root cause evaluation. Corrective actions included making a one hour report to the NRC for discovery of a condition that met the EAL classification for an NOUE after the fact. The inspectors determined the finding was associated with an actual event implementation problem, and assessed the significance using IMC 0609, Appendix B, "Emergency Preparedness Significance Determination Process." Using the Emergency Preparedness SDP, Sheet 1, "Failure to Implement (Actual Event) Significance Logic" the inspectors determined the finding was of very low safety significance (Green) because the licensee failed to implement a risk significant planning standard (10 CFR 50.47(b)(4)) during an actual Notice of Unusual Event.

• 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"

requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Licensee procedure ADM-NGGC-0107, Equipment Reliability Process Guideline, steps 9.4.9 and 9.4.10 required component experts and preventive maintenance (PM) optimization to determine if there was a cost effective PM to prevent failure and then to develop the PM model. Contrary to the above, the Unit 1 high pressure coolant injection (HPCI) ramp generator signal converter (RGSC) did not have the appropriate preventive maintenance to prevent failure.

As a result, the Unit 1 high pressure coolant injection (HPCI) system failed the Enclosure

• HPCI System Operability Test performed on April 30, 2012 and was declared inoperable. The licensee entered this issue into the CAP as NCR 534364.

Corrective actions included replacing the RGSC and creating a PM task to replace the RGSCs on a specified frequency. Using IMC 0609, Appendix A,

"Phase 1 Initial Screening and Characterization of Findings," the inspectors determined this finding required a Phase 2 analysis. The Phase 1 screened this Mitigating Systems Cornerstone finding to Phase 2 because the finding represented a loss of HPCI system and/or function. The inspectors, with the assistance of the regional Senior Risk Analyst, performed a Phase 2 analysis using the Saphire 8 Model. 109 hours0.00126 days <br />0.0303 hours <br />1.802249e-4 weeks <br />4.14745e-5 months <br /> of unavailability time was used for the analysis since HPCI was not required during the refueling outage from February 23, 2012 through April 29, 2012. Based on the results of the Phase 2 analysis, the inspectors determined the finding was of very low safety significance (Green).

ATTACHMENT: SUPPLEMENTAL INFORMATION Enclosure