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{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II SAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET, SW, SUITE 23T85 ATLANTA, GEORGIA 30303-8931
{{#Wiki_filter:UNITED STATES
  January 19, 2010  
                              NUCLEAR REGULATORY COMMISSION
 
                                              REGION II
EA-09-321  
                                SAM NUNN ATLANTA FEDERAL CENTER
Mr. Mano Nazar Executive Vice President and     Chief Nuclear Officer  
                                61 FORSYTH STREET, SW, SUITE 23T85
Florida Power & Light Company P.O. Box 14000 Juno Beach, FL 33408-0420  
                                    ATLANTA, GEORGIA 30303-8931
 
                                          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  
EA-09-321
Dear Mr. Nazar:  
Mr. Mano Nazar
On September 4, 2009, U. S. Nuclear Regulatory Commission (NRC) completed an inspection  
Executive Vice President and
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.  
  Chief Nuclear Officer
The inspection examined activities conducted under your license as they relate to safety and  
Florida Power & Light Company
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.  
P.O. Box 14000
Section 4OA5 of the enclosed report discusses an event which occurred in October 2008 when  
Juno Beach, FL 33408-0420
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  
SUBJECT:       ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES
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  
                INSPECTION - INSPECTION REPORT 05000335/2009006 AND
performance deficiency.
                05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS
   
Dear Mr. Nazar:
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
On September 4, 2009, U. S. Nuclear Regulatory Commission (NRC) completed an inspection
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.   
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 Commissions 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 licensees corrective
action evaluation did not identify the common cause failure vulnerability discussed in the first
performance deficiency.


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  
FP&L                                          2
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.
The findings associated with the common cause vulnerability and the inadequate corrective
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  
actions were assessed based on the best available information. The two issues were
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
preliminarily determined to be greater than Green findings using influencing assumptions and
=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  
the Significant Determination Process (SDP). The SDP analysis determined that the two
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.
findings are potentially greater than very low safety significance because they potentially
Please contact Mr. Steve Rose at (404) 562-4609 or Mr. Binoy Desai at (404) 562-4519 within  
impacted the availability and thus the accident mitigation capability of the CCW system. These
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.
findings do not represent a current safety concern because the containment IA system has been
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  
isolated from the CCW system. Additionally, increased station sensitivity exists for recognizing
the AVs violations may change as a result of further NRC review.
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
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
(AVs). The first performance deficiency is an AV of 10 CFR 50, Appendix B, Criterion III,
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
Design Control, for the failure to translate the design basis as specified in the license
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.
application, into specifications, drawings, procedures, and instructions resulting in the CCW
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
system being susceptible to a common cause failure. The second performance deficiency is an
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).
AV of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for the failure to identify and
Sincerely,        /RA/  Kriss M. Kennedy, Director 
correct a condition adverse to quality following the air intrusion event into the CCW system that
Division of Reactor Safety
occurred in October 2008. These AVs are being considered for escalated enforcement action in
Enclosure: Inspection Report 05000335/2009006, 05000389/2009006      w/Attachment:  Supplemental Information
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.
Docket Nos.: 50-335, 50-389 License Nos.: DPR-67 and NPF-16
In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our
cc w/encl:  (See page 4)
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.


_________________________  xx
FP&L                                          3
G  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 N
In addition, this report documents two NRC-identified findings of very low safety significance
OOFFICE RII:DRS RII:OE      SIGNATURE RA RA      NAME BDESAI CEVANS      DATE 1/11/2010 1/13/2010      E-MAIL COPY?    YES NO YES          NO     
which were determined to be violations of NRC requirements. The NRC is treating these two
FP&L 4  cc w/encl: Richard L. Anderson Site Vice President St. Lucie Nuclear Plant  
violations as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement
Electronic Mail Distribution
Policy because of their very low safety significance and because they were entered into your
Robert J. Hughes Plant General Manager St. Lucie Nuclear Plant  
corrective action program. If you contest these NCVs, you should provide a response within 30
Electronic Mail Distribution
days of the date of this inspection report, with the basis for your denial, to the Nuclear
Mark Hicks Operations Manager St. Lucie Nuclear Plant Electronic Mail Distribution
Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with
copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United
Rajiv S. Kundalkar Vice President - Fleet Organizational Support Florida Power & Light Company Electronic Mail Distribution
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 NRCs 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
NRCs 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: Inspection Report 05000335/2009006, 05000389/2009006
                w/Attachment: Supplemental Information
Docket Nos.: 50-335, 50-389
License Nos.: DPR-67 and NPF-16
cc w/encl: (See page 4)


