IR 05000335/2009006: Difference between revisions

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{{Adams|number = ML100210081}}
{{Adams
| number = ML100351155
| issue date = 02/04/2010
| title = 02/19/2010 Notice of Category 1 Public Meeting with Florida Power and Light Company to Discuss Risk Significance of the Two Preliminary Greater than Green Inspection Report 05000335-09-006 and 05000389-09-006
| author name = Desai B B
| author affiliation = NRC/RGN-II/DRS/EB1
| addressee name = Nazar M
| addressee affiliation = Florida Power & Light Co
| docket = 05000335, 05000389
| license number = DPR-067, NPF-016
| contact person =
| case reference number = IR-09-006
| document type = Letter, Meeting Notice
| page count = 5
}}


{{IR-Nav| site = 05000335 | year = 2009 | report number = 006 }}
{{IR-Nav| site = 05000335 | year = 2009 | report number = 006 }}


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II SAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET, SW, SUITE 23T85 ATLANTA, GEORGIA 30303-8931 February 4, 2010 Mr. Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power and Light Company P.O.Box 14000 Juno Beach, FL 33408-0420
[[Issue date::January 19, 2010]]


EA-09-321  Mr. Mano Nazar Executive Vice President and    Chief Nuclear Officer Florida Power & Light Company P.O. Box 14000 Juno Beach, FL 33408-0420
SUBJECT: MEETING ANNOUNCEMENT - PUBLIC MEETING CATEGORY 1 REGULATORY CONFERENCE - ST. LUCIE NUCLEAR PLANT DOCKET NOS. 50-335 AND 50-389
 
SUBJECT:  ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000335/2009006 AND 05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS


==Dear Mr. Nazar:==
==Dear Mr. Nazar:==
On September 4, 2009, U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your St. Lucie Nuclear Plant, Units 1 and 2. The enclosed inspection report documents the preliminary inspection results which were discussed with Mr. Gordon Johnston on September 4, 2009 and the final inspection results with Mr. Eric Katzman on December 10, 2009. The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The team reviewed selected procedures and records, observed activities, and interviewed personnel. Section 4OA5 of the enclosed report discusses an event which occurred in October 2008 when air from the containment instrument air (IA) system entered the Unit 1 Component Cooling Water (CCW) system. The air intrusion potentially rendered both trains of the safety-related CCW system inoperable. Two performance deficiencies were identified with this issue. The first performance deficiency involved a common cause failure vulnerability of the CCW system. Specifically, a non-safety system failure could result in a common cause failure of both trains of the CCW system. The second performance deficiency involved the failure to identify and correct a condition adverse to quality. Specifically, the licensee failed to properly determine the source of the air in-leakage into the CCW system and take appropriate corrective actions following the air intrusion event that occurred in October 2008. Further, the licensee's corrective action evaluation did not identify the common cause failure vulnerability discussed in the first performance deficiency.
This letter confirms the telephone conversation between Mr. Eric Katzman of your staff and Mr. Binoy Desai of the NRC, on February 1, 2010, concerning a Regulatory Conference, being conducted at your request, which has been scheduled for February 19, 2010, from 2 p.m. to 4:00 p.m. (EST). The meeting notice, which provides additional details, is enclosed. The purpose of the Regulatory Conference is to discuss the risk significance of the two preliminary greater than green inspection findings documented in NRC Inspection Report 05000335, 389/2009006. These findings concern apparent failures to identify and correct design issues associated with an October 8, 2008, air intrusion event from the Unit 1 containment instrument air system into the component cooling water system.
 
FP&L 2  The findings associated with the common cause vulnerability and the inadequate corrective actions were assessed based on the best available information. The two issues were preliminarily determined to be greater than Green findings using influencing assumptions and the Significant Determination Process (SDP). The SDP analysis determined that the two findings are potentially greater than very low safety significance because they potentially impacted the availability and thus the accident mitigation capability of the CCW system. These findings do not represent a current safety concern because the containment IA system has been isolated from the CCW system. Additionally, increased station sensitivity exists for recognizing and responding in a timely manner if a similar air intrusion event were to occur.
 
The performance deficiencies are documented in the enclosed report as two apparent violations (AVs). The first performance deficiency is an AV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to translate the design basis as specified in the license application, into specifications, drawings, procedures, and instructions resulting in the CCW system being susceptible to a common cause failure. The second performance deficiency is an AV of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for the failure to identify and correct a condition adverse to quality following the air intrusion event into the CCW system that occurred in October 2008. These AVs are being considered for escalated enforcement action in accordance with the NRC Enforcement Policy. The current Enforcement Policy is included on the NRC=s website at http://www.nrc.gov/reading-rm/adams.html.
 
In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our evaluation using the best available information and issue our final determination of safety significance within 90 days of this letter. The significance determination process encourages an open dialogue between the staff and the licensee; however, the dialogue should not impact the timeliness of the staff=s final determination. Before we make a final decision on this matter, we are providing you an opportunity to:  (1) present to the NRC your perspectives on the facts and assumptions used by the NRC to arrive at the finding and its significance at a Regulatory Conference or (2) submit your position on the finding to the NRC in writing. If you request a Regulatory Conference, it should be held within 30 days of the receipt of this letter and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective. If a Regulatory Conference is held, it will be open for public observation. The NRC will also issue a press release to announce the conference. If you decide to submit only a written response, such a submittal should be sent to the NRC within 30 days of the receipt of this letter. Please contact Mr. Steve Rose at (404) 562-4609 or Mr. Binoy Desai at (404) 562-4519 within 10 business days of the date of your receipt of this letter to notify the NRC of your intentions. If we have not heard from you within 10 days, we will continue with our significance determination and enforcement decisions and you will be advised by separate correspondence of the results of our deliberations on this matter. Since the NRC has not made a final determination in this matter, a Notice of Violation is not being issued at this time. In addition, please be advised that the number and characterization of the AVs violations may change as a result of further NRC review.
 
FP&L 3  In addition, this report documents two NRC-identified findings of very low safety significance which were determined to be violations of NRC requirements. The NRC is treating these two violations as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy because of their very low safety significance and because they were entered into your corrective action program. If you contest these NCVs, 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 resident inspector at the St. Lucie Nuclear Plant. In addition, if you disagree with the characterization of any finding in this report, 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 Resident Inspector at the St. Lucie Nuclear Plant. The information you provide will be considered in accordance with the Inspection Manual Chapter 0305. 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 (PARS) component of the 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/  Kriss M. Kennedy, Director Division of Reactor Safety 


===Enclosure:===
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 ins pection in the NRC Public Document Room (PDR) or from the Publicly Available Records (PARS) component of NRC's Agencywide Document Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Inspection Report 05000335/2009006, 05000389/2009006     


===w/Attachment:===
Should you have any questions concerning this meeting, please contact me at (404) 562-4519.
Supplemental Information Docket Nos.: 50-335, 50-389 License Nos.: DPR-67 and NPF-16  cc w/encl:  (See page 4)
FP&L 3  In addition, this report documents two NRC-identified findings of very low safety significance which were determined to be violations of NRC requirements. The NRC is treating these two violations as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy because of their very low safety significance and because they were entered into your corrective action program. If you contest these NCVs, 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 resident inspector at the St. Lucie Nuclear Plant. In addition, if you disagree with the characterization of any finding in this report, 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 Resident Inspector at the St. Lucie Nuclear Plant. The information you provide will be considered in accordance with the Inspection Manual Chapter 0305.


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 (PARS) component of the 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: Original signed by Mark Franke for/
Binoy B. Desai, Chief Engineering Branch 1 Division of Reactor Safety


Sincerely,/RA/  Kriss M. Kennedy, Director Division of Reactor Safety 
Docket Nos.: 50-335, 50-389 License Nos.: DPR-67, NPF-16


===Enclosure:===
===Enclosure:===
Inspection Report 05000335/2009006, 05000389/2009006 
As stated  
 
===w/Attachment:===
Supplemental Information  Docket Nos.: 50-335, 50-389 License Nos.: DPR-67 and NPF-16
 
cc w/encl:  (See page 4)  xx  PUBLICLY AVAILABLE G  NON-PUBLICLY AVAILABLE G  SENSITIVE      xx  NON-SENSITIVE ADAMS:  G Yes ACCESSION NUMBER:_________________________  xxG  SUNSI REVIEW COMPLETE OFFICE RII:DRS RII:DRS RII:DRS RII:DRP CONTRACTOR CONTRACTOR RII:DRP SIGNATURE RA RA RA RA RA RA RA NAME SROSE RMOORE JHAMMAN RTAYLOR MSHYLAMBERG NDELIAGRECA MSYKES DATE 11/30/2009 11/19/2009 1/12/2010 11/20/2009 11/18/2009 11/5/2009 1/13/2010 E-MAIL COPY?    YES NO  YES NO  YES NO  YES NO  YES NO    YES NO    YES NOOFFICE RII:DRS RII:OE      SIGNATURE RA RA      NAME BDESAI CEVANS      DATE 1/11/2010 1/13/2010      E-MAIL COPY?    YES NO YES          NO      OFFICIAL RECORD COPY DOCUMENT NAME:  S:\DRS\ENG BRANCH 1\BRANCH INSPECTION FILES\CDBI INSPECTIONS\CDBI INSPECTIONS\INSP REPORTS\CDBI FINAL INSPECTION REPORTS\REV 1 ST LUCIE 2009006 CDBI REPORT (SDR).DOC FP&L 4  cc w/encl: Richard L. Anderson Site Vice President St. Lucie Nuclear Plant Electronic Mail Distribution  Robert J. Hughes Plant General Manager St. Lucie Nuclear Plant Electronic Mail Distribution  Mark Hicks Operations Manager St. Lucie Nuclear Plant Electronic Mail Distribution Rajiv S. Kundalkar Vice President - Fleet Organizational Support Florida Power & Light Company Electronic Mail Distribution Eric Katzman Licensing Manager St. Lucie Nuclear Plant Electronic Mail Distribution Abdy Khanpour Vice President Engineering Support Florida Power and Light Company P.O. Box 14000 Juno Beach, FL  33408-0420  McHenry Cornell Director Licensing and Performance Improvement Florida Power & Light Company Electronic Mail Distribution  Alison Brown Nuclear Licensing Florida Power & Light Company Electronic Mail Distribution  Faye Outlaw County Administrator St. Lucie County Electronic Mail Distribution Mitch S. Ross Vice President and Associate General Counsel Florida Power & Light Company Electronic Mail Distribution  Marjan Mashhadi Senior Attorney Florida Power & Light Company Electronic Mail Distribution  William A. Passetti Chief Florida Bureau of Radiation Control Department of Health Electronic Mail Distribution Ruben D. Almaguer Director Division of Emergency Preparedness Department of Community Affairs Electronic Mail Distribution  J. Kammel Radiological Emergency Planning Administrator Department of Public Safety Electronic Mail Distribution  Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power & Light Company Electronic Mail Distribution  (Vacant)
Vice President Nuclear Plant Support Florida Power & Light Company Electronic Mail Distribution  Jack Southard Director Public Safety Department St. Lucie County Electronic Mail Distribution FP&L 5  Letter to Mano Nazar from Kriss Kennedy dated January 19, 2010.
 
SUBJECT:  ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000335/2009006 AND 05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS  Distribution w/encl: C. Evans, RII  L. Slack, RII OE Mail  RIDSNRRDIRS PUBLIC RidsNrrPMStLucie Resource Enclosure U. S. NUCLEAR REGULATORY COMMISSION  REGION II Docket Nos.: 50-335, 50-389    License Nos.: DPR-67 and NPF-16 Report Nos.: 05000335/2009006, 05000389/2009006    Licensee: Florida Power & Light Company (FP&L)
Facility: St. Lucie Nuclear Plant, Units 1 & 2    Location: Jensen Beach, FL 34957 Dates:  August 3-14 (Weeks 1 & 2)    August 31-September 4 (Week 3)
Inspectors: S. Rose, Senior Operations Inspector (Lead) R. Moore, Senior Reactor Inspector J. Hamman, Reactor Inspector R. Taylor, Senior Reactor Inspector M. Shylamberg, Contractor N. Della Greca, Contractor    Approved by: Binoy Desai, Chief Engineering Branch 1    Division of Reactor Safety Enclosure
 
=SUMMARY OF FINDINGS=
IR 05000335/2009006, 05000389/2009006; 8/3/2009 - 9/4/2009; St. Lucie Nuclear Plant, Units 1 and 2; NRC Component Design Bases Inspection. This inspection was conducted by a team of four NRC inspectors from the Region II office, and two NRC contract inspectors. Two findings of very low significance (Green) were identified during this inspection and were classified as non-cited violations. Also, two apparent violations (AV) with potential safety significance greater than Green were identified. The significance of most findings is indicated by their color (Green, White,
 
Yellow, Red) using IMC 0609, "Significance Determination Process" (SDP). 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," (ROP) Revision 4, dated December 2006.
 
===Cornerstone: Mitigating Systems===
: '''Green.'''
The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to translate the design basis as specified in the license application into specifications, drawings, procedures, and instructions. The licensee did not ensure that the component cooling water (CCW) surge tank design included adequate overpressure protection for all procedurally allowed configurations as required by the applicable ASME Boiler and Pressure Vessel Code, Section VIII,
Division 1. The code requires that no intervening stop valves be between the vessel and its protective device or devices or between the protective devices and the point of discharge. The team concluded that stop valve V6466 was an intervening stop valve for the CCW surge tank vent path to the chemical drain tank (CDT). The issue was entered in the licensee's corrective action program as condition report (CR)2009-23473. Immediate licensee corrective actions included verification that the valve was in its open position and the implementation of administrative controls to maintain the valve open. This finding is associated with the Mitigating Systems Cornerstone attribute of
Design Control, i.e. initial design, was determined to be more than minor because it impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined that if left uncorrected, this design deficiency had the potential to impact the operability of safety-related systems and, thus, become a more significant safety concern in that a closed intervening valve had the potential for overpressurizing the CCW surge tank. The team assessed this finding for significance in accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, Significance Determination Process (SDP) for Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green), in that no actual loss of safety system function was identified. The team reviewed the finding for cross-cutting aspects and concluded that this finding did not have an associated cross-cutting aspect because the design of the CCW surge tank relief was established in an original plant design, and therefore, was not representative of current licensee performance.  [Section 1R21.2.2]
Enclosure
: '''Green.'''
The inspectors identified a finding involving a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensee's failure to maintain the safety-related 125V DC system design basis information consistent with the plant configuration. Specifically, a revision to the Unit 1, safety-related 125V DC system analysis incorporated incorrect design input specifications. The issue was entered in the licensee's corrective action program as CR 2009-24517. Licensee corrective actions included incorporating the correct design input and specifications by revising the calculations. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Design Control. It impacted the cornerstone objective because if left uncorrected, it had the potential to lead to a more significant safety concern in that future design activity or operability assessments would assume the lower voltage (100V DC vs. actual 105V DC) value acceptable for assuring the adequacy of voltage to the safety-related inverters. The team assessed this finding for significance in accordance with NRC Manual Chapter 0609, using the Phase I SDP worksheet for mitigating systems and determined that the finding was of very low safety significance (Green) since it was a design deficiency determined not to have resulted in a loss of safety function. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to ensure that procedures (specifically ENG-QI 1.5) were available and adequate to assure nuclear safety (specifically, complete, accurate and up-to-date design documentation): H.2(c).  [Section 1R21.2.20]
* TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensee's failure to identify that the CCW system met its license specifications related to common cause failure vulnerabilities. Specifically, a non-safety system failure (i.e. waste gas compressor aftercoolers affecting both units, or containment IA compressors affecting Unit 1 only) could result in a common cause failure of both trains of a safety system (i.e. CCW system). The issue was entered into the licensee's corrective action program as CR 2009-22929 with actions to evaluate the past operability of the CCW system during the air intrusion event. Licensee corrective actions included isolating the CCW system from the containment IA compressors. The finding was determined to be more than minor because if left uncorrected, it could affect the availability, reliability and capability of a safety system to perform its intended safety function. Specifically, with this vulnerability, a failure of the waste gas aftercooler (both units) or a failure of the containment IA compressors (Unit 1 only) could cause air intrusion into the CCW system and lead to a loss of CCW event, therefore, failing to ensure that adequate cooling would be available or maintained to essential equipment used to mitigate design bases accidents. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, using the Phase I and Phase II SDP worksheets for mitigating systems. It was determined that a Phase III analysis was required since this finding represented a potential loss of safety system function for multiple trains which was not addressed by the Phase II pre-solved tables/worksheets. Based on the Phase III SDP, the finding was preliminarily determined to be greater than
: '''Green.'''
The team reviewed the finding for cross-cutting aspect and concluded that this finding did not have an
Enclosure associated cross-cutting aspect because the design of the CCW system was established in an original plant design, and therefore, was not representative of current licensee performance.  [Section 4OA5]
* TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to implement adequate corrective actions associated with the CCW air intrusion event that occurred in October, 2008. The corrective actions were inadequate in that the licensee failed to identify and correct the cause of air intrusion. The issue was entered in the licensee's corrective action program as CR 2009-25209 to address the ineffective corrective actions for the air intrusion event. Licensee corrective actions included isolating the CCW system from the containment IA compressors. The finding was determined to be more than minor because it affected the availability, reliability and capability of a safety system to perform its intended safety function. Specifically, without knowing the leak path from the containment IA compressors to the CCW system, the licensee could not ensure that adequate cooling would be available or maintained to essential equipment used to mitigate design bases accidents. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, using the Phase I and Phase II SDP worksheets for mitigating systems. It was determined that a Phase III analysis was required since this finding represented a loss of safety system function for multiple trains which was not addressed by the Phase II pre-solved tables/worksheets. Based on the Phase III
SDP, the finding was preliminarily determined to be greater than
: '''Green.'''
This finding was determined to have a cross-cutting aspect in the area of Human Performance, Decision Making, specifically H.1(a).  [Section 4OA5]
 