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


  Abdy Khanpour Vice President Engineering Support Florida Power and Light Company  
_________________________                        xxG SUNSI REVIEW COMPLETE
P.O. Box 14000 Juno Beach, FL   33408-0420  
  OFFICE              RII:DRS          RII:DRS          RII:DRS          RII:DRP        CONTRACTOR      CONTRACTOR      RII:DRP
McHenry Cornell Director  
SIGNATURE          RA              RA              RA                RA              RA              RA              RA
Licensing and Performance Improvement Florida Power & Light Company Electronic Mail Distribution  
NAME                SROSE            RMOORE          JHAMMAN          RTAYLOR        MSHYLAMBERG NDELIAGRECA        MSYKES
Alison Brown Nuclear Licensing Florida Power & Light Company  
DATE                  11/30/2009      11/19/2009        1/12/2010        11/20/2009      11/18/2009      11/5/2009    1/13/2010
Electronic Mail Distribution  
E-MAIL COPY?          YES        NO  YES          NO  YES          NO  YES        NO  YES        NO  YES        NO  YES      NO
Faye Outlaw County Administrator St. Lucie County
OFFICE              RII:DRS          RII:OE
Electronic Mail Distribution Mitch S. Ross Vice President and Associate General
SIGNATURE          RA              RA
Counsel Florida Power & Light Company  
NAME                BDESAI          CEVANS
Electronic Mail Distribution  
DATE                    1/11/2010        1/13/2010
Marjan Mashhadi Senior Attorney Florida Power & Light Company
E-MAIL COPY?          YES        NO YES        NO
Electronic Mail Distribution  
       
William A. Passetti Chief Florida Bureau of Radiation Control Department of Health Electronic Mail Distribution  
FP&L                                  4
cc w/encl:                              Mitch S. Ross
Richard L. Anderson                    Vice President and Associate General
Site Vice President                    Counsel
St. Lucie Nuclear Plant                Florida Power & Light Company
Electronic Mail Distribution            Electronic Mail Distribution
Robert J. Hughes                        Marjan Mashhadi
Plant General Manager                  Senior Attorney
St. Lucie Nuclear Plant                Florida Power & Light Company
Electronic Mail Distribution            Electronic Mail Distribution
Mark Hicks                              William A. Passetti
Operations Manager                      Chief
St. Lucie Nuclear Plant                Florida Bureau of Radiation Control
Electronic Mail Distribution            Department of Health
                                        Electronic Mail Distribution
Rajiv S. Kundalkar
Vice President - Fleet Organizational  Ruben D. Almaguer
Support                                Director
Florida Power & Light Company          Division of Emergency Preparedness
Electronic Mail Distribution            Department of Community Affairs
                                        Electronic Mail Distribution
Eric Katzman
Licensing Manager                      J. Kammel
St. Lucie Nuclear Plant                Radiological Emergency Planning
Electronic Mail Distribution            Administrator
                                        Department of Public Safety
Abdy Khanpour                           Electronic Mail Distribution
Vice President
Engineering Support                     Mano Nazar
Florida Power and Light Company         Executive Vice President and Chief Nuclear
P.O. Box 14000                         Officer
Juno Beach, FL 33408-0420               Florida Power & Light Company
                                        Electronic Mail Distribution
McHenry Cornell
Director                               (Vacant)
Licensing and Performance Improvement   Vice President
Florida Power & Light Company           Nuclear Plant Support
Electronic Mail Distribution           Florida Power & Light Company
                                        Electronic Mail Distribution
Alison Brown
Nuclear Licensing                      Jack Southard
Florida Power & Light Company           Director
Electronic Mail Distribution           Public Safety Department
                                        St. Lucie County
Faye Outlaw                            Electronic Mail Distribution
County Administrator
St. Lucie County
Electronic Mail Distribution


Ruben D. Almaguer Director Division of Emergency Preparedness Department of Community Affairs
FP&L                                     5
Electronic Mail Distribution
Letter to Mano Nazar from Kriss Kennedy dated January 19, 2010.
J. Kammel Radiological Emergency Planning Administrator
SUBJECT:       ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES
Department of Public Safety Electronic Mail Distribution
              INSPECTION - INSPECTION REPORT 05000335/2009006 AND
Mano Nazar Executive Vice President and Chief Nuclear
              05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS
Officer Florida Power & Light Company Electronic Mail Distribution
Distribution w/encl:
(Vacant)
C. Evans, RII
Vice President Nuclear Plant Support Florida Power & Light Company Electronic Mail Distribution
L. Slack, RII
Jack Southard Director
OE Mail
Public Safety Department St. Lucie County Electronic Mail Distribution 
RIDSNRRDIRS
FP&L 5 Letter to Mano Nazar from Kriss Kennedy dated January 19, 2010.  
PUBLIC
SUBJECT: ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000335/2009006 AND  
RidsNrrPMStLucie Resource
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  
          U. S. NUCLEAR REGULATORY COMMISSION
    Licensee: Florida Power & Light Company (FP&L)  
                            REGION II
Docket Nos.: 50-335, 50-389
  Facility: St. Lucie Nuclear Plant, Units 1 & 2  
License Nos.: DPR-67 and NPF-16
    Location: Jensen Beach, FL 34957  
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


    Dates:  August 3-14 (Weeks 1 & 2)    August 31-September 4 (Week 3)
                              SUMMARY OF FINDINGS
IR 05000335/2009006, 05000389/2009006; 8/3/2009 - 9/4/2009; St. Lucie Nuclear
  Inspectors: S. Rose, Senior Operations Inspector (Lead) R. Moore, Senior Reactor Inspector J. Hamman, Reactor Inspector R. Taylor, Senior Reactor Inspector
Plant, Units 1 and 2; NRC Component Design Bases Inspection.
M. Shylamberg, Contractor N. Della Greca, Contractor
This inspection was conducted by a team of four NRC inspectors from the Region II
    Approved by: Binoy Desai, Chief
office, and two NRC contract inspectors. Two findings of very low significance (Green)
    Engineering Branch 1    Division of Reactor Safety 
were identified during this inspection and were classified as non-cited violations. Also,
Enclosure SUMMARY OF FINDINGS  
two apparent violations (AV) with potential safety significance greater than Green were
IR 05000335/2009006, 05000389/2009006; 8/3/2009 - 9/4/2009; St. Lucie Nuclear Plant, Units 1 and 2; NRC Component Design Bases Inspection.  
identified. The significance of most findings is indicated by their color (Green, White,
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
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.  
which the SDP does not apply may be Green or be assigned a severity level after NRC
Cornerstone: Mitigating Systems  
management review. The NRC's program for overseeing the safe operation of
* 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,  
commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight
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)  
Process, (ROP) Revision 4, dated December 2006.
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.  
Cornerstone: Mitigating Systems
This finding is associated with the Mitigating Systems Cornerstone attribute of  
*   Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion
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  
    III, Design Control, for failure to translate the design basis as specified in the license
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  
    application into specifications, drawings, procedures, and instructions. The licensee
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]  
    did not ensure that the component cooling water (CCW) surge tank design included
 
    adequate overpressure protection for all procedurally allowed configurations as
3  Enclosure  
    required by the applicable ASME Boiler and Pressure Vessel Code, Section VIII,
* 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
    Division 1. The code requires that no intervening stop valves be between the vessel
actions included incorporating the correct design input and specifications by revising the calculations.   
    and its protective device or devices or between the protective devices and the point
The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Design Control.  It impacted the cornerstone
    of discharge. The team concluded that stop valve V6466 was an intervening stop
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
    valve for the CCW surge tank vent path to the chemical drain tank (CDT). The issue
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]
    was entered in the licensees corrective action program as condition report (CR)
* 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
    2009-23473. Immediate licensee corrective actions included verification that the
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.
    valve was in its open position and the implementation of administrative controls to
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
    maintain the valve open.
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
    This finding is associated with the Mitigating Systems Cornerstone attribute of
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 
    Design Control, i.e. initial design, was determined to be more than minor because it
4  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]
    impacted the cornerstone objective to ensure the availability, reliability, and capability
* 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
    of systems that respond to initiating events to prevent undesirable consequences.
containment IA compressors.
    The team determined that if left uncorrected, this design deficiency had the potential
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
    to impact the operability of safety-related systems and, thus, become a more
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
    significant safety concern in that a closed intervening valve had the potential for
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]
    overpressurizing the CCW surge tank. The team assessed this finding for
 
    significance in accordance with NRC Manual Chapter 0609, Appendix A, Attachment
Enclosure REPORT DETAILS
    1, Significance Determination Process (SDP) for Reactor Inspection Findings for At-
1. REACTOR SAFETY
    Power Situations, and determined that it was of very low safety significance (Green),
  Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
    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


1R21 Component Design Bases Inspection (71111.21)
                                            3
  .1 Inspection Sample Selection Process
* Green. The inspectors identified a finding involving a violation of 10 CFR 50,
 