=REPORT DETAILS=
 
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
{{a|1R21}}
==1R21 Component Design Bases Inspection==
{{IP sample|IP=IP 71111.21}}
===.1 Inspection Sample Selection Process===
 
The team selected risk significant components and operator actions for review using information contained in the licensee's Probabilistic Risk Assessment (PRA). In general, this included components and operator actions that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1 X10-6. The sample included 20 components, six operator actions, and five operating experience items. Additionally, the team reviewed one permanent plant modification by performing activities identified in IP 71111.17, "Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications."  The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases had been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modifications, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These reliability issues included review of items related to performance and surveillance test failures, corrective actions due to repeat maintenance, maintenance rule (a)1 status, Regulatory Issue Summary (RIS) 05-020 (formerly Generic Letter (GL) 91-18) conditions, NRC resident inspector input of problem equipment, system health reports, industry operating experience and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.
 
===.2 Results of Detailed Reviews===
 
===.2.1 Component Cooling Water (CCW) Pumps 1A/1B/1C===
 
====a. Inspection Scope====
The team reviewed the design bases documents (DBD), related design basis documentation, drawings, technical specifications (TS), and the final safety analysis report (FSAR) to identify design, maintenance, and operational requirements for the CCW pumps. The team reviewed the system configuration and design calculations to verify that adequate net positive suction head (NPSH) would be available during accident conditions. Maintenance history, as demonstrated by system health reports, corrective maintenance documentation, condition reports (CRs), and surveillance test results, were reviewed to verify the design bases had been maintained; potential degradation was being monitored; and that identified degradation or malfunctions had been adequately addressed. The team reviewed normal, abnormal, and emergency 
 
operating procedures to verify correct implementation of design bases. The team verified that the equipment periodic maintenance performed was consistent with vendor recommendations. Additionally, the team conducted a field walkdown of the CCW pumps with the licensee staff to assess observable material condition and to verify that the installed configuration was consistent with the design basis and plant drawings. The team reviewed voltage drop calculations to confirm that the voltage available at the motor terminals as well as at the circuit breakers was adequate to ensure that the pumps can perform their safety function when called upon. Additionally, the team verified that the horsepower rating of the motors were correctly identified in the load flow analysis and that adequate protection was provided for the motors. The team reviewed control wiring diagrams to confirm that the operation of the pumps conformed to their intended function.
 
====b. Findings====
No findings of significance were identified; however, see section
{{a|4OA5}}
==4OA5 for two findings related to the CCW system.==
 
===.2.2 Component Cooling Water Surge Tank===
 
====a. Inspection Scope====
For the CCW surge tank the team reviewed DBDs, Technical Specifications, FSAR, calculations, and drawings. Specific design requirements for the CCW surge tank levels, tank leakage and make up rate, minimum level vs. NPSH allowed and vortex limits, tank baffle location and height, and tank implosion and overpressure protection were reviewed and compared to as-built configuration. The team also reviewed all CCW system operating conditions to verify that design, maintenance, and operational requirements were appropriate. The CCW flow assumptions in the FSAR accident analysis were also reviewed to verify that the surge tank was capable of performing the intended safety functions. Calculations were also reviewed to verify that the surge tank met applicable ASME requirements. Maintenance, corrective actions, and design change history were reviewed to assess potential component degradation and subsequent impacts on design margins.
 
====b. Findings====
 
=====Introduction:=====
The inspectors identified a finding of very low safety significance (Green) involving a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensee's failure to translate the design basis as specified in the license application, into specifications, drawings, procedures, and instructions. Specifically, the licensee's failure to assure that the CCW surge tank design included adequate overpressure protection for all configurations allowed by plant procedures, as required by the applicable ASME Boiler and Pressure Vessel Code, Section VIII, Division 1, was identified by the inspectors as a performance deficiency.
 
=====Description:=====
The review of the Unit 1 CCW surge tank's design and operation identified that the tank pressure relief required by the ASME Code (ASME Section VIII) was provided via a 2-inch vent line. This vent line was routed to a diverting air-operated 
 
valve, RCV-14-1. This valve was normally open to atmosphere; however, in the event of high radiation, this valve re-aligns the relief path from the atmosphere and diverts the vent/overflow to the liquid waste management system chemical drain tank (CDT) 1A. A similar re-alignment would take place on a loss of instrument air. The CDT 1A was a closed tank and was vented to a sump pit by a 1-1/2" line. A maintenance valve, V6466, was installed between the diverting air-operated valve RCV-14-1 and CDT 1A. A similar configuration existed for Unit 2. ASME Section VIII, Division 1, 1971 Edition, paragraph UG-134(e) states, "There shall be no intervening stop valves between the vessel and its protective device or devices or between the protective devices and the point of discharge-"  The requirement to comply with the ASME Code requirements was based on Unit 1 FSAR Table 3.2-2, which states the minimum code requirements for Quality Group C pressure vessels must comply with ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. The Quality Group C designation for the safety-related portion of the CCW system was provided in the Unit 1 DBD for CCW. The Unit 1 CCW Tank was procured per specification FLO-8770-764, originally issued on October 31, 1971. Therefore, the inspectors concluded that ASME Section VIII, Division 1, 1971 edition applied. The team concluded that valve V6466 was an intervening stop valve for the CCW Surge Tank vent path to the CDT. The licensee issued CRs 2009-25276 and 2009-23473 to evaluate this condition. The licensee's review determined that valve V6466 was a normally open valve. Additionally, there were a number of floor drains (although not formally maintained clear of blockages) that tie in the header between valves RCV-14-1 and V6466 that would provide an alternate relief path should valve V6466 be closed. The licensee's review of records for the past 10 years identified that for Unit 1, valve V6466 was never closed. The licensee identified that for Unit 2, the valve had been closed in the past, however, during that time, the drains were rerouted to an alternate tank, thus providing the required relief path. The team concluded from this information that this design deficiency did not represent an actual loss of safety system function.
 
The team reviewed the finding for cross-cutting and concluded that this finding did not have an associated cross-cutting aspect because the design of the CCW surge tank relief was established in an original plant design, therefore, not representative of current licensee performance.
 
=====Analysis:=====
The licensee's failure to assure the CCW surge tank design included adequate overpressure protection as required by the applicable ASME Boiler and Pressure Vessel Code was identified as a performance deficiency. This finding, associated with the Mitigating Systems Cornerstone attribute of Design Control, i.e. initial design, was determined to be more than minor because it impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined that if left uncorrected, this design deficiency had the potential to impact the operability of safety-related systems and, thus, become a more significant safety concern. Specifically, during an overpressure event, if intervening valve V6466 was shut and the floor drain lines clogged, the CCW surge tank vent path to the CDT would be obstructed to the point that a loss of CCW surge tank could occur, therefore, increasing the likelihood of a loss of CCW. The team assessed this finding for significance in 
 
accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, Significance Determination Process (SDP) for Reactor Inspection Findings for At-Power Situations, and determined that it was of very low safety significance (Green), in that no actual loss of safety system function was identified. The team concluded that this finding did not have an associated cross-cutting aspect because the performance deficiency was not reflective of current plant performance. The design of the CCW surge tank relief was established during original plant design; and the last design change associated with the CCW surge tank was in 2001.
 
=====Enforcement:=====
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications. Contrary to the above, the licensee failed to assure that applicable regulatory requirements and the design bases were correctly translated into actual plant specifications. The installed CCW surge tank pressure relief protection did not meet the Code requirements described in the Unit 1 FSAR Table 3.2-2. The FSAR required that the minimum code requirements for Quality Group C pressure vessels to be ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. Specifically, ASME Boiler and Pressure Vessel Code, Section VIII, Division 1 requirements for the overpressure protection for the CCW surge tank were not properly implemented. This design deficiency was an original plant design and has existed since the operating licenses were issued. Because this violation was of very low safety significance (Green) and it was entered into the licensee's corrective action program as CR 2009-25276 and CR 2009-23473, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000335,389/2009006-01, Failure to Meet the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1 Requirements for the Overpressure Protection for the CCW Surge Tank.
 
===.2.3 Instrument Air Emergency Cooling System===
 
====a. Inspection Scope====
The team reviewed the drawings, TS, and the FSAR to identify the design, maintenance, and operational requirements for the instrument air (IA) emergency cooling system. The team reviewed the system configuration and normal, abnormal, and emergency operating procedures to verify correct implementation of the design bases. Maintenance history, as demonstrated by system health reports, corrective maintenance documentation, CRs, and surveillance test results, was reviewed to verify that the design bases had been maintained and correctly implemented; potential degradation was being monitored; and that identified degradation or malfunctions had been adequately addressed. The team verified that the equipment periodic maintenance performed was consistent with vendor recommendations. Additionally, the team conducted a field walkdown of the IA emergency cooling system with the licensee staff to assess observable material condition and to verify that the installed configuration was consistent with the design basis and plant drawings.
 
====b. Findings====
 
=====Introduction:=====
An unresolved item (URI) was identified related to the performance monitoring of the IA emergency cooling system. The team determined that the performance monitoring did not provide reasonable assurance that the system was capable of fulfilling its intended function. This failure to monitor the performance of the IA emergency cooling system was a performance deficiency. The system was identified to be in the scope of the maintenance rule (MR), 10 CFR 50.65(a)(1), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because it is included in the St. Lucie emergency operating procedures.
 
=====Description:=====
The IA emergency cooling system is an alternate source of cooling for IA compressors A and B. The system is a small, closed cooling system with a pump, head tank, fan cooled radiator and connecting piping and valves to the IA compressors. The normal cooling water to the compressors is provided by the turbine cooling water (TCW) system which does not have power available after a loss of offsite power (LOOP)accident. These IA compressors and the emergency cooling system pump are provided with vital power so that the compressors can be manually loaded in accordance with 1[2]-EOP-09, Loss of Offsite Power, Rev. 38. During the inspection, the team requested design, maintenance, or operational documentation that would provide reasonable assurance that the emergency cooling system could perform its intended function of providing adequate cooling for IA compressors A and B during a LOOP event. There were no documented specifications, analysis, or testing available to verify the adequacy of the emergency cooling water system to support continued operation of the IA compressors. The team reviewed the routine testing performed on the emergency cooling system and concluded that this testing did not verify the system adequacy or provide the capability to identify potential degradation of the equipment. For example, Procedure OSP-69.13A, "ESF-18 Month Surveillance for SIAS/CIS/CSAS," Rev. 2, aligned the IA emergency cooling system to the 2B IA compressor; however, the test configuration was in parallel with the higher capacity 2C IA compressor and, therefore, it was not possible to determine if the 2B IA compressor was loaded and the emergency cooling system was capable of sustaining loaded compressor operation.
 
Procedure 2-0330020, Appendix H, "Instrument Air Emergency Cooling Test," Rev. 56, required the recirculation pump to be run for 30 minutes but stated that starting the IA compressor was an option. The licensee did not provide past test information that demonstrated the IA compressor was run or loaded during this routine test. The inspectors concluded that the routine testing performed verified the flow path to the unloaded compressor but did not verify that the cooling system was capable of supporting sustained operation of the compressor. The licensee documented this issue in CR 2009-22766 and planned to perform a formal test of the system to demonstrate its capabilities.
 
The team noted that the IA system at St. Lucie was a non-safety related system. Station design was that air-operated components fail to a safe position or are provided with an air accumulator. The emergency cooling system for the IA compressors was identified to be in the scope of the MR because it is a non-safety related system that was used in the emergency operating procedures (10 CFR 50.65(b)(2)).
 
This item will remain unresolved pending the completion of the station's testing, and NRC review of the results of the IA emergency cooling system's capability to provide cooling for the IA compressors under conditions comparable to those expected during a LOOP event. The item is identified as URI 05000335,389/2009006-02, Adequacy of Performance Monitoring of the IA Compressor Emergency Cooling System.
 
===.2.4 GD-1/2 Gravity Damper On HVS-5A/B Outlet===
 
====a. Inspection Scope====
The team reviewed the DBD, related design basis documentation, drawings, TS, and the FSAR to identify design, maintenance, and operational requirements for the GD-1/2 Gravity Damper. The team reviewed the system configuration and normal, abnormal, and emergency operating procedures to verify correct implementation of design bases. Maintenance history, as demonstrated by system health reports, corrective maintenance documentation, and CRs was reviewed to verify the design bases had been maintained; potential degradation was being monitored; and that identified degradation or malfunctions had been adequately addressed. The team verified that the equipment periodic maintenance performed was consistent with vendor recommendations. Additionally, the team conducted a field walkdown of the GD-1/2 Gravity Damper with the licensee staff to assess observable material condition and to verify that the installed configuration was consistent with the design basis and plant drawings.
 
====b. Findings====
No findings of significance were identified.
 
===.2.5 Pressurizer Relief Valve Isolation Valves, V1403 and V1405===
 
====a. Inspection Scope====
The team reviewed the system DBD, related design basis support documentation, drawings, TS, and the FSAR to identify design, maintenance, and operational requirements for these motor operated valves (MOVs). Maintenance history, as demonstrated by system health reports, preventive and corrective maintenance, and CRs, was reviewed to verify that potential degradation was being monitored and addressed. The MOV sizing calculations were reviewed to verify that the valves could operate during all credited design bases events and that the licensee appropriately translated the correct valve dimensions and other significant characteristics into the sizing calculations. A review was conducted of the licensee's testing procedures and results from diagnostic valve testing to verify that the MOVs were tested in a manner that would detect a malfunctioning valve and verify proper operation of the valve. The team reviewed vendor recommendations for preventative maintenance and operation to verify that the maintenance practices were consistent with design basis requirements.
 
====b. Findings====
No findings of significance were identified
 
===.2.6 Battery Charger 1B===
 
====a. Inspection Scope====
The team reviewed the Class 1E DC electrical distribution system DBD, related design basis support documents, drawings, appropriate sections of the TS, and the FSAR to identify the design bases, maintenance requirements and the operational design requirements of the battery charger. The team reviewed the battery charger sizing calculation, its conformance to the original design, and its capability to support current load demands and battery charging requirements. The team also reviewed testing requirements and test procedures developed to demonstrate the design capabilities of the charger under various plant conditions. The review included the vendor manual and the procedures that were developed to verify that the installation, operation, and maintenance were in accordance with manufacturer's recommendations. The team reviewed the health report and the results of recent tests to verify that the current performance was within accepted limits. Additionally, the team reviewed selected corrective action reports to verify that anomalies were addressed and corrected. A field walkdown was performed to assess the observable material condition of the batteries, battery chargers, and inverters.
 
====b. Findings====
No findings of significance were identified.
 
===.2.7 125V DC Bus 1B Power Panel & Cross-Tie Breakers to 125V DC Bus 1AB===
 
====a. Inspection Scope====
The team reviewed the Class 1E DC electrical distribution system DBD, applicable drawings and documents, including appropriate sections of the FSAR, to identify the design bases, maintenance and design requirements and to verify conformance of the design to the licensing bases. The team reviewed preventive maintenance and testing procedures to confirm that the bus and breakers were maintained in accordance with manufacturer's recommendations. The team also addressed short circuit capabilities and circuit breaker/protective device coordination to verify that the power panels and breakers were applied within the vendor published interruptive ratings and to confirm the capability of the bus to support load demands under accident and station blackout conditions. Additionally, the team reviewed recent system modifications and selected corrective action reports to verify that anomalies were addressed and corrected. The team reviewed operation requirements for the system and the interlocks provided to prevent paralleling of divisional power through DC bus 1AB. The team reviewed the interfaces between the safety-related bus and non-safety-related loads and the protection provided to ensure that the safety-related bus and battery were not overloaded beyond calculated limits. A field walkdown of the power panels was performed to assess their installation, observable material conditions and to verify the current alignment of the buses.
 
====b. Findings====
No findings of significance were identified.
 
===.2.8 Engineered Safety Features Actuation System and Diverse Scram System===
 
====a. Inspection Scope====
The team reviewed the engineered safety features actuation system (ESFAS) and diverse scram system (DSS) design basis document and applicable sections of the TS and FSAR to identify the design bases and the operational and maintenance requirements for the ESFAS and DSS. The team reviewed the DSS components including transmitters, logic modules, control and monitoring instrumentation, actuation relays and contactors, selected components, and instrument loops associated with the ESFAS. The review included a detailed evaluation of instrument loop diagrams, control logic, and wiring diagrams to confirm that the design conformed to the intended operation of the systems. The review also addressed voltage requirements and voltage available at the various components, circuit protection, channel separation, and electrical isolation. The team reviewed test procedures and evaluated the tests performed to demonstrate the capability of the systems to perform the design basis functions. The review included instrument and loop calibration procedures, test results, and adequacy of overlapping tests. The team confirmed that system and component maintenance was conducted per vendor recommendations. Additionally, a review of the latest system health report and recent problem reports was conducted to evaluate whether component concerns were adequately addressed and corrected and that their aging issues were appropriately addressed. The team conducted a field verification of selected components to evaluate installation criteria used and to assess their observable material condition.
 
====b. Findings====
No findings of significance were identified.
 
===.2.9 Pressurizer Pressure Instrumentation===
 
====a. Inspection Scope====
The team reviewed applicable sections of the pressurizer system DBD and applicable sections of the TS and FSAR to identify the design bases and the operational and maintenance requirements for the low range pressure control functions and components, including transmitters, logic modules, control and monitoring instrumentation, and actuation relays. The team conducted a detailed review of instrument loop diagrams and control logic and wiring diagrams to confirm that the design conformed to the intended functions of the instrument loops. The review also evaluated voltage requirements and voltage available at the instrument components, circuit protection, channel separation, and electrical isolation. Additionally, the team reviewed test procedures and evaluated the periodic tests performed to demonstrate the capability of the instrument loops to perform their design basis functions. The review included component and loop calibration procedures, test results, and adequacy of overlapping 
 
tests. The team reviewed the latest system health report and recent corrective action reports to evaluate whether component concerns were adequately addressed and corrected and that aging issues were appropriately addressed. The team conducted a field walkdown of accessible instrument loop components to assess their observable material condition.
 
====b. Findings====
No findings of significance were identified.
 
===.2.10 Start-Up Transformers 1A and 1B and associated supply and feeder breakers===
 
====a. Inspection Scope====
The team reviewed the TS, DBD, FSAR, and alternate current (AC) load flow analysis, as well as the Unit 1 computer modeling to assess whether station startup transformers would have sufficient capacity to support required loads in accident/event conditions. The team further reviewed coordination studies to assess the effects of inrush currents and protective schemes in transformer relays to determine if adequate protection was provided. The team reviewed maintenance records, system health reports and corrective action records to assess any adverse operating trends. A walk down of the Start-Up Transformers 1A and 1B was performed to observe material condition and vulnerability to hazards.
 
====b. Findings====
No findings of significance were identified.
 
===.2.11 480VAC Load Center 1AB Cross-Tie Breaker (to either 480V 1A Load Center or 1B Load Center)===
 
====a. Inspection Scope====
The team reviewed the TS, DBD, design drawings, calculations, vendor manuals and plant procedures to identify the design, maintenance and operational requirements for the cross-tie breaker. Electrical elementary drawings and wiring diagrams were reviewed to verify that power sources would be available and adequate to power the appropriate safety loads during accident/event conditions. The team reviewed preventive maintenance and testing results to determine if the breakers were maintained in accordance with industry and vendor standards and recommendations. The team reviewed short circuit and protection calculations to ensure that the breakers could provide the appropriate interrupting and coordination protection. Selected corrective action reports were reviewed to determine if conditions adverse to quality were appropriately addressed and corrected. A walk down of the cross-tie breaker to load center 1A was performed to assess installation, configuration, observable material condition and vulnerability to hazards.
 
====b. Findings====
No findings of significance were identified.
 
===.2.12 480V Switchgear 1B2 (feeder and supply breakers and transformers)===
 
====a. Inspection Scope====
:  The team reviewed the TS, DBD, design drawings, calculations, vendor data and manuals and plant procedures to identify the design, maintenance and operational requirements. Electrical elementary drawings and wiring diagrams were reviewed to verify that power sources would be available and adequate to power the appropriate safety loads during accident/event conditions. The team reviewed preventive maintenance and testing procedures and results to determine if the breakers were maintained in accordance with industry and vendor standards and recommendations.
 
The team reviewed short circuit and protection calculations to ensure that the breakers could provide the appropriate interrupting and coordination protection. Selected corrective action reports were reviewed to determine if conditions adverse to quality were appropriately addressed and corrected. A walk down of the 480V 1B2 Breaker panel was performed to assess installation, configuration, observable material condition and vulnerability to hazards.
 
====b. Findings====
No findings of significance were identified.
 
===.2.13 Temperature Indication Switches for Reactor Coolant Pump (RCP) 1A and 1B CCW Seal Cooler Heat Exchanger Outlet (TIS-14-32A1/B1/B2/A2)===
 
====a. Inspection Scope====
The team reviewed design and licensing basis documents, drawings and vendor manuals to identify the design requirements for the temperature indication switches. The team reviewed set point calculations to verify that set points were established in accordance with vendor data, equipment capability and system design parameters. Procedures were reviewed to verify alarm levels had been consistently translated from calculation data to ensure appropriate protection for an RCP seal leak. The team reviewed calibration records and procedures to verify that instrument accuracy was monitored and maintained. Maintenance history, as demonstrated by work orders and corrective action records, was reviewed to note any anomalies in equipment history and to verify corrective actions were accomplished in a timely matter.
 
====b. Findings====
No findings of significance were identified.
 
===.2.14 Intersystem Loss of Coolant Accident (LOCA) Instrumentation===
 
====a. Inspection Scope====
The team reviewed design and licensing basis documents, drawings and vendor manuals to identify the design requirements and capabilities of the intersystem LOCA instrumentation. The following instrumentation was included in the review:  CCW Surge Tank Level (LS-14-1A and B; LS-14-5, LG-14-2A and B); CCW System Radiation Monitors; Reactor/Auxiliary Building (RAB) Sump Level; and RAB Radiation Monitors.
 
The team reviewed set point and level calculations to verify that set points and levels were established in accordance with vendor data, equipment capability and system design parameters. Appropriate procedures were reviewed to verify set point data and alarm points had been consistently translated. The team reviewed calibration records and procedures to verify that instrument accuracy was monitored and maintained. Maintenance history, as demonstrated by work orders and corrective action records, was reviewed to note any anomalies in equipment history and to verify corrective actions were accomplished in a timely matter.
 
====b. Findings====
No findings of significance were identified.
 
===.2.15 Safety Injection Tank (SIT) Instrumentation===
 
====a. Inspection Scope====
The team reviewed design and licensing basis documents, drawings and vendor manuals to identify the design requirements and capability of the safety injection tank instrumentation. The team reviewed set point calculations to verify that set points and levels were established in accordance with vendor data, equipment capability and system design parameters. Appropriate procedures were reviewed to verify alarm levels and set point data had been consistently translated. The team reviewed calibration records and procedures to verify that instrument accuracy was monitored and maintained. Maintenance history, as demonstrated by work orders and CR's, was reviewed to note any anomalies in equipment history and to verify corrective actions were accomplished in a timely matter.
 
====b. Findings====
No findings of significance were identified.
 
===.2.16 Safety Injection (SI) System Check Valves (V3227, V07174, V07172, V3106, V3107)===
 
====a. Inspection Scope====
The team reviewed the DBD, related design basis documentation, drawings, TS, and the FSAR to identify design, maintenance, and operational requirements for selected SI system check valves. Maintenance history, as demonstrated by system health reports, preventive and corrective maintenance, and CRs, was reviewed to verify that potential 
 
degradation was being monitored and addressed. The team conducted interviews with the SI System Engineer to obtain additional information and verify the station's implementation and analysis of industry operating experience related to check valves.
 
====b. Findings====
No findings of significance were identified.
 
===.2.17 Volume Control Tank (VCT) MOVs 2501 & 2504===
 
====a. Inspection Scope====
The team reviewed the system DBD, related design basis support documentation, drawings, TS, and the FSAR to identify design, maintenance, and operational requirements for these MOVs. Maintenance history, as demonstrated by system health reports, preventive and corrective maintenance, and CRs, was reviewed to verify that potential degradation was being monitored and addressed. The MOV sizing calculations were reviewed to verify that the valves could operate during all credited design bases events and that the licensee appropriately translated the correct valve dimensions and other significant characteristics into the sizing calculations. A review was conducted of the licensee's testing procedures and results from diagnostic valve testing to verify the MOVs were tested in a manner that would detect a malfunctioning valve and verify proper operation of the valve. The team reviewed vendor recommendations for preventative maintenance and operation to determine if maintenance practices were consistent with design basis requirements.
 
====b. Findings====
No findings of significance were identified.
 
===.2.18 CCW Control Valves (HCV-14-8A, HCV-14-8B, & HCV-14-9)===
 
====a. Inspection Scope====
The team reviewed applicable portions of the FSAR, DBD, and drawings to identify design basis requirements for these valves. The air operator sizing calculations were reviewed to verify inputs were consistent with the most limiting design basis operating conditions. Procurement documentation for the solenoids was reviewed to verify compliance with environmental qualification (EQ) requirements. Stroke time surveillance test procedures/results were reviewed to verify that stroke times were consistent with design basis requirements and to identify any adverse trends. The vendor manual was reviewed to identify recommendations for inspection and maintenance. The CR history was reviewed to identify failures and determine whether they were entered into the MR data base as appropriate.
 
====b. Findings====
No findings of significance were identified.
 
===.2.19 SIT Outlet Valves (V3634, V3614, V3624, & V3644)===
 
====a. Inspection Scope====
The team reviewed the system DBD, related design basis support documentation, drawings, TS, and the FSAR to identify design, maintenance, and operational requirements for these MOVs. Maintenance history, as demonstrated by system health reports, preventive and corrective maintenance, and CRs, was reviewed to verify that potential degradation was being monitored and addressed. A review was conducted of the licensee's testing procedures and results from diagnostic valve testing to verify the MOVs were tested in a manner that would detect a malfunctioning valve and verify proper operation of the valve. The team reviewed maintenance practices and vendor recommendations for preventative maintenance and operation to verify that the valves were being maintained consistent with design basis requirements.
 
====b. Findings====
No findings of significance were identified.
 
===.2.20 Motors and Electrical Components in Inspection Scope===
 
====a. Inspection Scope====
The team reviewed AC and direct current (DC) load flow and voltage (V) drop calculations to determine if each motor within the inspection sample had adequate terminal voltage to start and operate under worst case design basis events. This review was also performed to determine if each component had sufficient voltage to perform its design function. The review addressed power supply, cable amp capacity, and voltage drop during all modes of operation. For MOVs, the team evaluated valve motor starting requirements to determine correct modeling in the voltage analysis. The team reviewed the electrical control schematics associated with the motors to evaluate if the control circuits had adequate voltage to start or stop the motor when required. The team also reviewed the protection provided for each of the inspection sample components and the coordination of protective devices to determine if the components were adequately protected for overcurrent conditions and the protection was selected to ensure satisfactory operation during worst-case bus voltages. The team reviewed the AC and DC bus system health reports and recent corrective action reports to determine if circuit breaker issues were being adequately resolved. Additionally, the team reviewed preventive maintenance and testing procedures to verify conformance to manufacturer recommendations. For MOVs, the team reviewed the electrical terminal voltages provided as design inputs to the mechanical torque and thrust calculations to verify the values were consistent with analyzed system conditions.
 
====b. Findings====
 
=====Introduction:=====
The inspectors identified a finding of very low safety significance (Green) involving a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensee's failure to maintain the safety-related 125V DC system design basis information consistent with the plant configuration. Specifically, a revision to the Unit 1, 
 
safety-related 125V DC system analysis (Calculation PSL-1FSE-05-002) incorporated incorrect design input specifications related to the inverter, resulting in inaccurate design basis information. The licensee's failure to maintain the vital 125V DC design basis information consistent with the plant configuration was identified as a performance deficiency.
 
=====Description:=====
The current revision of DC Calculation PSL-1FSE-05-002 did not reflect the current configuration of the Unit 1 DC system. In 2006, the licensee prepared two station modification packages to replace the existing safety-related inverters with new ones. The replacement of these components, however, did not occur as scheduled and had not occurred at the time of inspection. Based on licensee verbal information, the installation of the inverters was scheduled for 2012. The licensee issued Revision 1 of the above calculation on December 10, 2008. This revision included the proposed replacement inverter equipment specifications as design inputs. The specifications for the replacement inverters were less limiting than the presently installed inverters. In particular, the installed inverters require a minimum of 105V DC to operate and have an efficiency of 75 percent. The replacement inverters require 100V DC and have an efficiency of 81 percent. Through discussions with the licensee pertaining to the discrepancy between the current plant configuration and the 125V DC system design analysis, the inspection team determined that such discrepancies are permitted by the station's Quality Assurance Procedure ENG-QI 1.5. Specifically, Section 5.1.C of ENG-QI 1.5 states: "Calculations may be created or revised to support modifications and issued before completion of the modification. Since calculations are issued "as-engineered," when a modification is cancelled it may be necessary to revise calculations to return them to the correct configuration."  Since the QA procedure did not establish a time limit when a discrepancy was allowed to exist between the design documentation and the configuration of the plant, such discrepancy could exist for years, as in the case of the postponed replacement of the inverters. The team was concerned that the existence of official design documents that are inconsistent with the configuration of the plant might invalidate conclusions pertaining to the operability and performance of structures, systems, and components, particularly if, during the intervening period, other design changes and plant modifications were developed on the assumption that the documents of record reflect the current plant configuration. Regarding the incorrect inverter minimum voltage information, the team was concerned that degradation of the battery in subsequent years combined with the implementation of other potential modifications could result in the nuclear safety-related inverters being unable to perform their design safety function.
 