  Appendix B, Criterion III, Design Control, for the licensees failure to maintain the
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
  safety-related 125V DC system design basis information consistent with the plant
-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
  configuration. Specifically, a revision to the Unit 1, safety-related 125V DC system
Modifications."
  analysis incorporated incorrect design input specifications. The issue was entered in
  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,
  the licensees corrective action program as CR 2009-24517. Licensee corrective
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-
  actions included incorporating the correct design input and specifications by revising
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
  the calculations.
is included in the following sections of the report.
  The finding was more than minor because it was associated with the Mitigating
.2 Results of Detailed Reviews
  Systems Cornerstone attribute of Design Control. It impacted the cornerstone
  .2.1 Component Cooling Water (CCW) Pumps 1A/1B/1C
  objective because if left uncorrected, it had the potential to lead to a more significant
    a. Inspection Scope
  safety concern in that future design activity or operability assessments would
  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
  assume the lower voltage (100V DC vs. actual 105V DC) value acceptable for
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 
  assuring the adequacy of voltage to the safety-related inverters. The team assessed
6  Enclosure 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
  this finding for significance in accordance with NRC Manual Chapter 0609, using the
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
  Phase I SDP worksheet for mitigating systems and determined that the finding was
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.
  of very low safety significance (Green) since it was a design deficiency determined
    b. Findings
  not to have resulted in a loss of safety function. This finding has a cross-cutting
  No findings of significance were identified; however, see section 4OA5 for two findings related to the CCW system.
  aspect in the area of human performance because the licensee failed to ensure that
.2.2 Component Cooling Water Surge Tank
  procedures (specifically ENG-QI 1.5) were available and adequate to assure nuclear
    a. Inspection Scope
  safety (specifically, complete, accurate and up-to-date design documentation):
  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
  H.2(c). [Section 1R21.2.20]
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
* TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design
ASME requirements.  Maintenance, corrective actions, and design change history were reviewed to assess potential component degradation and subsequent impacts on design margins.    b. Findings
  Control, for the licensees failure to identify that the CCW system met its license
  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
  specifications related to common cause failure vulnerabilities. Specifically, a non-
Boiler and Pressure Vessel Code, Section VIII, Division 1, was identified by the inspectors as a performance deficiency.  
  safety system failure (i.e. waste gas compressor aftercoolers affecting both units, or
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
  containment IA compressors affecting Unit 1 only) could result in a common cause
provided via a 2-inch vent line.  This vent line was routed to a diverting air-operated 
  failure of both trains of a safety system (i.e. CCW system). The issue was entered
7  Enclosure 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  
  into the licensees corrective action program as CR 2009-22929 with actions to
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.
  evaluate the past operability of the CCW system during the air intrusion event.
ASME Section VIII, Division 1, 1971 Edition, paragraph UG-134(e) states, "There shall
  Licensee corrective actions included isolating the CCW system from the containment
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
  IA compressors.
specification FLO-8770-764, originally issued on October 31, 1971.  Therefore, the inspectors concluded that ASME Section VIII, Division 1, 1971 edition applied.
   The finding was determined to be more than minor because if left uncorrected, it
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
  could affect the availability, reliability and capability of a safety system to perform its
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
  intended safety function. Specifically, with this vulnerability, a failure of the waste
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.  
   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
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.
  event, therefore, failing to ensure that adequate cooling would be available or
  maintained to essential equipment used to mitigate design bases accidents. The
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
  finding was assessed for significance in accordance with NRC Manual Chapter 0609,
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
  using the Phase I and Phase II SDP worksheets for mitigating systems. It was
likelihood of a loss of CCW. The team assessed this finding for significance in
  determined that a Phase III analysis was required since this finding represented a
8  Enclosure 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
  potential loss of safety system function for multiple trains which was not addressed
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.  
  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
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
  the finding for cross-cutting aspect and concluded that this finding did not have an
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
                                                                                    Enclosure
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.
 
 
9  Enclosure    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)).


    
                                            4
10  Enclosure 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
   associated cross-cutting aspect because the design of the CCW system was
Performance Monitoring of the IA Compressor Emergency Cooling System.
  established in an original plant design, and therefore, was not representative of
.2.4 GD-1/2 Gravity Damper On HVS-5A/B Outlet
  current licensee performance. [Section 4OA5]
    a. Inspection Scope
* TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI, Corrective
   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;
  Action, for the licensees failure to implement adequate corrective actions associated
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
  with the CCW air intrusion event that occurred in October, 2008. The corrective
configuration was consistent with the design basis and plant drawings.  
  actions were inadequate in that the licensee failed to identify and correct the cause
    b. Findings
  of air intrusion. The issue was entered in the licensees corrective action program as
   No findings of significance were identified.
  CR 2009-25209 to address the ineffective corrective actions for the air intrusion
.2.5 Pressurizer Relief Valve Isolation Valves, V1403 and V1405
  event. Licensee corrective actions included isolating the CCW system from the
    a. Inspection Scope
  containment IA compressors.
  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
   The finding was determined to be more than minor because it affected the
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
  availability, reliability and capability of a safety system to perform its intended safety
that the maintenance practices were consistent with design basis requirements.  
  function. Specifically, without knowing the leak path from the containment IA
    b. Findings
  compressors to the CCW system, the licensee could not ensure that adequate
  No findings of significance were identified
  cooling would be available or maintained to essential equipment used to mitigate
 
  design bases accidents. The finding was assessed for significance in accordance
11  Enclosure .2.6 Battery Charger 1B
  with NRC Manual Chapter 0609, using the Phase I and Phase II SDP worksheets for
    a. Inspection Scope
  mitigating systems. It was determined that a Phase III analysis was required since
  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
  this finding represented a loss of safety system function for multiple trains which was
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.  
  not addressed by the Phase II pre-solved tables/worksheets. Based on the Phase III
The team reviewed the health report and the results of recent tests to verify that the  
  SDP, the finding was preliminarily determined to be greater than Green. This finding
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.  
  was determined to have a cross-cutting aspect in the area of Human Performance,
  Decision Making, specifically H.1(a). [Section 4OA5]
  b. Findings
                                                                                  Enclosure
  No findings of significance were identified.  
 