=====Analysis:=====
The licensee's failure to maintain the vital 125V DC design basis information consistent with the plant configuration was identified as a performance deficiency. This finding, associated with the Mitigating Systems Cornerstone attribute of Design Control was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern in that future design activity or operability assessments would assume the lower voltage (100V DC vs. actual 105V DC) value was acceptable for assuring the adequacy of voltage to the safety-related inverters. The team assessed this finding for significance in accordance with NRC Manual Chapter 0609, using the Phase I SDP worksheet for mitigating systems and determined that the finding was of 
 
very low safety significance (Green) since it was a design deficiency determined not to have resulted in a loss of safety function. Regarding the programmatic concern about configuration discrepancies permitted by procedure ENG-QI 1.5, the team did not identify any other design document that was inconsistent with the current plant configuration. This finding reflects current station performance because the identified performance deficiency occurred in a calculation revision dated December 10, 2008. The issue was identified to be programmatic because the station procedure for controlling engineering calculations (ENG-QI-1.5) contributed to the performance deficiency. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to ensure that procedures (i.e. ENG-QI 1.5) were available and adequate to assure nuclear safety (specifically, complete, accurate and up-to-date design documentation).  [H.2(c)]
 
=====Enforcement:=====
10 CFR 50 Appendix B, Criterion III, Design Control, requires that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Contrary to the above, design changes were not subject to design control measures commensurate with those applied to the original design in that a revision to the Unit 1, safety-related 125V DC system analysis (Calculation PSL-1FSE-05-002) incorporated incorrect design input specifications related to the system inverter equipment. As a result, the station's Unit 1, safety-related 125V DC system analysis, revised on December 10, 2008, did not reflect the actual plant configuration and was not conservative in that it concluded that a minimum voltage of 100V DC was adequate to assure operation of the safety-related inverters. Because the finding was of very low safety significance and was entered into the licensee's corrective action program (CR 2009-24517), this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy:  NCV 05000335,389/2009006-03, Failure to Maintain the Safety-Related 125V DC System Design Basis Information Consistent with the Plant Configuration.
 
===.3 Review of Low Margin Operator Actions===
 
====a. Inspection Scope====
The team performed a margin assessment and detailed review of six risk significant and time critical operator actions. Where possible, margins were determined by the review of the assumed design basis and FSAR response times. For the selected operator actions, the team performed a walkthrough of associated Emergency Operating procedures (EOPs) abnormal operating procedures (AOPs), Normal Operating Procedures (OPs), and other operations procedures with appropriate plant operators and engineers to assess operator knowledge level, adequacy of procedures, availability of special equipment when required, and the conditions under which the procedures would be performed. The inspection team conducted detailed reviews with operations and training department leadership, and observed operator training on the plant simulator, to assess the procedural rationale and approach to meeting the design basis and FSAR response and performance requirements. Operator actions were observed on the plant simulator and during plant walk downs. Additionally, the team reviewed the station configuration control for risk significant manual valves. This review included field verification that the valve positions for a selected sample of risk significant manual valves was consistent with applicable drawings and system operating procedures.
 
Operator actions associated with the following events/evolutions were reviewed:
* Reactor coolant system feed and bleed and Power Operated Relief Valve (PORV) fails open (block valve use)
* Inner-system Loss of Coolant Accident (LOCA)
* Anticipated Transient Without a Scram (ATWS) - Emergency Boration
* Cross-tie 480V 1AB load center
* Condensate storage tank makeup from the treated water storage tank
* Restoration of non-essential CCW following Safety Injection Actuation Signal (SIAS)
 
====b. Findings====
 
=====Introduction:=====
The team identified a URI related to the licensee's failure to provide adequate procedures for restoration of non-essential CCW following a SIAS. Specifically, emergency operating procedure, 1-EOP-99, Appendix A, "Sampling Steam Generators," and Appendix J, "Restoration of CCW and CBO to the RCPs," Rev. 38, did not address the potential adverse impact on essential cooling flow required to mitigate a LOCA when the non-essential CCW was restored.
 
=====Description:=====
Emergency Operating Procedure 1-EOP-99, Appendix A and J, step 2, directed the operator to restore non-essential CCW if the related isolation valve closed due to the SIAS. Additionally, an input to isolate non-essential CCW was provided by a low CCW surge tank level signal. The purpose of both signals was to assure adequate cooling flow was provided to essential loads for design basis accident conditions. The station CCW flow balance procedure (1-NOP-14.02, Rev. 20, Appendix I) positioned system flow balance valves to establish cooling flow to the essential components based on assumptions in the LOCA Containment Analysis, JPN-PSL-SENP-93-001, Rev. 0. When establishing the essential cooling flow balance per this procedure, the non-essential portion of the CCW system was isolated. Therefore, adequate essential cooling flow was assured only when the non-essential portion of the system was isolated. The EOP assured that CCW train separation was maintained when the non-essential header was restored but did not address that the essential cooling load flow would be diverted with the potential adverse impact on cooling capability for the essential components, primarily the containment coolers used in containment pressure control, the shutdown heat exchanger used for decay heat removal, and cooling for emergency core cooling system (ECCS) pumps. The team concluded that the procedure action to restore non-essential CCW flow after an SIAS signal adversely impacted the licensee's capability to assure adequate cooling of essential components following a LOCA induced SIAS. In particular, this concern applied to the circumstance of only one train of CCW being available during LOCA, assuming a single failure event resulted in the loss of the redundant train. Following identification by the team, the licensee initiated CR 2009-22623 to assess this issue. The immediate compensatory action was to issue a standing order to the operators related to Emergency Operating Procedure 1, 2-EOP-99 directing them to not restore the non-essential CCW when responding to a SIAS when only one CCW train was available. Additionally, the licensee initiated an evaluation to assess the impact on essential CCW flow if non-essential CCW was restored to allow cooling of the RCPs and 
 
the steam generator sample coolers. The licensee's failure to provide adequate procedures for restoration of non-essential CCW following a SIAS was identified as a performance deficiency. The licensee's evaluation, and the NRC review of this evaluation, is needed to determine if adequate cooling would be available to essential equipment following the LOCA induced SIAS when the non-essential CCW was restored. This issue is being documented as URI 05000335, 389/2009006-04, Inadequate Procedure for Restoration of Non-Essential CCW Flow Following a SIAS.
 
===.4 Review of Industry Operating Experience===
 
====a. Inspection Scope====
The team reviewed selected operating experience issues that had occurred at domestic and foreign nuclear facilities for applicability at the St. Lucie Nuclear Plant. The team performed an independent applicability review for issues that were identified as applicable to the St. Lucie Nuclear Plant and were selected for a detailed review. The issues that received a detailed review by the team included:
* Generic Letter 07-01, Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients.
* Generic Letter 98-02, Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition.
* NRC Information Notice 07-09, Equipment Operability Under Degraded Voltage Conditions.
* Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient, dated July 25, 1984
* NRC Information Notice 2008-02: Findings Identified During Component Design Bases Inspections, March 19, 2008
 
====b. Findings====
No findings of significance were identified.
 
===.5 Review of Permanent Plant Modifications===
 
====a. Inspection Scope====
The team reviewed one permanent modification related to the selected risk-significant components in detail to verify that the design bases, licensing bases, and performance capability of the components have not been degraded through modifications. The adequacy of design and post-modification testing of these modifications was reviewed by performing activities identified in IP 71111.17, Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications. The following modification was reviewed:
* PC/M: 04028, Medium Voltage Switchgear Circuit Breaker Replacement - Phase III
 
====b. Findings====
No findings of significance were identified.
 
{{a|4OA5}}
==4OA5 Other Activities==
CCW Air Intrusion Event
 
====a. Inspection Scope====
The team performed a detailed review of the condition reports related to the air intrusion into the CCW system event that took place from 2:13 a.m. on October 16, 2008, through 4:02 a.m. on October 17, 2008.
 
====b. Findings====
 
=====Introduction:=====
The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensee's failure to translate the design basis, as specified in the license application, into specifications, drawings, procedures, and instructions. Specifically, a non-safety system failure (i.e. containment IA compressors) could cause a common cause failure of both trains of a safety system (i.e. CCW system).
 
=====Description:=====
The Unit 1 design included IA compressors inside containment. The Unit 1 CCW system non-essential header provided cooling and seal makeup to these IA compressors. On October 16, 2008, an air intrusion event occurred in which air from the IA compressors located inside containment entered into the CCW system. The licensee determined the air intrusion into the CCW system was caused by the failures of IA system check valves V1818A and V18060 to the IA receiver tank combined with the failure of the IA unloading solenoid SE1814A. Additionally, leakage through the IA seal water cooler, which interfaces with the CCW system, created pathways for air to enter the CCW system.
 
The inspector's review of the CCW system CRs identified that the air intrusion event occurred from 2:13 a.m. on October 16, 2008 through 4:02 a.m. on October 17, 2008. The team's review identified that this event resulted in the degraded performance of both trains of the Unit 1 CCW system and a potential loss of the CCW safety function. Review of the control room operational logs, CR 2008-31947, CR 2008-34697 and, CR2008-35753 identified that both CCW pumps exhibited motor amp fluctuations due to the air intrusion. Subsequent to this, operators vented a significant amount of air from the CCW system in order to return the system parameters to normal. The air intrusion event demonstrated an original design deficiency on Unit 1 such that a non-safety system (IA) could adversely impact the reliability, capability, and availability of the safety-related CCW system. In this case, the design deficiency was a common cause failure mechanism.


In addition to the air intrusion source discussed above, the team also determined that this vulnerability potentially existed on the waste gas compressors since non-essential CCW flow was also used for waste gas compressor aftercooler cooling. The waste gas 
cc w/encl: (See page 2)
FP&L 2 cc w/encl: Richard L. Anderson Site Vice President St. Lucie Nuclear Plant Electronic Mail Distribution


compressors run at approximately 160 psig system pressure and the CCW system pressure is approximately 120 psig. The common cause failure vulnerability of the CCW system from a failure in the waste compressor units was applicable to both Unit 1 and Unit 2.
Robert J. Hughes Plant General Manager St. Lucie Nuclear Plant Electronic Mail Distribution


The CCW system essential header cools the containment fan coolers (CFCs), shutdown cooling heat exchanger, and bearing/seal coolers for the containment spray, high pressure safety injection, and low pressure safety injection pumps. The CCW trains are normally cross-connected during normal operation. The team concluded that the air intrusion affecting both CCW trains could have prevented the CCW system from delivering the flow specified by the TS Surveillance Requirement 4.6.2.1.1 (1,200 gpm to each cooling train fan unit), and reduced flow to the remaining safety-related heat exchangers below the analyzed/required values. An additional impact of the air intrusion into the CCW system was potential degradation of the safety-related heat exchangers' performance. The team concluded that given enough air introduction, the possibility existed that the heat exchangers could become fully or partially air bound (e.g., upper tube regions), thus significantly decreasing the heat transfer capability. The combined effects of the reduced flow and the reduced heat transfer could lead to the inability of the CCW system to perform the following safety-related functions:
Gene St. Pierre Vice President - Fleet Organizational Support Florida Power & Light Company Electronic Mail Distribution
* Providing adequate cooling for those safety-related components associated with containment and reactor decay heat removal during accident conditions.
* Providing adequate cooling for those safety-related components associated with achieving safe shutdown. This event simultaneously affected both redundant trains of the CCW system (i.e. introduced a common cause failure mechanism). FSAR section 9.2.2.3.2, Single Failure Analysis, states in part: "there is no single failure that could prevent the component cooling system from performing its safety function."  The licensee's evaluation of the air intrusion event failed to evaluate the operability consequences of the air intrusion on the CCW flow reduction to the safety-related heat exchangers and failed to consider the effect of the air intrusion on the heat exchangers' performance. The licensee initiated CR 2009-22929 with actions to evaluate the past operability of the CCW system during the air intrusion event.


=====Analysis:=====
Mark Hicks Operations Manager St. Lucie Nuclear Plant Electronic Mail Distribution
An original plant design deficiency was revealed by the CCW air intrusion event of October 16, 2008. This design deficiency involved the potential for a non-safety system (IA or waste gas) adversely impacting the reliability, capability, and availability of the safety-related CCW system. This design deficiency was identified as a performance deficiency. In this case, the design deficiency introduced a common cause failure mechanism. FSAR section 9.2.2.3.2, Single Failure Analysis, states, in part: "there is no single failure that could prevent the CCW system from performing its safety function."  This single failure vulnerability existed on Units 1 and 2 from potential failure of the aftercoolers on the waste gas compressors and on Unit 1 from the potential failure of the containment IA system.


The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance. It impacted the cornerstone objective because, if left uncorrected, it would affect the availability, reliability and capability of a safety system to perform its intended safety function.
Eric Katzman Licensing Manager St. Lucie Nuclear Plant Electronic Mail Distribution


Specifically, with this vulnerability, a failure of the waste gas aftercooler (either unit) or a failure of the containment IA compressors (Unit 1 only) could cause air intrusion into the CCW system and potentially lead to a loss of CCW event. A loss of CCW could result in inadequate cooling to essential equipment used to mitigate design bases accidents. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, using the Phase I and Phase II SDP worksheets for mitigating systems.
Abdy Khanpour Vice President Engineering Support Florida Power and Light Company P.O. Box 14000 Juno Beach, FL 33408-0420


It was determined that a Phase III analysis was required since this finding represented a potential loss of safety system function for multiple trains which was not addressed by the Phase II pre-solved tables/worksheets.
McHenry Cornell Director Licensing and Performance Improvement Florida Power & Light Company Electronic Mail Distribution


The preliminary Phase III analysis determined that for the air intrusion event of October 2008, it was reasonable to assume the initiating event frequency increased from the baseline by at least one magnitude and therefore the performance deficiency was preliminarily characterized as greater than Green. The preliminary Phase III analysis is attached.
Alison Brown Nuclear Licensing Florida Power & Light Company Electronic Mail Distribution


The team concluded that this finding did not have an associated cross-cutting aspect because the design of the CCW system was established in an original plant design, and therefore, was not representative of current licensee performance.
Mitch S. Ross Vice President and Associate General Counsel Florida Power & Light Company Electronic Mail Distribution


=====Enforcement:=====
Marjan Mashhadi Senior Attorney Florida Power & Light Company Electronic Mail Distribution
10 CFR 50, Appendix B, Criterion III, Design Control, requires that the design basis specified in the license application be correctly translated into specifications, drawings, procedures, and instructions. FSAR section 9.2.2.3.2, Single Failure Analysis, states in part: "there is no single failure that could prevent the component cooling system from performing its safety function."  Contrary to the above, the licensee failed to correctly translate the original design basis into specifications for the design of the CCW system. Specifically, a non-safety system failure (i.e. waste gas compressor aftercoolers, both units, or containment IA compressors, Unit 1 only) could result in a common cause failure of both trains of a safety system (i.e. CCW system). The air intrusion event revealed an original design deficiency that a non-safety system (IA) could adversely impact the reliability, capability, and availability of safety related CCW system. In this case, the design deficiency was a common cause failure mechanism. This design deficiency was established in the original plant design and has existed since the operating licenses were issued. This issue is being documented as AV 05000335, 389/2009006-05, Failure to Translate Design Basis Specifications to Prevent Single Failure of CCW.


=====Introduction:=====
William A. Passetti Chief Florida Bureau of Radiation Control Department of Health Electronic Mail Distribution
The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to identify a condition adverse to quality associated with the CCW air intrusion event that occurred in October 2008. Following the October 2008 event, the licensee failed to properly identify and correct the source of the air intrusion into the CCW system prior to closing the associated Condition Report.