.2.7 125V DC Bus 1B Power Panel & Cross-Tie Breakers to 125V DC Bus 1AB
                                        REPORT DETAILS
    a. Inspection Scope
1.   REACTOR SAFETY
  The team reviewed the Class 1E DC electrical distribution system DBD, applicable drawings and documents, including appropriate sections of the FSAR, to identify the
      Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
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  
1R21 Component Design Bases Inspection (71111.21)
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
.1   Inspection Sample Selection Process
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.
      The team selected risk significant components and operator actions for review using
 
      information contained in the licensees Probabilistic Risk Assessment (PRA). In general,
12  Enclosure    b. Findings
      this included components and operator actions that had a risk achievement worth factor
  No findings of significance were identified.  
      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
                                                                                        Enclosure
 
                                                6
      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 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
      licensees failure to translate the design basis as specified in the license application, into
      specifications, drawings, procedures, and instructions. Specifically, the licensees 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 tanks 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
                                                                                        Enclosure
 
                                            7
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 licensees 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 licensees 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 licensees 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
                                                                                    Enclosure
 
                                              8
    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 licensees
    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.
                                                                                      Enclosure
 
                                              9
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)).
                                                                                      Enclosure
 
                                              10
    This item will remain unresolved pending the completion of the stations testing, and
    NRC review of the results of the IA emergency cooling systems 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 licensees 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
                                                                                      Enclosure
 
                                              11
.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 manufacturers 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
    manufacturers 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.
                                                                                        Enclosure
 
                                              12
  b. Findings
    No findings of significance were identified.
.2.8 Engineered Safety Features Actuation System and Diverse Scram System
.2.8 Engineered Safety Features Actuation System and Diverse Scram System
    a. Inspection Scope
  a. Inspection Scope
  The team reviewed the engineered safety features actuation system (ESFAS) and  
    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  
    diverse scram system (DSS) design basis document and applicable sections of the TS
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  
    and FSAR to identify the design bases and the operational and maintenance
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  
    requirements for the ESFAS and DSS. The team reviewed the DSS components
components to evaluate installation criteria used and to assess their observable material condition.  
    including transmitters, logic modules, control and monitoring instrumentation, actuation
    b. Findings
    relays and contactors, selected components, and instrument loops associated with the
 
    ESFAS. The review included a detailed evaluation of instrument loop diagrams, control
No findings of significance were identified.  
    logic, and wiring diagrams to confirm that the design conformed to the intended
.2.9 Pressurizer Pressure Instrumentation
    operation of the systems. The review also addressed voltage requirements and voltage
    a. Inspection Scope
    available at the various components, circuit protection, channel separation, and electrical
  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  
    isolation. The team reviewed test procedures and evaluated the tests performed to
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  
    demonstrate the capability of the systems to perform the design basis functions. The
component and loop calibration procedures, test results, and adequacy of overlapping
    review included instrument and loop calibration procedures, test results, and adequacy
13 Enclosure 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  
    of overlapping tests. The team confirmed that system and component maintenance was
material condition.  
    conducted per vendor recommendations. Additionally, a review of the latest system
    b. Findings
    health report and recent problem reports was conducted to evaluate whether component
  No findings of significance were identified.  
    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
                                                                                    Enclosure
 
                                              13
      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.
                                                                                      Enclosure


.2.10 Start-Up Transformers 1A and 1B and associated supply and feeder breakers
                                              14
    a. Inspection Scope
   b.  Findings
   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
      No findings of significance were identified.
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.
 
 
14  Enclosure    b. Findings
  No findings of significance were identified.  
.2.12 480V Switchgear 1B2 (feeder and supply breakers and transformers)
.2.12 480V Switchgear 1B2 (feeder and supply breakers and transformers)
    a. Inspection Scope
  a. Inspection Scope:
:  The team reviewed the TS, DBD, design drawings, calculations, vendor data and  
      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  
      manuals and plant procedures to identify the design, maintenance and operational
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.  
      requirements. Electrical elementary drawings and wiring diagrams were reviewed to
    b. Findings
      verify that power sources would be available and adequate to power the appropriate
  No findings of significance were identified.  
      safety loads during accident/event conditions. The team reviewed preventive
      maintenance and testing procedures and results to determine if the breakers were
.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)
      maintained in accordance with industry and vendor standards and recommendations.
    a. Inspection Scope
      The team reviewed short circuit and protection calculations to ensure that the breakers
  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  
      could provide the appropriate interrupting and coordination protection. Selected
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.  
      corrective action reports were reviewed to determine if conditions adverse to quality
    b. Findings
      were appropriately addressed and corrected. A walk down of the 480V 1B2 Breaker
  No findings of significance were identified.
      panel was performed to assess installation, configuration, observable material condition
 
      and vulnerability to hazards.
 