The licensee's failure to identify the source (i.e. leak path from the containment IA compressors to the CCW system) of air intrusion into the CCW system was identified as a performance deficiency.
Ruben D. Almaguer Director Division of Emergency Preparedness Department of Community Affairs Electronic Mail Distribution


=====Description:=====
J. Kammel Radiological Emergency Planning Administrator Department of Public Safety Electronic Mail Distribution
The team reviewed CRs for the CCW system air intrusion event that took place from October 16, 2008 through October 17, 2008. Review of the control room operational logs, CR 2008-31947, CR 2008-34697 and, CR2008-35753 identified that both CCW pumps exhibited motor amp fluctuations due to the air in the system. Subsequent to this, operators vented a significant amount of air from the CCW pumps and heat exchangers in order to return the system parameters to normal. As discussed in section
{{a|4OA5}}
==4OA5 b.1, the licensee identified that the containment IA compressors provided a pathway for which air intrusion into the CCW system could occur.==
The team's review of the station data identified that the indicated maximum containment IA pressure was approximately 113 psig during normal operation of the compressor. The maximum identified pressure during the air intrusion event was 129 psig (CCW system pressure is approximately 120 psig). The licensee identified that the elevated IA pressure was attributed to a failure of the pressure switch that activates the unloader solenoid or the solenoid itself, such that it remained closed keeping the unit loaded and allowing header pressure to reach 129 psig. The licensee determined that the most likely path for air intrusion into the CCW system to be through the 1A containment IA compressor's aftercooler (as documented in CR 2008-34697). Listed below is a summary of actions taken by the licensee:
* Initial troubleshooting performed on November 10, 2008, under CR 2008-31947, determined that IA aftercoolers, when tested to 100 psig with compressed air, did not leak. CR 2008-31947 was subsequently closed to CR 2008-34697.
* CR 2008-34697 identified that CCW to the IA compressor aftercoolers was not needed and should remain isolated. TSA-1-08-013 was developed to accomplish this task and CR 2008-34697 was closed to CR 2008-35753.
* Subsequent troubleshooting was performed on November 18, 2008, under WO 38025447 and determined that IA aftercoolers, when tested to 120 psig with argon gas, also did not leak.
* CR 2008-35753 was closed on November 19, 2008. The closure was based on isolation of the CCW from the aftercoolers to remove the risk of compressed air entering the CCW System from this high pressure source.
* The licensee performed an operability review of the CCW system and determined the system was operable (CR 2008-31947). The corrective action documents did not provide a basis for this determination.
* The 1A compressor unloading solenoid valve body and internals were replaced on November 21, 2008 (after the event). The licensee's decision-making at the time of the event resulted in the isolation of CCW cooling to both aftercoolers. The team questioned the evaluation performed for the CCW air intrusion event which included the operability evaluation, the basis for the conclusions and the suspected air intrusion path. CR 2009-24030 was initiated to evaluate why a prompt operability determination was not requested by the licensee's operations department at the time of the event. The licensee had not performed an engineering evaluation to support the 


operability determination. Consequently, the licensee had not evaluated if the air intrusion was significant enough to block cooling flow to safety-related components during an accident. CR 2009-22929 was initiated to perform a past operability review to address this concern. The team identified to the licensee an additional air intrusion path, not previously identified by the licensee. The team concluded that the most likely source for the air intrusion was the CCW seal makeup interface with the IA compressor. The licensee issued CR 2009-25209 to address the ineffective corrective actions for the air intrusion event. The potential source of air intrusion into the CCW system from the containment IA system was re-reviewed and re-evaluated by the licensee. The licensee documented, in CR 2009-25209, that the most probable cause of the air intrusion into the CCW system was the failure of 1A IA compressor unloader solenoid (SE-18-14A) in conjunction with failure of check valves V1818A and V18060 to fully seat, which could have allowed instrument air to enter the CCW system via the make-up line.
Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power & Light Company Electronic Mail Distribution


This failure mechanism explained why leak testing of the aftercoolers and seal water cooler for containment IA compressor did not identify any leaks. The original evaluation documented in CR 2008-31947 failed to identify or address this susceptibility. As detailed above, the team's review of the troubleshooting and corrective actions documented in CR 2008-31947, CR 2008-34697, CR 2008-35753, and Work Order (WO) 38025447 determined that the licensee did not correctly identify the source of the air intrusion. This vulnerability also potentially exists on both units should the aftercoolers on the waste gas compressors fail. The waste gas compressors run at approximately 160 psig pressure and the CCW system pressure is approximately 120 psig. The team concluded that the failure of a non-safety system (i.e. containment IA or waste gas compressor) that could cause a common cause failure of both trains of a safety-related system (i.e. CCW system) was a condition adverse to quality.
Senior Resident Inspector St. Lucie Nuclear Plant U.S. Nuclear Regulatory Commission P.O. Box 6090 Jensen Beach, FL 34957-2010


The licensee initiated CR 2009-23882 to address this concern.
(Vacant)
Vice President Nuclear Plant Support Florida Power & Light Company Electronic Mail Distribution


=====Analysis:=====
Faye Outlaw County Administrator St. Lucie County Electronic Mail Distribution
The licensee's failure to identify and correct the source (i.e. leak path from the containment IA compressors to the CCW system) of air intrusion into the CCW system was identified as a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance. It impacted the cornerstone objective because it affected the availability, reliability and capability of a safety system to perform its intended safety function. Specifically, the failure to identify and correct the source of air intrusion into the CCW system affected the ability of the system to ensure that adequate cooling would be available or maintained to essential equipment used to mitigate design bases accidents. The finding was assessed for significance in accordance with NRC Manual Chapter 0609, using the Phase I and Phase II SDP worksheets for mitigating systems. It also was determined that a Phase III analysis was required since this finding represented a potential loss of safety system function for multiple trains which was not addressed by the Phase II pre-solved tables/worksheets.


The preliminary Phase III analysis determined that for the air intrusion event of October 2008, it was reasonable to assume the initiating event frequency increased from the baseline by at least one magnitude and therefore the performance deficiency was preliminarily characterized as greater than Green. The preliminary Phase III analysis is attached.
Jack Southard Director Public Safety Department St. Lucie County Electronic Mail Distribution


This finding was determined to have a cross-cutting aspect in the area of Human Performance, Decision Making, specifically, H.1(a), which states, the licensee makes safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained. The inspectors determined that the licensee's decision to close the associated corrective action documents without finding the cause of the air intrusion contributed to extending the length of time that the CCW system was susceptible to this common cause failure mode.
_
x G SUNSI REVIEW COMPLETE /RA M. Thomas/ OFFICE RII:DRS RII:DRP RII:DRS SIGNATURE RA RA RA NAME JHAMMON MSYKES BDESAI DATE 2/4/2010 2/4/2010 2/4/2010 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO Enclosure


=====Enforcement:=====
SUBJECT: PUBLIC MEETING WITH FLORIDA POWER & LIGHT FACILITY:
10 CFR 50 Appendix B Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, following the discovery of air in the CCW system on October 16, 2008, the licensee failed to identify and correct the source of the air intrusion into the CCW system and closed the associated Condition Report. As a result, the plant remained susceptible to a non-safety system failure (i.e. containment IA compressors), which could cause a common cause failure of both trains of a safety system (i.e. CCW System), for approximately one year. This issue is being documented as AV 05000335, 389/2009006-06, Failure to Identify and Correct a Condition Adverse to Quality such that a Non-Safety Related System Could Cause a Common Mode Failure of Both Trains of a Safety-Related System.
ST. LUCIE NUCLEAR PLANT DOCKETS: 05000335 and 05000389 DATE & TIME:
{{a|4OA6}}
Friday, February 19, 2010 - 2:00 p.m. to 4:00 p.m.
==4OA6 Meetings, Including Exit==
On September 4, 2009, the team presented the preliminary inspection results to Mr. Johnston and other members of the licensee's staff. Although proprietary information was reviewed as part of this inspection, all proprietary information was returned and no proprietary information is documented in the report. On October 19, 2009, the NRC presented preliminary inspection results in a telephone with Mr. Jim Porter and other members of the licensee's staff. On December 3, 2009, the NRC presented preliminary inspection results in a telephone  with Mr. Eric Katzman and other members of the licensee's staff. On December 10, 2009, the NRC presented inspection results in a telephone exit with Mr. Eric Katzman and other members of the licensee's staff.


ATTACHMENT: SUPPPLEMENTAL INFORMATION 
LOCATION:
Suite 24T25 Region II Office, Sam Nunn Atlanta Federal Center, 61 Forsyth Street, S.W., Atlanta, GA PURPOSE: The purpose of the meeting is to discuss the risk significance of the two preliminary greater than green inspection findings documented in NRC Inspection Report 05000335, 389/2009006. These findings concern apparent failures to identify and correct design issues associated with an October 8, 2008, air intrusion event from the Unit 1 containment instrument air system into the component cooling water system. The NRC's Policy Statement, "Enhancing Public Participation in NRC Meetings," effective May 28, 2002, applies to this meeting. The policy statement may be found on the NRC Web site, www.nrc.gov and contains information regarding conduct of NRC meetings.


=SUPPLEMENTAL INFORMATION=
CATEGORY:
This is a Category 1 meeting. The public is invited to observe this meeting and will have one or more opport unities to communicate with the NRC after the business portion, but before the meeting is adjourned. The NRC provides


==KEY POINTS OF CONTACT==
reasonable accommodation to individuals with disabilities, where appropriate. If you need a reasonable accommodation to participate in this meeting, or need this meeting notice or the transcript or other information from the meeting in another format (e.g., braille, large print), please notify the NRC
=s meeting contact listed below. Determinations on requests for reasonable accommodation will be made on a case-by-case basis.


===Licensee personnel===
NRC OFFICE PARTICIPATING:
:
Region II OTHER ORGANIZATIONS PARTICIPATING:
: [[contact::P. Barnes]], Mechanical Engineering Design Supervisor
None PARTICIPANTS:
: [[contact::D. Cecchett]], Licensing
NRC (partial list) L. Reyes, Regional Administrator, Region II (RII)
: [[contact::G. Johnston]], Site Vice President
V. McCree, Deputy Regional Administrator for Operations, RII K. Kennedy, Director, Division of Reactor Safety (DRS), RII L. Wert, Director, Division of Reactor Projects (DRP), RII B. Desai, Chief, Engineering Branch 1, DRS, RII W. Rogers, Senior Reactor Analyst  
: [[contact::E. Katzman]], Licensing Manager
: [[contact::D. Lany]], Operations Senior Reactor Operator 
: [[contact::J. Porter]], Manager Design Engineering
: [[contact::S. Short]], Electrical Engineering Design Supervisor 
===NRC personnel===
: [[contact::D. Jones]], Acting Chief, Engineering Branch Chief 1, Division of Reactor Safety, RII  
: [[contact::T. Hoeg]], Senior Resident Inspector, St. Lucie
: [[contact::W. Rogers]], Senior Risk Analyst, RII
: [[contact::S. Sanchez]], Resident Inspector, St. Lucie 
==LIST OF ITEMS==
OPENED, CLOSED AND DISCUSSED 


===Opened and Closed===
LICENSEE (partial list) Richard Anderson, Site Vice President Bob Hughes, Plant General Manager Eric Katzman, Licensing Manager Tom Cosgrove, Engineering Manager Abdy Khanpour, Engineering Support Jim Porter, Engineering Design Manager Cheng Guey, Risk Assessment George Tullidge, Risk Assessment 2 Enclosure MEETING CONTACT:
: 05000335,389/2009006-01 NCV Failure to Meet the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1 Requirements for the Overpressure Protection
Binoy Desai 404-562-4519 Binoy.Desai@nrc.gov AUDIO TELE-CONFERENCING Interested members of the public may participate in this meeting via a toll-free audio teleconference. Please contact the NRC meeting contact in advance of the meeting, indicating your intent to attend and to obtain teleconferencing information no later than 4:30 p.m. Wednesday, February 17, 2010.
for the CCW Surge Tank (1R21.2.2)
: 05000335,389/2009006-03 NCV Failure to Maintain the Safety-Related 125V DC System Design Basis Information Consistent with the Plant Configuration (1R21.2.20)
===Opened===
: 05000335,389/2009006-02 URI Adequacy of Performance Monitoring of the IA Compressor Emergency Cooling System. (1R21.2.3)
: 05000335, 389/2009006-04 URI Inadequate Procedure for Restoration of Non-Essential CCW Flow Following a SIAS (1R21.3)
: 05000335, 389/2009006-05 AV Failure to Translate Design Basis Specifications to Prevent Single Failure of CCW (4OA5)
: 05000335,389/2009006-06 AV Failure to Identify and Correct a Condition Adverse to Quality such that  Non-Safety Related System Could Cause a Common Mode Failure of Both Trains of a Safety-Related System (4OA5) 


==LIST OF DOCUMENTS REVIEWED==
VIDEO TELE-CONFERENCING Not Available
===Calculations===
: 128-42A.6002, Component Cooling Water (CCW) System SIAS Operation, Rev. 0,
: CRN 07127-17201
: PSL-1FJM-93-06, Intake Cooling Water System Performance, Rev. 2
: 007-AS93-C-004
: PSL-1CHN-93-002A, Unit 1 LOCA Containment Pressure/Temperature (P/T) Analysis for 102% Power (2754 MWt), Rev. 0
: NSSS-040, Component Cooling Water System, Rev. 3
: PSL-1FJI-91-006,
: FIS-14-12A, B, C, & D Setpoints, Rev. 1
: PSL-BSFM-01-014, Acceptable Corrosion Allowance on the Units 1 and 2 CCW Surge Tank for a 50 psi Design Pressure, Rev. 0
: PSL-BSFM-01-019, Component Cooling Water Surge Tank Pressure Analysis, Rev. 0 32-82-6001, HVAC System HVS 5A &5B, RAB El. 43'- 0, Heat Load Verification, Rev 0 C2-B-9, HVAC System HVS 5A &5B, RAB El. 43'- 0, Heat Load Verification, Rev 0
: JPN-PSL-SEIP-92-025, Evaluation of CE's PPS Setpoint Calculation, Rev 4
: JPN-PSL-SEMP-91-007, Safety Evaluation for RAB Electrical Equipment and Battery Room, Rev 1
: PSL-1FJE-90-013, St. Lucie Unit 1 Emergency Diesel Generator 1A and 1B Electrical Loads, Rev 6
: PSL-1FJE-90-0014, Unit 1 Battery Charger 1A, 1AA, 1B, 1BB, and 1AB Sizing, Rev 1
: PSL-1FJE-90-026, St. Lucie - Unit 1 Short Circuit, Voltage Drop and
: PSB-1 Analysis, Rev 6
: PSL-1-FJE-91-002, Instrument Inverters 1A, 1B, 1C, & 1D AC Output Loading, Rev 05
: PSL-1FSE-03-009, Unit 1 ELECTRICAL System Computer Model (ETAP) Documentation, Rev 1
: PSL-1FSE-05-002, Unit 1 125V DC System ETAP Model & Analysis, Rev 1
: PSL-1FSE-05-002, Unit 1 -125V DC System ETAP Model & Analysis, Rev. 1
: PSL-1FJI-92-035, Unit 1 Pressurizer NR Pressure Uncertainty Determination, Rev 1
: PSL-1FJI-92-047, St. Lucie Unit 1 Safety Injection Tank Pressure Setpoints, 2/7/94
: IC.0004, Safety Injection Tank Level Instrumentation, Rev 4
: PSL-1-FJE-98-001, Review of Selective Coordination for the Electrical Circuits on the St. Lucie Unit 1 Essential Equipment List, rev 5, 9/26/02
: PSL-1FSE-03-009, Unit 1 Electrical System Computer Model (ETAP) Documentation, Rev 1
: PSL-1-FJE-90-0026, Unit 1 Short Circuit, Voltage Drop and
: PSB-1 Analysis, Rev 6
: Specifications
: FLO-8770-764, Unit 1 CCW Surge Tank, original issue 10/31/71
===Procedures===
: 1-NOP-14.02, Component Cooling Water System Operation, Rev. 25 2-NOP-14.02, Component Cooling Water System Operation, Rev. 15 1-NOP-50.01A, 125V DC Bus 1A (Class 1E) Normal Operation, Rev 2 1-NOP-50.01AB, 125V DC Bus 1AB (Class 1E) Normal Operation, Rev 0A 2-GOP-403, Reactor Plant Heatup - Mode 4 to Mode 3, Rev. 32
: 1-0330020, Turbine Cooling Water System, Rev. 57C
: 1-0330030, Turbine Cooling Water System, Rev. 16A
: 1-1010030, Loss Of Instrument Air, Rev. 33a 1-EOP-99, Appendices / Figures / Tables / Data Sheets, Rev. 38
: Attachment 1-OSP-14.01A, Component Cooling Water Pump Code Run, Rev. 0B 1-OSP-14.01B, Component Cooling Water Pump Code Run, Rev. 0B 1-OSP-14.01C, Component Cooling Water Pump Code Run, Rev. 0B 1-OSP-100.01, Schedule of Periodic Tests, Checks and Calibrations Week 1, Rev. 34B
: 0-EMP-50.01, 125V DC System Battery Charger 18 Month Operability Testing, Rev 8D 0-EMP-50.01A, 125V DC Bus 1A (Class 1E) Normal Operation, Rev 2 0-EMP-50.05, Safety Battery Performance Test, Rev 4A 0-EMP-50.05, Safety Battery Performance Test, Rev 6 0-EMP-50.08, Safety Battery Emergency Load Profile Test (Service Test), Rev 11 
: 0-EMP-80.11, Votes Testing of Globe and Gate Valves, Rev 6 0-EMP-80.06, Preventative Maintenance of Limitorque MOV Actuators, Rev 17B 0-CME-50.21, Safety Related Battery Cell Charging and Replacement, Rev 1A
: 0-PMI-69-01, Anticipated Transient Without a Scram (ATWS) Functional Test, Rev 2 1-EMP-50.01, Safety Battery 18 Month Maintenance, Rev 4E 1-IMP-01.37L, Pressurizer Pressure Low Range Loop Calibration, Rev 4D 1-IMP-01.37T, Pressurizer Pressure Low Range Transmitter Calibration, Rev 2B
: 1-IMP-01.39T, Pressurizer Pressure Safety Channel Transmitter Calibration, Rev 4 1-IMP-26.14, Containment Atmosphere Process Monitor Functional and Calibration Instruction, Rev 12A 1-IMP-26.19, Component Cooling Water Process Monitor Functional and Secondary Calibration Instruction, Rev 8 1-IMP-69.01, Safeguards Group Actuation Procedure, Rev 0B 1-IMP-69.02, ESFAS Monthly Channel Functional Test, Rev 11 1-OSP-50.01, 125V DC Bus 1AB Crosstie Breaker In-place Undervoltage Testing, Rev 1B
: 1-1400052, Engineered Safeguards Actuation System, Channel Functional Test, Rev 54 1-IMP-14.02, CCW to RCP Seal Temperature Switch Calibration, Rev 3
: OP-1-0010125, Schedule of Periodic Tests, Checks and Calibrations St. Lucie Unit 1, Rev 79
: OP-1-0010125A, Surveillance Data Sheets, Rev. 125
: 1-1400064L, Installed Plant Instrumentation Calibration (Level), Rev 47 1-PTP-21, Bus 1A1 SF6 Breakers Pre-Operational Testing, Rev 0A 1-PTP-24, Bus 1B1 SF6 Breakers Pre-Operational Testing, Rev 0A
: 0310080, Preoperational Test Procedure, Component Cooling Water Functional Test, Rev. 2
: ENG-QI 1.5, Quality Instruction Nuclear Engineering Calculations, Rev 8
: IMP-76.01, Rosemount Transmitter Repair & Calibration (Model 1153 & 1154) IMG-.04, Magnetrol Level Switch Calibration, Rev 10A 
: Completed Procedures 1-OSP-14.01A, Component Cooling Water Pump Code Run, performed on: 6/12/09, 3/12/09, 12/11/08, 9/12/08, 7/7/08, 3/15/08 1-OSP-14.01B, Component Cooling Water Pump Code Run, performed on: 6/26/09, 3/27/09, 12/26/08, 9/26/08, 6/26/08, 3/27/08 1-OSP-14.01C, Component Cooling Water Pump Code Run, performed on: 6/26/09, 3/27/09, 12/26/08, 9/26/08, 6/26/08, 3/27/08 1-NOP-14.02, Component Cooling Water System Operation, Appendix I, Essential CCW Load Flow Balance, performed on: 11/18/08 2-NOP-14.02, Component Cooling Water System Operation, Appendix I, Essential CCW Load Flow Balance, performed on: 05/29/09 
: Attachment Drawings 8770-G-078, Sheet 162A, Flow Diagram, Waste Management System, Rev. 14
: 8770-G-082, Sheet 1, Flow Diagram, Circulating & Intake Cooling Water System, Rev. 50 8770-G-082, Sheet 2, Flow Diagram, Circulating & Intake Cooling Water System, Rev. 23
: 8770-G-083, Sheet 1A, Flow Diagram, Component Cooling System, Rev. 59 8770-G-083, Sheet 1B, Flow Diagram, Component Cooling System, Rev. 57 8770-G-083, Sheet 2, Flow Diagram, Component Cooling System, Rev. 4 8770-G-085, Sheet 2A, Instrument Air System, Rev. 39 8770-G-085, Sheet 4B, Instrument Air System, Rev. 31
: 8770-G-089, Sheet 1A, Flow Diagram, Turbine Cooling Water System, Rev. 26 8770-G-089, Sheet 1B, Flow Diagram, Turbine Cooling Water System, Rev. 26 8770-G-089, Sheet 2, Flow Diagram, Turbine Cooling Water System, Rev. 25 8770-G-100, Flow Diagram Symbols, Rev. 10 8770-G-125, Sheet
: CC-H-5, Large Bore Piping Isometric, Component Cooling Piping, Rev. 5 8770-G-125, Sheet
: CC-H-7, Large Bore Piping Isometric, Component Cooling Piping, Rev. 5 8770-G-862, HVAC - Air Flow Diagram, Rev. 31
: 8770-G-879, HVAC - Control Diagrams - Sheet 2, Rev. 39 8770-16336, Bettis Actuator, Spring Return, Rev. 1 8770-5624, Component Cooling Water Surge Tank, Rev. 4 8770-B-326, Sh. 269, Schematic Diagram Safety Injection Tank Isolation Valve V-3626, Rev 8
: 8770-B-327, Sh. 118, Control Wiring Diagram Pressurizer Relief Isolation Valve V-1403, Rev 16 
: 8770-B-327, Sh. 120, Control Wiring Diagram Pressurizer Relief Isolation Valve V-1405, Rev 15 8770-B-327, Sh. 140, Control Wiring Diagram Measurement Channels L-1103, L-1116, & P-1103, Rev 15 8770-B-327, Sh. 141, Control Wiring Diagram Measurement Channels
: PS-1118,
: PT-1116, &
: PT-1104, Rev 24 8770-B-327, Sh. 161, Control Wiring Diagram Volume Control Tank Discharge Valve V-2501, Rev 8 8770-B-327, Sh. 162, Control Wiring Diagram Refueling Water to Discharge Pumps V-2504, Rev 7 8770-B-327, Sh. 201, Control Wiring Diagram Component Cooling Water Pump 1A, Rev 16
: 8770-B-327, Sh. 205, Control Wiring Diagram Component Cooling Water Pump 1B, Rev 22 8770-B-327, Sh. 209, Control Wiring Diagram Component Cooling Water Pump 1C, Rev 23 8770-B-327, Sh. 211, Control Wiring Diagram Component Cool Wtr Shutdown Heat Exch & Surge Tank Fill Valves, Rev 13 8770-B-327, Sh. 250, Control Wiring Diagram Shutdown Cooling Isolation Valve V-3481,
: Rev 12 8770-B-327, Sh. 253, Control Wiring Diagram Shutdown Cooling Isolation Valve V-3651,
: Rev 15 8770-B-327, Sh. 269, Control Wiring Diagram Injection Tank 1A1 Isolation Valve V-3624, Rev 8 8770-B-327, Sh. 270, Control Wiring Diagram Injection Tank 1A2 Isolation Valve V-3614,
: Rev 13 8770-B-327, Sh. 271, Control Wiring Diagram Injection Tank 1B1 Isolation Valve V-3634,
: Rev 13 8770-B-327, Sh. 272, Control Wiring Diagram Injection Tank 1B2 Isolation Valve V-3644, Rev 8 8770-B-327, Sh. 409, Control Wiring Diagram CEA Drive MG Set 1A Pnl, Rev 9 8770-B-327, Sh. 410, Control Wiring Diagram CEA Drive MG Set 1B Pnl, Rev 8
: Attachment 8770-B-327, Sh. 476, Control Wiring Diagram Electrical Equipment Room Supply Fan
: HVS-5A, Rev 7 8770-B-327, Sh. 477, Control Wiring Diagram Electrical Equipment Room Supply Fan
: HVS-5B, Rev 7 8770-B-327, Sh. 532, Control Wiring Diagram Safeguards Room "A" Sump Pumps, Rev 9 8770-B-327, Sh. 533, Control Wiring Diagram Safeguards Room "B" Sump Pumps, Rev 10 8770-B-327, Sh. 583, Control Wiring Diagram Equipment Drain Sump Pump, Rev 6 8770-B-327, Sh. 600, Control Wiring Diagram Instrument Air Compressor Emergency Cooling System, Rev 2 8770-B-327, Sh. 934, Control Wiring Diagram 4160V Swgr 1A2 Fdr to Bus 1A3, Rev 13 8770-B-327, Sh. 935, Control Wiring Diagram 4160V Swgr 1B2 Fdr to Bus 1B3, Rev 13 8770-B-327, Sh. 978, Control Wiring Diagram 480V Switchgear 1A2 - 1AB Tie, Rev 9 8770-B-327, Sh. 979, Control Wiring Diagram 480V Switchgear 1AB - 1A2 Tie, Rev 10 8770-B-327, Sh. 980, Control Wiring Diagram 480V Switchgear 1B2 Fdr, Rev 11 8770-B-327, Sh. 1002, Control Wiring Diagram Battery 1B & Battery Charger 1B, Rev 24 8770-B-327, Sh. 1003, Control Wiring Diagram Battery Charger 1AB, Rev 16
: 8770-B-327, Sh. 1601, Control Wiring Diagram Battery Charger 1BB, Rev 6 8770-G-272, Main One Line Wiring Diagram, Rev 25 8770-G-274, Auxiliary One Line Diagram, Rev 16 8770-G-275, 6.9KV Swgr. & 4.16 KV Swgr. One Line Wiring Diagram, Rev 17 8770-G-275, Sh.2, 480V Swgr. & Pressurizer Htr. Bus One Line Wiring Diagram, Rev 20
: 8770-G-332, Sh. 1, 480V Miscellaneous, 125V DC and Vital AC One Line, Rev 23 8770-G-332, Sh. 2, 480V Miscellaneous, 125V DC and Vital AC One Line, Rev 6
: 8770-3639, CEA Drive MG Sets Elementary Connection Diagram, Rev 11 8770-5515, Sh. 3, Electrical Schematic, Safety Features Actuation System SB, Rev 20 8770-5516, Sh. 3, Electrical Schematic, Safety Features Actuation System SA, Rev 19
: 8770-5517, Sh. 1, Electrical Schematic, Safety Features Actuation System MA, Rev 14 8770-5518, Sh. 1, Electrical Schematic, Safety Features Actuation System MC, Rev 15 8770-5519, Sh. 1, Electrical Schematic, Safety Features Actuation System MB, Rev 15 8770-5520, Sh. 1, Electrical Schematic, Safety Features Actuation System MD, Rev 14 8770-12315, Sh. 2, Electrical Schematic, Safety Features Actuation System MD, Rev 0
: 8770-12316, Sh. 3, Electrical Schematic, Safety Features Actuation System MA, Rev 0 8770-12317, Sh. 2, Electrical Schematic, Safety Features Actuation System MB, Rev 0 8770-12318, Sh. 2, Electrical Schematic, Safety Features Actuation System MC, Rev 0 8770-G-083, sheet 1A, Flow Diagram Component Cooling System - Unit 1, Rev 59 8770-G-083, sheet 1B, Flow Diagram Component Cooling System - Unit 1, Rev 57
: 8770-G-227, sheet 1, Reactor Auxiliary Building Instrument Arrangement, Rev 21 8770-G-272, Unit 1 Main One Line Wiring Diagram, Rev 25 8770-G-274, Unit 1 Auxiliary One Line Diagram, Rev 17 8770-G275, sheet 1, 6.9KV Switchgear & 4.16KV Switchgear One Line Wiring Diagram, Rev 19 8770-B-327, sheet 202, Control Wiring Diagram Normal Supply Header & Normal Return Header Isolation Valves - Unit 1, Rev 6 8778-B-327, sheet 211, Component Cooling Water Shutdown Heat Exchanger & Surge Tank Fill Valves Unit 1 Control Wiring Diagrams, Rev 13 8770-B-327, sheet 280, Safety Injection Tank 1A-2 Instrument and Check Valve Leakage Drain to RWT
: HCV-3618 Unit 1 Control Wiring Diagram, Rev 19, 5/30/07 8770-B-327, sheet 353, CCW Rad Mon Channels 56 and 57 Unit 1 Control Wiring Diagram, Rev 6 8770-B-327, sheet 449, Control Wiring Diagram Process Radiation Channels 31 &32, Rev 11
: Attachment 8770-B-327, sheet 450, Control Wiring Diagram Process Radiation Channels 31 &32 & Iodine Pumping System control, Rev 5 8770-B-327, sheet 532, Safeguards Room A Sump Pumps Unit 1 Control Wiring Diagram, Rev 9 8770-B-327, sheet 533, Safeguards Room B Sump Pumps Unit 1 Control Wiring Diagram, Rev 10 8770-B-327, sheet 583, Equipment Drain Sump Pump Unit 1 Control Wiring Diagram, Rev 6 8770-B-327, sheet 906, Unit 1 Control Wiring Diagram Startup Transformer 1A-2 Breaker, Rev 14 8770-B-327, sheet 907, Unit 1 Control Wiring Diagram Startup Transformer 1B-2 Breaker, Rev 17 8770-B-327, sheet 934, Unit 1 Control Wiring Diagram 4160 Switchgear 1A2 Feeder to Bus 1A3, Rev 13 8770-B-327, sheet 935, Unit 1 Control Wiring Diagram 4160 Switchgear 1B2 Feeder to Bus 1B3, Rev 17 8770-B-327, sheet 948, Unit 1 Control Wiring Diagram 480 V Station Service Transformer 1B2 4160V Feeder Breaker, Rev 13 8770-B-327, sheet 978, 480 V Switchgear 1A2-1AB Tie, Unit 1, Rev 9 8770-B-327, sheet 979, 480 V Switchgear 1A2-1AB Tie, Unit 1, Rev 10 8770-B-327, sheet 980, Control Wiring Diagram, 480 V Switchgear 1B2 Feeder Breaker, Rev 11 2998-G-083, sheet 1, Flow Diagram Component Cooling System - Unit 2, Rev 41
: 2998-B-327, sheet 202, Control Wiring Diagram Normal Supply Header & Normal Return Header Isolation Valves - Unit 2, Rev 11 T/RCO/0711502-F1-R10, Unit 1 Main Power Distribution E-57953, 230KV Switchyard Operating Diagram, Rev 49 8770-G-078SH.131, Flow Diagram Safety Injection System, Rev 19
: 8770-G-088SH.2, Flow Diagram Containment Spray and Refueling Water Systems, Rev 52 8770-G-078SH.110, Flow Diagram Reactor Coolant System, Rev 30 8770-G-083SH.1A, Flow Diagram Component Cooling System, Rev 59 8770-G-078SH.121, Flow Diagram CVCS, Rev 39
===Condition Reports===
(CRs) 1998-1584, Unit 1 & 2 Charging Pump Surveillance Test Flow Rates Do Not Meet the Design Maximum Flow Rates 2005-1294, 1C CCW Pump Exceed The Maintenance Rule Unavailability Limit Of 200 Hours/Year/Pump 2005-2969, Letdown HX CCW Relief Was Found Lifting After 2A CCW Pump Start 2005-30300, Inter-System LOCA Detection Instrumentation for Reactor Pressure Boundary Not Prioritize for Deficiency Resolution Prior to Mode 4 2007-27048, Incorrect Safety Classification of a DBD Function for Valve
: TCV-14/4A/4B 2007-28391, Parameter Limits for ICW Operability Performance Curves 2007-35587, PMs Being Changed From Daily to Outage During the Outage 2008-31947, Air introduction into CCW System
: 2008-34697, Air introduction into CCW System per
: CR 2008-31947 2008-35753, Isolate CCW to Containment IA Compressors' Aftercoolers 2008-37070, St. Lucie Engineering Self-Assessment - Component Design Basis Inspection 2009-19025, Site Glass accidentally broken Attachment 2009-23473, CCW Surge Tank Design Basis Requirements for Code Pressure Relief Capacity and Design 2005-6815, Low Margin Issue - Degraded Grid Action Plan 2006-1885, 1A Battery Charger DC Output Breaker Lead Observed to Be Overheating
: 2006-19927, Develop PMCRs for New SF6 Breakers 2006-20023, During Performance of Breaker PM, 480V Brkr Found with Misaligned Contact 2006-20094, EDG Breaker Closure Failure during Post-Maintenance Testing of
: EDG 2006-22579, K-600 Breaker found to Have Several Problems 2006-25939, Spare Load Center Breaker Could Not Be Set per Procedure
: 2006-30383, UNUSED toc Switch Contacts Do Not Function Properly 2007-1920, 480 V Swgr Breaker Received Refurbishment by ABB in Trip Free Mode 2007-23473, EDG Loading Increase Due to Operation at Upper TS Frequency Limit 2007-4859, K3000 Breaker Refurbishment by ABB Unsatisfactory 2007-7456, 480V Swgr Breaker Failed to Trip during Testing 2007-8304, Problem Found on 480V Swgr Breaker after Refurbishment by
: ABB 2007-9889, Negative Trend in Performance of MCCBs Installed in MCCs
: 2007-10302, Hot Connection Observed on 1BB Battery Charger Neutral Lead Connection 2007-13704, Review of
: IN 2007-09, Equipment Operability Under Degraded Voltage
: Conditions. 2007-14099, 1A 125V DC System Swgr Undervoltage Relay out of Adjustment 2007-15321, 4.16KV Breaker for 1A LPSI Pump Did Not Charge Spring when Racked in
: 2007-28789, Track & Trend CR for Revision of DC Calculation due to Battery Inter-Cell Resistance. 2007-29985, Jumpering-out of two battery cells 2007-34306, Medium Voltage Breaker Cluster Finger Problem 2007-36484, 1D Battery Charger Control Board B Found to Be Defective during PM
: 2007-39837, 480V Swgr 2A3-5B Loose Breaker Power Stab 2008-14926, Adequacy of 1-OSP-50.01 to satisfy NRC Commitment to Test UV Trip
: Feature of DC Cross-Tie Breakers 2008-26139, Hard Ground on 125V DC Bus 1BDefective Masterpact Circuit Breaker Trip Unit
: 2008-33033, Defective Masterpact Circuit Breaker Trip Unit 
: 2008-35540, 1A EDG Output Breaker Failed to Open during Engineered Safeguard Test 2009-8276, Potential Part 21 Notification for ABB K-Line Circuit Breakers 2009-9055, Potential Part 21 from ABB Due to Possible Tension Spring Failure 2009-15659, 480V Swgr Breaker Had Trip Indication Illuminated 2009-15807, During the ESF Testing, the CEA MG Set Breaker Failed to Trip as Expected
: 2007-12838,
: HVS-1C field cables megger readings were identified out of spec low, 4/27/2007 2005-10351,Potentail for Motor Degradation, 4/11/2005 2008-21053, Leakage into the 1A2 Safety Injection Tank, 6/26/2008 2006-17344, V1403 Did Not Stoke Closed as Expected, 6/4/2005 2009-20554, SIT Outlet Valves V3614 and V3624 Failed to Open, 7/20/2009 2007-42630, 2A1 SIT Iso Valve, V3624 Breaker Trip, 12/26/2007 2005-1469, 2A2 SIT Iso Valve Failed to Open, 1/23/2005LOL
: 2004-9733, SIT Outlet Valve V3614 Failed to Open.
: Completed Work Orders (WOs)
: 38025447, Air Leaked into CCW - Troubleshoot, dated 11/18/08
: 31023348-01, Unit 1 Replacement of AM507 and AM517, 12/12/01
: Attachment
: 33001113-01, 125V DC Battery 1B Capacity Test, 4/9/04
: 34013031-01, Pressurizer Pressure P-1103/1104 Loop Calibration, 11/5/05
: 34013455-01, 125V DC Battery 1B Performance Maintenance, 11/22/05
: 35027963-01, 125V DC Battery 1B Capacity Test, 11/22/05
: 36001038-01,125V DC Battery 1B Battery Profile Test, 4/2/07
: 36001576-01, 125V DC Battery 1B Performance Maintenance, 4/9/07
: 36006266-01,125V DC System Battery Charger 1B 18-Month Operability Test, 7/3/07
: 36008706-01,
: PT-1103 EQ Rosemount Replacement, 4/18/07
: 36008707-01,