  b. Findings
15 Enclosure .2.14 Intersystem Loss of Coolant Accident (LOCA) Instrumentation
      No findings of significance were identified.
    a. Inspection Scope
.2.13 Temperature Indication Switches for Reactor Coolant Pump (RCP) 1A and 1B CCW
  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.
      Seal Cooler Heat Exchanger Outlet (TIS-14-32A1/B1/B2/A2)
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  
  a. Inspection Scope
were accomplished in a timely matter.  
      The team reviewed design and licensing basis documents, drawings and vendor
    b. Findings
      manuals to identify the design requirements for the temperature indication switches.
  No findings of significance were identified.  
      The team reviewed set point calculations to verify that set points were established in
.2.15 Safety Injection Tank (SIT) Instrumentation
      accordance with vendor data, equipment capability and system design parameters.
    a. Inspection Scope
      Procedures were reviewed to verify alarm levels had been consistently translated from
  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  
      calculation data to ensure appropriate protection for an RCP seal leak. The team
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.  
      reviewed calibration records and procedures to verify that instrument accuracy was
    b. Findings
      monitored and maintained. Maintenance history, as demonstrated by work orders and
  No findings of significance were identified.  
      corrective action records, was reviewed to note any anomalies in equipment history and
.2.16 Safety Injection (SI) System Check Valves (V3227, V07174, V07172, V3106, V3107)
      to verify corrective actions were accomplished in a timely matter.
    
  b. Findings
  a. Inspection Scope
      No findings of significance were identified.
  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,  
                                                                                      Enclosure
preventive and corrective maintenance, and CRs, was reviewed to verify that potential
 
16 Enclosure 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.  
                                              15
.2.14 Intersystem Loss of Coolant Accident (LOCA) Instrumentation
  b. Findings
  a. Inspection Scope
  No findings of significance were identified.  
      The team reviewed design and licensing basis documents, drawings and vendor
.2.17 Volume Control Tank (VCT) MOVs 2501 & 2504
      manuals to identify the design requirements and capabilities of the intersystem LOCA
    a. Inspection Scope
      instrumentation. The following instrumentation was included in the review: CCW Surge
  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  
      Tank Level (LS-14-1A and B; LS-14-5, LG-14-2A and B); CCW System Radiation
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  
      Monitors; Reactor/Auxiliary Building (RAB) Sump Level; and RAB Radiation Monitors.
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.  
      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
  b. Findings
      design parameters. Appropriate procedures were reviewed to verify set point data and
  No findings of significance were identified.  
      alarm points had been consistently translated. The team reviewed calibration records
.2.18 CCW Control Valves (HCV-14-8A, HCV-14-8B, & HCV-14-9)
      and procedures to verify that instrument accuracy was monitored and maintained.
    a. Inspection Scope
      Maintenance history, as demonstrated by work orders and corrective action records, was
  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 note any anomalies in equipment history and to verify corrective actions
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  
      were accomplished in a timely matter.
data base as appropriate.  
  b. Findings
    b. Findings
      No findings of significance were identified.
  No findings of significance were identified.  
.2.15 Safety Injection Tank (SIT) Instrumentation
 
  a. Inspection Scope
17 Enclosure .2.19 SIT Outlet Valves (V3634, V3614, V3624, & V3644)
      The team reviewed design and licensing basis documents, drawings and vendor
    a. Inspection Scope
      manuals to identify the design requirements and capability of the safety injection tank
  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  
      instrumentation. The team reviewed set point calculations to verify that set points and
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.  
      levels were established in accordance with vendor data, equipment capability and
    b. Findings
      system design parameters. Appropriate procedures were reviewed to verify alarm levels
  No findings of significance were identified.  
      and set point data had been consistently translated. The team reviewed calibration
.2.20 Motors and Electrical Components in Inspection Scope
      records and procedures to verify that instrument accuracy was monitored and
    