: PT-1104 EQ Rosemount Replacement, 4/15/07
: 37011755-01,125V DC Battery 1B Battery Profile Test, 11/13/08
: 37011878-01, 125V DC Battery 1B Performance Maintenance, 11/14/08
: 37019185-01,125V DC System Battery Charger 1B 18-Month Operability Test, 7/17/08
: 37024229-01, Pressurizer Pressure PT1103 Transmitter Calibration, 11/6/08
: 37024231-01, Pressurizer Pressure PT1104 Transmitter Calibration, 11/5/08
: 38003350-01, ATWS Functional Test, 6/6/09
: 38005460-01, Unit 1 Battery Charger 1AA Charger PM, 7/2/08
: 38008615-01, Pressurizer Pressure P-1103/1104 Loop Calibration, 12/5/08
: 38013076-01, 125V DC Battery 1B Quarterly PM, 1/6/09
: 38025268-01, 125V DC Battery 1B Quarterly PM, 3/19/09
: 39001705-01, Engineered Safeguards Monthly, 6/21/09
: 39001717-01, 125V DC Battery 1B Quarterly PM, 5/28/09
: 39003272-01, ESFAS Monthly PM, 6/9/2009 W/R
: 39005930, Replace relay 27-4, 8/6/09 W/R
: 37008352, Oil leak by north oil pump on Startup Transformer 1B, 7/30/07 W/R
: 38013946, Breaker binding when racking in or out, 11/12/08 W/O
: 34020981, Calibration of Safeguards Room B Level Alarm Switch
: LS-06-1B and High-High Alarm Level Switch
: LS-06-41 W/O
: 38005217, Calibration of Safeguards Room B Level Alarm Switch
: LS-06-1B and High-High Alarm Level Switch
: LS-06-41 W/O
: 34020329, Calibration of Safeguards Room A Level Alarm Switch
: LS-06-1A and High-High Alarm Level Switch
: LS-06-40 W/O
: 38011496, Channel 31 and 21 18 month Calibration, 7/9/08 W/O
: 37017925,
: RE 26-56 & 57 Calibration, 8/16/07 W/O
: 32013060, Spare Breaker PM, 12/18/03 W/O
: 33016593, Breaker 1B2-7B 54 Month PM, 9/10/04 W/O
: 33022130, Breaker 1B2-2C 54 Month PM, 7/27/04
: W/O
: 34018962, Breaker 1A1-7B 54 Month PM, 3/9/06 W/O
: 35002739, 4/16KV SWGR 1B3-2 Breaker Replacement and Testing , 6/3/05 W/O
: 35009733, Breaker 1A1-5D PM and Swap, 7/15/05 W/O
: 35020492, Breaker 1B2-6B 54 Month PM, 2/13/06 W/O
: 36008492, Breaker 1B2-7A PM and Swap, 10/04/06
: WO31022173-01, V3106 Check Valve Inspection
: WO31022495-01, V07174 IST Check Valve Inspection
: WO33003927-01, V07172 IST Check Valve Inspection
: WO34019438-01, V07174 IST Check Valve Inspection
: WO36000672-01, V07172 IST Check Valve Inspection
: WO37015831-01, V07174 IST Check Valve Inspection
: WO38018501-01, V07174 IST Check Valve Inspection Attachment Modifications Change Request Notice
: CRN 03167-13230, Vendor Manual for Shutdown Cooling Heat Exchanger Update to Show the Correct Tube Plugs, Rev. 0 Change Request Notice
: CRN 18362, Install Temporary Protection on
: LG-14-2A and
: LG-14-2B, Rev. 0 Change Request Notice
: CRN 00048-9446, Permanent Removal of Gravity Damper Cover Plates on
: GD-1 and
: GD-2, Rev. 0
===Miscellaneous Documents===
: DBD-CCW-1, Component Cooling Water System, Rev. 2
: DBD-ICW-1, Intake Cooling Water System, Rev. 2
: DBD-HVAC-1, Safety Related HVAC Systems, Rev. 2
: DBD-120V-AC-1, Class 1E 120 V AC Power System, Rev 2
: DBD-480V-AC-1, 480 VAC Distribution System, Rev 2A
: DBD-4160
: VAC-1, 4160 VAC Distribution System, Rev 2
: DBD-EDG-1, Emergency Diesel Generatoor System, Rev 3
: DBD-ESF-1, Engineered Safety Features Actuation System, Rev 2
: DBD-HVAC-1, Safety Related HVAC Systems Design Basis Document, Rev 2
: DBD-PZR-1, Pressurizer System, Rev 2
: DBD-VDC-1, Class 1E DC Electrical Distribution System, Rev 2 8770-5756, Component Cooling Water Pump, Rev. 6
: 8770-7248, I/M Centrifugal Fans
: HVS-4A, 4B, 5A, 5B,
: HVE-1, 2, 4, 5, 7A, 7B, 8A, 8B, 9A, 9B, 10A, 10B, 15, 16A, 16B, 21A, 21B, 13A, 13B, 14 & 33, Rev. 5
: 0711209, Component Cooling Water System, Rev. 12
: 0702209, Component Cooling Water System, Rev. 8 Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient, Dated July 25, 1984
: EPO-84-1662, CCW Surge Tank Overpressurization, Dated, August 20, 1984
: SLN-88-021-10-20,
: JPN-PSL-SEICP-92-28, Evaluation of the Design basis for Fisher & Porter Indicating Controllers for Temperature Control Valves
: TCV-14-4A and
: TCV-14-4B.
: JPN-PSL-SENP-93-001, Inputs for the LOCA Containment Re-Analysis, Rev. 0
: NRC
: NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, April 1995 NRC Information Notice 2008-02: Findings Identified During Component Design Bases Inspections, March 19, 2008
: FPL-09-366, Westinghouse Letter to Phil Barnes, CCW Flowrates Used in Containment Analysis for St. Lucie, September 2, 2009
: JPN-PSL-SEMP-91-007, Safety Evaluation for RAB Electrical Equipment and Battery Room HVAC, St. Lucie Unit 1, Rev 1
: 0711401, Engineered Safety Features Actuation System, Rev 1 00809-0100-4388, Rosemount 1153 Series D Alphaline Nuclear Pressure Transmitter, Rev
: BA 2998-15662, Instruction Manual for Safety Features Actuation System (SFAS) Vol. 1 & Vol. 2, Manual No. TM9N38, Rev 8 8770-7423, Instruction Manual Turbine Building Battery Charger, C&D Battery, Manual No.
: MCB-2010, Rev 5
: 8770-10459, Instruction Manual for Battery Chargers 1AA, 1BB, and 1D, C&D Battery, Manual No.
: RS-421, Rev 5
: Attachment 8770-15662, Instruction Manual for Safety Features Actuation System (SFAS) Vol 1 & Vol. 2, Manual No. TM9N38, Rev 8 Unit 1 System 47, 480 VAC System Health Report, 6/30/2009 Unit 1 System 50, 125V DC System Health Report, 6/30/2009 
: Unit 1 System 52, 4.16 KV System Health Report, 6/30/2009 Unit 1 System 63, Reactor Protection System Health Report, 6/30/09 IEEE Standard 450-1995, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid for Stationary Applications. Vendor Manual 8770-15227, OTEK
: HI-Q2000 Instruction Manual, Rev 1, 5/11/.06
: Vendor Manual 8770-12474, Beckman Industrial Model 500T Digital Panel Indicator Operator's Manual, Rev 0, 2/13/91 L-2007-067, Response to Generic Letter 2007-01, 5/8/2007 Component Evaluation Sheet, page 2, Rosemount 1153-GD-7 Series D Pressure Transmitter, Rev 11 Maintenance Rule Scoping for Switchyard System, Rev 3 Maintenance Rule Scoping for 480V Switchgear, Breakers and MCC's Nuclear Plant Switchyard Inspection Report (weekly) for breakers 8W23, 8W40, 8W61 and 8W64 Control Room Log, 7/19-21/2009
: CRs  and WOs Initiated Due to CDBI Activity: 2009-22430, Inadequate housekeeping in the PSL1 CCW Surge Tank Room
: 2009-22556, Lid on Head Tank for Instrument Air Compressor Cooling Water Fan Cooler Was Rusted Shut 2009-22623, Alignment of the Non-Essential CCW Header to the Only Remaining Essential CCW Header Under Certain Accident Scenarios Could Potentially Place the Plant in the Unanalyzed Condition 2009-22766, Instrument Air Compressor Cooling Water Cooler Original Design Documentation Can Not Be Located 2009-22811, Two Steel Angles (Support on HVAC duct) Extend Down Into The Walk Path Around The East Side Of The Surge Tank In The Unit 1 CCW Surge Tank Room 2009-22892, 1(2)A and 1(2)B Instrument Air Compressor Emergency Cooling Lineup Issues 2009-22929, A NRC inspector for the CBDI team has questioned the operability determination previously done for Air Intrusion into CCW Event from October, 2008 2009-22959, Missing Information from Calculation
: PSL-1CHN-93-002, Rev. 0 about 3 Plugged Tubes in the 1A Shutdown Cooling Heat Exchanger 2009-23011, CCW Pump IST Procedure (1-OSP-14.01A/B/C) does not address affect (sic) of pump degradation on SIAS CCW System flow rates 2009-23473, CCW surge tank design basis requirements for code pressure relief capacity and design 2009-23882, Investigate possible sources of air ingress into the CCW System on PSL1 and PSL2 2009-24030, A Past Operability Review of an Air Intrusion into CCW event from October 2008
: 2009-25209, Evaluation of the Air/Gas Intrusion into CCW event from 10/16/08 which was documented in 3/C
: CR 2008-34697 2009-25276, Unit 1 CCW Surge Tank Overpressure Protection configuration is not in compliance with the ASME Code Attachment 2009-17349, Calculation
: PSL-1-FSE-002, Rev. 1, Transmitted to Document Control as the Calculation of Record Without an FPL Acceptance Signature
: 2009-22338, Loose Tools Identified by NRC at 2A4/2B4 Switchgear and 1AB Load Center. 2009-22998, Technical Specification Battery Inter-Cell Connection Resistance Limit of 150 Micro-Ohms not Used in DC System Analysis. 2009-22999, Possible Calculation Procedure Enhancement. 2009-24649, Current Revision of Calculation
: PSL-1FSE-05-002 Does not Reflect the As-Built status of Unit 1. 2009-25088, During
: SL1-22 Both Narrow Range Pressurizer Pressure Transmitters Found out of Calibration High. 2009-25178, Battery Profile (Service) Test Procedure Enhancement 2009-22338, Loose Tools Identified by NRC at 2A4/2B4 Switchgear and 1AB Load Center, 8/5/2009
: PHASE III ANALYSIS
: SRA Analysis Number:
: STL0904 Analysis Type:
: SDP Phase III
: Inspection Report:
: 05000335, 389/2009006
: Plant Name:
: St. Lucie
: Unit Numbers:
: 1 & 2 Enforcement Action
: EA-09-321
: BACKGROUND - Air intrusion into the CCW system occurred on October 16, 2008, and was originally documented in
: CR 2008-31947.
: This air intrusion event on Unit 1 affected the CCW system to the extent that both operating CCW pumps, one in each train, were cavitating as evidenced by fluctuating amp indication.
: It was identified that the containment instrument air compressors provided a pathway for which air intrusion into the system occurred.
: This vulnerability also exists, on both units, should the aftercoolers on the waste gas compressors fail.
: The waste gas compressors run at approximately 160 psig and the CCW system pressure is approximately 120 psig.
: Original design deficiency: Non-safety related instrument air compressor inside containment (Unit 1 only) and waste gas air compressor (both units) provide a common vulnerability for safety related component cooling water (CCW) system.
: FSAR
section 9.2.2.3.2, Single Failure Analysis for the CCW system, states in part: "there is no single failure that could prevent the component cooling system from performing its safety function."
: Therefore, the air intrusion that affected both trains of the CCW system was a significant condition adverse to quality.
: PERFORMANCE DEFICIENCY - Section 4OA5 of the report discusses the air intrusion in detail.
: The air intrusion potentially rendered both trains of the safety-related CCW system inoperable.
: Two performance deficiencies were identified associated with this issue.
: The first performance deficiency involved a common cause failure vulnerability of the CCW system.
: Specifically, a non-safety system failure could result in a common cause failure of both trains of the CCW system.
: The second performance deficiency involved the failure to identify and correct a condition adverse to quality.
: Specifically, the licensee failed to properly determine the source of the air in-leakage into the CCW system and take appropriate corrective actions following the air intrusion event that occurred in October 2008.
: Further, the licensee's corrective action evaluation did not identify the common cause failure vulnerability discussed in the first performance deficiency.
: EXPOSURE TIME - One year will be used.
: DATE OF OCCURRENCE - October 2008
: SAFETY IMPACT > Green
: RISK ANALYSIS/CONSIDERATIONS
: Assumptions
: 1.
: The performance deficiency will be modeled as an increase in the probability of an initiating event, Loss of the CCW system.
: Attachment 2.
: With respect to Unit 1 the performance deficiency caused a failure or an imminent failure of the CCW system.
: Given the condition of the pumps and the surge tank level perturbations, the probability of failure will be set at 1.0 for the one year exposure time.
: 3.
: Given the response of the operators to the abnormal condition of the CCW system, recovery credit is appropriate.
: A 0.1 failure probability will be assigned to operators failing to recognize and mitigate the air intrusion before air binding of the pumps happens.
: 4.
: With respect to Unit 2 a non-conforming case initiating event frequency will be set at 1/55
years.
: This is based upon the number of years that Unit 1 and 2 have been in service since their operating licenses were issued.
: Recovery will be applied here also.
: 5.
: No recovery will be considered after air intrusion severe enough to cause CCW pump failure.
: 6.
: The non-conforming case will be considered the delta core damage frequency case.
: This is due to at least a magnitude shift in the core damage frequency results between the non-
conforming and conforming cases.
: PRA Model used for basis of the risk analysis: Licensee's full scope model
: Significant Influence Factor(s) [if any]:
: How severe the air intrusion was on the CCW system's ability to perform its numerous risk significant functions.
: CALCULATIONS 
: The top 10,000 cutsets from the full scope model were screened for a loss of CCW system initiator.
: A loss of Train A Surge Tank and Train B in test and maintenance was selected.
: Those cutsets with these events were extracted and are shown in Appendix 2.
: Once the initiating event is removed, only one basic event remained in the accident sequence, operators fail to trip the operating Reactor Coolant Pumps.
: This basic event failure probability was 3.3E-3 and represents the conditional core damage probability given a Loss of CCW.
: This CCDP was comparable to SPAR in the GEM mode.
: Applying the Unit 1 non-conforming case initiating event frequency of 1.0 yields a core damage frequency of 3.3E-3 for the exposure period.
: Applying the non-recovery term (see Attachment 3 for its detailed development) of 0.1 yields a core damage frequency of 3.3E-4 for the exposure period.
: Applying the Unit 2 non-conforming case initiating event frequency of 1.8E-2/yr to the CCDP of 3.3E-3 yields a core damage frequency of 6E-5.
: Applying the non-recovery term of 0.1 yields a final core damage frequency of 6E-6 for the exposure period.
: EXTERNAL EVENTS CONSIDERATIONS - Due to the nature of the performance deficiency which increases the frequency of an internal events initiator, external events consideration is not warranted.
: LARGE EARLY RELEASE FREQUENCY IMPACT - Since there is not an increase in SGTR or ISLOCA accident sequences, LERF is not the appropriate decision making metric.
: Attachment RECONCILIATION BETWEEN PHASE III AND PLANT NOTEBOOK/ PHASE II RESULTS - The dominant accident sequence from the Phase II Notebook is Loss of CCW followed by operators failing to trip the RCP leading directly to a large seal LOCA and core damage.
: The sequence is assigned a nominal value of 6 - four for the initiating event frequency and two for the operator error.
: Phase III results in a lower probability of operators tripping the RCP of 3E-3.
: Therefore, the color is the same in both phases but, numerically a magnitude higher in the Phase II result.
: This shows reconciliation between the two phases.
: CONCLUSIONS/RECOMMENDATIONS - Given the present information associated with the air intrusion of October 2008, it is reasonable to assume the initiating event frequency increased by at least a magnitude.
: Such a shift with recovery is in the White zone of safety characterization.
: Assuming that CCW was in imminent failure the safety characterization shifts into the red zone, even with recovery.
: Therefore, this performance deficiency should be preliminarily characterized as >Green with the intent to acquire as much information about air intrusion into the CCW system as:  * an initiator for Loss of the CCW system  * an undetected failure mechanism of any CCW functions while the equipment is in standby       
: APPENDICES: 1.
: Full Scope Model Output
: 2.
: Recovery Development Analyst:
: W. Rogers
: Date: 10/30/09
: Reviewed By:
: G. MacDonald Date: 11/02/2009 
: Appendix 1 EDITED FULL SCOPE MODEL CUTSETS FOR TOTAL LOSS OF COMPONENT COOLING WATER SYSTEM
: TOP 10,000 Cutsets for PSL1
: C:\CAFTA32\06R0B\PSL1\Cuts-B\PSL1Top10K.CUT    # Cutset Prob Event Prob EventDescription
: 4832 8.06E-11 1.00E+00 %ZZCCWU1 LOSS OF CCW IE
: 1.00E+00 CHFPRCPTRP FAILURE TO TRIP REACTOR COOLANT PUMPS FOLLOWING LOSS OF COMPONENT COOLING WATER
: 3.50E-06 CTKJ1STAIE CCW SURGE TANK RUPTURE FAILS TRAIN A (1 YR EXPOSURE)
: 6.97E-03 CTM1CCWHXB CCW HX B IN TEST OR MAINTENANCE
: 1.00E+00 RCPSL RCP SEAL LOCA FLAG EVENT
: 3.30E-03 ZHFPRCPTRP FAILURE TO TRIP RCPS LOSS OF CCW
: EDITED FOR LOSS OF COMPONENT COOLING WATER
: 4832 3.30E-03 1.00E+00 %ZZCCWU1 LOSS OF CCW IE
: 1.00E+00 CHFPRCPTRP FAILURE TO TRIP REACTOR COOLANT PUMPS FOLLOWING LOSS OF COMPONENT COOLING WATER
: 1.00E+00 CTKJ1STAIE CCW SURGE TANK RUPTURE FAILS TRAIN A (1 YR EXPOSURE)
: 1.00E+00 CTM1CCWHXB CCW HX B IN TEST OR MAINTENANCE
: 1.00E+00 RCPSL RCP SEAL LOCA FLAG EVENT
: 3.30E-03 ZHFPRCPTRP FAILURE TO TRIP RCPS LOSS OF CCW
: Report Summary:
: Filename: C:\CAFTA32\06R0B\PSL1\Cuts-B\PSL1Top10K.CUT
: Print date: 7/14/2009 2:28 PM
: Not sorted
: Printed in full Appendix 2 RECOVERY DEVELOPMENT
: Two perspectives will be applied to the recovery development, since the time variable could be applied differently.
: The more liberal of the two calculations will be applied in the quantification.
: DIAGNOSIS
: Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging
: BASE 1.0E-02
: TIME 1.0E+01 limited information available as to how much time was left prior to sys failure
: STRESS 2.0E+00 Unusual condition
: COMPLEXITY 1.0E+00 Nominal
: EXPERIENCE/TRAIN 1.0E+00 Nominal
: PROCEDURES 1.0E+00 Nominal
: ERGONOMICS 1.0E+00 Nominal
: FIT FOR DUTY 1.0E+00 Nominal
: WORK PROCESS 1.0E+00 Nominal
: DIAGNOSITIC TOTAL 2.0E-01
: ACTION
: BASE 1.0E-03
: TIME 1.0E+00 although limited information available time penalty applied to diagnosis
: STRESS 2.0E+00
: COMPLEXITY 5.0E+00 numerous actions with multiple sub-tasks outside Main Control Room
: EXPERIENCE/TRAIN 1.0E+00 Nominal
: PROCEDURES 1.0E+00 Nominal
: ERGONOMICS 1.0E+00 Nominal
: FIT FOR DUTY 1.0E+00 Nominal
: WORK PROCESS 1.0E+00 Nominal
: ACTION TOTAL 1.0E-02
: TOTAL 2.1E-01           
: Appendix 2
: DIAGNOSIS
: Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging
: BASE 1.0E-02
: TIME 1.0E+00 limited information available as to how much time was left prior to sys failure
: STRESS 2.0E+00 Unusual condition
: COMPLEXITY 1.0E+00 Nominal
: EXPERIENCE/TRAIN 1.0E+00 Nominal
: PROCEDURES 1.0E+00 Nominal
: ERGONOMICS 1.0E+00 Nominal
: FIT FOR DUTY 1.0E+00 Nominal
: WORK PROCESS 1.0E+00 Nominal
: DIAGNOSITIC TOTAL 2.0E-02
: ACTION
: BASE 1.0E-03
: TIME 1.0E+01 apply time penalty that after diagnosis, time available = time req'd
: STRESS 2.0E+00
: COMPLEXITY 5.0E+00 numerous actions with multiple sub-tasks outside Main Control Room
: EXPERIENCE/TRAIN 1.0E+00 Nominal
: PROCEDURES 1.0E+00 Nominal
: ERGONOMICS 1.0E+00 Nominal
: FIT FOR DUTY 1.0E+00 Nominal
: WORK PROCESS 1.0E+00 Nominal
: ACTION TOTAL 1.0E-01
: TOTAL 1.2E-01
}}
}}