      maintained. Maintenance history, as demonstrated by work orders and CRs, was
  a. Inspection Scope
      reviewed to note any anomalies in equipment history and to verify corrective actions
  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  
      were accomplished in a timely matter.
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  
  b. Findings
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.  
      No findings of significance were identified.
.2.16 Safety Injection (SI) System Check Valves (V3227, V07174, V07172, V3106, V3107)
  b. Findings
   a. Inspection Scope
  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  
      The team reviewed the DBD, related design basis documentation, drawings, TS, and the
information consistent with the plant configuration. Specifically, a revision to the Unit 1,
      FSAR to identify design, maintenance, and operational requirements for selected SI
18 Enclosure 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  
      system check valves. Maintenance history, as demonstrated by system health reports,
deficiency.  
      preventive and corrective maintenance, and CRs, was reviewed to verify that potential
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  
                                                                                      Enclosure
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.  
                                                16
Through discussions with the licensee pertaining to the discrepancy between the current plant configuration and the 125V DC system design analysis, the inspection team  
      degradation was being monitored and addressed. The team conducted interviews with
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  
      the SI System Engineer to obtain additional information and verify the stations
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  
      implementation and analysis of industry operating experience related to check valves.
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  
  b. Findings
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.  
      No findings of significance were identified.
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  
.2.17 Volume Control Tank (VCT) MOVs 2501 & 2504
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  
  a. Inspection Scope
Phase I SDP worksheet for mitigating systems and determined that the finding was of
      The team reviewed the system DBD, related design basis support documentation,
19  Enclosure 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  
      drawings, TS, and the FSAR to identify design, maintenance, and operational
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  
      requirements for these MOVs. Maintenance history, as demonstrated by system health
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)]  
      reports, preventive and corrective maintenance, and CRs, was reviewed to verify that
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  
      potential degradation was being monitored and addressed. The MOV sizing calculations
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  
      were reviewed to verify that the valves could operate during all credited design bases
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  
      events and that the licensee appropriately translated the correct valve dimensions and
Policy: NCV 05000335,389/2009006-03, Failure to Maintain the Safety-Related 125V DC System Design Basis Information Consistent with the Plant Configuration.  
      other significant characteristics into the sizing calculations. A review was conducted of
.3 Review of Low Margin Operator Actions
      the licensees 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
  a. Inspection Scope
      proper operation of the valve. The team reviewed vendor recommendations for
  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  
      preventative maintenance and operation to determine if maintenance practices were
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  
      consistent with design basis requirements.
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  
  b. Findings
valves was consistent with applicable drawings and system operating procedures.
      No findings of significance were identified.
20 Enclosure Operator actions associated with the following events/evolutions were reviewed:  
.2.18 CCW Control Valves (HCV-14-8A, HCV-14-8B, & HCV-14-9)
* Reactor coolant system feed and bleed and Power Operated Relief Valve (PORV) fails open (block valve use)  
  a. Inspection Scope
* Inner-system Loss of Coolant Accident (LOCA)
      The team reviewed applicable portions of the FSAR, DBD, and drawings to identify
* Anticipated Transient Without a Scram (ATWS) - Emergency Boration  
      design basis requirements for these valves. The air operator sizing calculations were
* Cross-tie 480V 1AB load center  
      reviewed to verify inputs were consistent with the most limiting design basis operating
* Condensate storage tank makeup from the treated water storage tank  
      conditions. Procurement documentation for the solenoids was reviewed to verify
* Restoration of non-essential CCW following Safety Injection Actuation Signal (SIAS)  
      compliance with environmental qualification (EQ) requirements. Stroke time surveillance
    b. Findings
      test procedures/results were reviewed to verify that stroke times were consistent with
  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  
      design basis requirements and to identify any adverse trends. The vendor manual was
not address the potential adverse impact on essential cooling flow required to mitigate a LOCA when the non-essential CCW was restored.  
      reviewed to identify recommendations for inspection and maintenance. The CR history
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  
      was reviewed to identify failures and determine whether they were entered into the MR
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.  
      data base as appropriate.
The station CCW flow balance procedure (1-NOP-14.02, Rev. 20, Appendix I) positioned  
  b. Findings
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  
      No findings of significance were identified.
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  
                                                                                      Enclosure
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
                                                17
21 Enclosure 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  
.2.19 SIT Outlet Valves (V3634, V3614, V3624, & V3644)
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.  
  a. Inspection Scope
.4 Review of Industry Operating Experience
      The team reviewed the system DBD, related design basis support documentation,
    a. Inspection Scope
      drawings, TS, and the FSAR to identify design, maintenance, and operational
  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  
      requirements for these MOVs. Maintenance history, as demonstrated by system health
issues that received a detailed review by the team included:  
      reports, preventive and corrective maintenance, and CRs, was reviewed to verify that
* Generic Letter 07-01, Inaccessible or Underground Power Cable Failures that Disable Accident Mitigation Systems or Cause Plant Transients.
      potential degradation was being monitored and addressed. A review was conducted of
* Generic Letter 98-02, Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition.
      the licensees testing procedures and results from diagnostic valve testing to verify the
* NRC Information Notice 07-09, Equipment Operability Under Degraded Voltage Conditions.  
      MOVs were tested in a manner that would detect a malfunctioning valve and verify
* Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient, dated July 25, 1984  
      proper operation of the valve. The team reviewed maintenance practices and vendor
* NRC Information Notice 2008-02: Findings Identified During Component Design Bases Inspections, March 19, 2008  
      recommendations for preventative maintenance and operation to verify that the valves
    b. Findings
      were being maintained consistent with design basis requirements.
  No findings of significance were identified.  
  b. Findings
.5 Review of Permanent Plant Modifications
      No findings of significance were identified.
    a. Inspection Scope  
.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
      licensees failure to maintain the safety-related 125V DC system design basis
      information consistent with the plant configuration. Specifically, a revision to the Unit 1,
                                                                                        Enclosure
 
                                          18
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 licensees 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 stations 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 licensees 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
                                                                                    Enclosure
 
                                                19
    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 stations 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 licensees 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.
                                                                                        Enclosure
 
                                            20
  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 licensees 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 licensees 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
                                                                                    Enclosure
 
                                                21
    the steam generator sample coolers. The licensees failure to provide adequate
    procedures for restoration of non-essential CCW following a SIAS was identified as a
    performance deficiency. The licensees 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
                                                                                        Enclosure


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
                                                22
by performing activities identified in IP 71111.17, Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications. The following modification was reviewed:
  b. Findings
* PC/M: 04028, Medium Voltage Switchgear Circuit Breaker Replacement - Phase III
    No findings of significance were identified.
 