Revision as of 04:38, 24 August 2018

02/19/2010 Notice of Category 1 Public Meeting with Florida Power and Light Company to Discuss Risk Significance of the Two Preliminary Greater than Green Inspection Report 05000335-09-006 and 05000389-09-006
ML100351155
Person / Time
Site: Saint Lucie  NextEra Energy icon.png
Issue date: 02/04/2010
From: Desai B B
NRC/RGN-II/DRS/EB1
To: Nazar M
Florida Power & Light Co
References
IR-09-006
Download: ML100351155 (5)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II SAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET, SW, SUITE 23T85 ATLANTA, GEORGIA 30303-8931 February 4, 2010 Mr. Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power and Light Company P.O.Box 14000 Juno Beach, FL 33408-0420

SUBJECT: MEETING ANNOUNCEMENT - PUBLIC MEETING CATEGORY 1 REGULATORY CONFERENCE - ST. LUCIE NUCLEAR PLANT DOCKET NOS. 50-335 AND 50-389

Dear Mr. Nazar:

This letter confirms the telephone conversation between Mr. Eric Katzman of your staff and Mr. Binoy Desai of the NRC, on February 1, 2010, concerning a Regulatory Conference, being conducted at your request, which has been scheduled for February 19, 2010, from 2 p.m. to 4:00 p.m. (EST). The meeting notice, which provides additional details, is enclosed. The purpose of the Regulatory Conference is to discuss the risk significance of the two preliminary greater than green inspection findings documented in NRC Inspection Report 05000335, 389/2009006. These findings concern apparent failures to identify and correct design issues associated with an October 8, 2008, air intrusion event from the Unit 1 containment instrument air system into the component cooling water system.

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 ins pection in the NRC Public Document Room (PDR) or from the Publicly Available Records (PARS) component of NRC's Agencywide Document Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Should you have any questions concerning this meeting, please contact me at (404) 562-4519.

Sincerely,/RA: Original signed by Mark Franke for/

Binoy B. Desai, Chief Engineering Branch 1 Division of Reactor Safety

Docket Nos.: 50-335, 50-389 License Nos.: DPR-67, NPF-16

Enclosure:

As stated

cc w/encl: (See page 2)

FP&L 2 cc w/encl: Richard L. Anderson Site Vice President St. Lucie Nuclear Plant Electronic Mail Distribution

Robert J. Hughes Plant General Manager St. Lucie Nuclear Plant Electronic Mail Distribution

Gene St. Pierre Vice President - Fleet Organizational Support Florida Power & Light Company Electronic Mail Distribution

Mark Hicks Operations Manager St. Lucie Nuclear Plant Electronic Mail Distribution

Eric Katzman Licensing Manager St. Lucie Nuclear Plant Electronic Mail Distribution

Abdy Khanpour Vice President Engineering Support Florida Power and Light Company P.O. Box 14000 Juno Beach, FL 33408-0420

McHenry Cornell Director Licensing and Performance Improvement Florida Power & Light Company Electronic Mail Distribution

Alison Brown Nuclear Licensing Florida Power & Light Company Electronic Mail Distribution

Mitch S. Ross Vice President and Associate General Counsel Florida Power & Light Company Electronic Mail Distribution

Marjan Mashhadi Senior Attorney Florida Power & Light Company Electronic Mail Distribution

William A. Passetti Chief Florida Bureau of Radiation Control Department of Health Electronic Mail Distribution

Ruben D. Almaguer Director Division of Emergency Preparedness Department of Community Affairs Electronic Mail Distribution

J. Kammel Radiological Emergency Planning Administrator Department of Public Safety Electronic Mail Distribution

Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power & Light Company Electronic Mail Distribution

Senior Resident Inspector St. Lucie Nuclear Plant U.S. Nuclear Regulatory Commission P.O. Box 6090 Jensen Beach, FL 34957-2010

(Vacant)

Vice President Nuclear Plant Support Florida Power & Light Company Electronic Mail Distribution

Faye Outlaw County Administrator St. Lucie County Electronic Mail Distribution

Jack Southard Director Public Safety Department St. Lucie County Electronic Mail Distribution

_

x G SUNSI REVIEW COMPLETE /RA M. Thomas/ OFFICE RII:DRS RII:DRP RII:DRS SIGNATURE RA RA RA NAME JHAMMON MSYKES BDESAI DATE 2/4/2010 2/4/2010 2/4/2010 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO Enclosure

SUBJECT: PUBLIC MEETING WITH FLORIDA POWER & LIGHT FACILITY:

ST. LUCIE NUCLEAR PLANT DOCKETS: 05000335 and 05000389 DATE & TIME:

Friday, February 19, 2010 - 2:00 p.m. to 4:00 p.m.

LOCATION:

Suite 24T25 Region II Office, Sam Nunn Atlanta Federal Center, 61 Forsyth Street, S.W., Atlanta, GA PURPOSE: The purpose of the meeting is to discuss the risk significance of the two preliminary greater than green inspection findings documented in NRC Inspection Report 05000335, 389/2009006. These findings concern apparent failures to identify and correct design issues associated with an October 8, 2008, air intrusion event from the Unit 1 containment instrument air system into the component cooling water system. The NRC's Policy Statement, "Enhancing Public Participation in NRC Meetings," effective May 28, 2002, applies to this meeting. The policy statement may be found on the NRC Web site, www.nrc.gov and contains information regarding conduct of NRC meetings.

CATEGORY:

This is a Category 1 meeting. The public is invited to observe this meeting and will have one or more opport unities to communicate with the NRC after the business portion, but before the meeting is adjourned. The NRC provides

reasonable accommodation to individuals with disabilities, where appropriate. If you need a reasonable accommodation to participate in this meeting, or need this meeting notice or the transcript or other information from the meeting in another format (e.g., braille, large print), please notify the NRC

=s meeting contact listed below. Determinations on requests for reasonable accommodation will be made on a case-by-case basis.

NRC OFFICE PARTICIPATING:

Region II OTHER ORGANIZATIONS PARTICIPATING:

None PARTICIPANTS:

NRC (partial list) L. Reyes, Regional Administrator, Region II (RII)

V. McCree, Deputy Regional Administrator for Operations, RII K. Kennedy, Director, Division of Reactor Safety (DRS), RII L. Wert, Director, Division of Reactor Projects (DRP), RII B. Desai, Chief, Engineering Branch 1, DRS, RII W. Rogers, Senior Reactor Analyst

LICENSEE (partial list) Richard Anderson, Site Vice President Bob Hughes, Plant General Manager Eric Katzman, Licensing Manager Tom Cosgrove, Engineering Manager Abdy Khanpour, Engineering Support Jim Porter, Engineering Design Manager Cheng Guey, Risk Assessment George Tullidge, Risk Assessment 2 Enclosure MEETING CONTACT:

Binoy Desai 404-562-4519 Binoy.Desai@nrc.gov AUDIO TELE-CONFERENCING Interested members of the public may participate in this meeting via a toll-free audio teleconference. Please contact the NRC meeting contact in advance of the meeting, indicating your intent to attend and to obtain teleconferencing information no later than 4:30 p.m. Wednesday, February 17, 2010.

VIDEO TELE-CONFERENCING Not Available

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