 
22  Enclosure    b. Findings
  No findings of significance were identified.  
4OA5 Other Activities
4OA5 Other Activities
  CCW Air Intrusion Event
    CCW Air Intrusion Event
    a. Inspection Scope
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.  
    The team performed a detailed review of the condition reports related to the air intrusion
    b. Findings
    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.
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).  
b. Findings
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  
    Introduction: The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design
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.  
    Control, for the licensees failure to translate the design basis, as specified in the license
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.  
    application, into specifications, drawings, procedures, and instructions. Specifically, a
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-
    non-safety system failure (i.e. containment IA compressors) could cause a common
related CCW system. In this case, the design deficiency was a common cause failure mechanism.  
    cause failure of both trains of a safety system (i.e. CCW system).
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  
    Description: The Unit 1 design included IA compressors inside containment. The Unit 1
CCW flow was also used for waste gas compressor aftercooler cooling. The waste gas
    CCW system non-essential header provided cooling and seal makeup to these IA
23 Enclosure 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.  
    compressors. On October 16, 2008, an air intrusion event occurred in which air from the
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  
    IA compressors located inside containment entered into the CCW system. The licensee
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  
    determined the air intrusion into the CCW system was caused by the failures of IA
tube regions), thus significantly decreasing the heat transfer capability.  
    system check valves V1818A and V18060 to the IA receiver tank combined with the
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:  
    failure of the IA unloading solenoid SE1814A. Additionally, leakage through the IA seal
* Providing adequate cooling for those safety-related components associated with containment and reactor decay heat removal during accident conditions.  
    water cooler, which interfaces with the CCW system, created pathways for air to enter
* Providing adequate cooling for those safety-related components associated with achieving safe shutdown.  
    the CCW system.
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  
    The inspectors review of the CCW system CRs identified that the air intrusion event
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.  
    occurred from 2:13 a.m. on October 16, 2008 through 4:02 a.m. on October 17, 2008.
Analysis: 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  
    The teams review identified that this event resulted in the degraded performance of both
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.
    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
24 Enclosure 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.
    the air intrusion. Subsequent to this, operators vented a significant amount of air from
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,  
    the CCW system in order to return the system parameters to normal. The air intrusion
using the Phase I and Phase II SDP worksheets for mitigating systems.  
    event demonstrated an original design deficiency on Unit 1 such that a non-safety
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.  
    system (IA) could adversely impact the reliability, capability, and availability of the safety-
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.    
    related CCW system. In this case, the design deficiency was a common cause failure
    mechanism.
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.  
    In addition to the air intrusion source discussed above, the team also determined that
Enforcement: 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  
    this vulnerability potentially existed on the waste gas compressors since non-essential
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.  
    CCW flow was also used for waste gas compressor aftercooler cooling. The waste gas
  Introduction: 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.  
                                                                                          Enclosure
25  Enclosure 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.  
 
Description: 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  
                                          23
and heat exchangers in order to return the system parameters to normal. As discussed in section 4OA5 b.1, the licensee identified that the containment IA compressors provided a pathway for which air intrusion into the CCW system could occur.  
compressors run at approximately 160 psig system pressure and the CCW system
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  
pressure is approximately 120 psig. The common cause failure vulnerability of the 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.  
system from a failure in the waste compressor units was applicable to both Unit 1 and
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:  
Unit 2.
* 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.  
The CCW system essential header cools the containment fan coolers (CFCs), shutdown
* 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.  
cooling heat exchanger, and bearing/seal coolers for the containment spray, high
* 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.  
pressure safety injection, and low pressure safety injection pumps. The CCW trains are
* 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.  
normally cross-connected during normal operation. The team concluded that the air
* 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.  
intrusion affecting both CCW trains could have prevented the CCW system from
* 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.  
delivering the flow specified by the TS Surveillance Requirement 4.6.2.1.1 (1,200 gpm to
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  
each cooling train fan unit), and reduced flow to the remaining safety-related heat
the event. The licensee had not performed an engineering evaluation to support the
exchangers below the analyzed/required values. An additional impact of the air intrusion
26 Enclosure 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.  
into the CCW system was potential degradation of the safety-related heat exchangers
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  
performance. The team concluded that given enough air introduction, the possibility
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.  
existed that the heat exchangers could become fully or partially air bound (e.g., upper
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.
tube regions), thus significantly decreasing the heat transfer capability.
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  
The combined effects of the reduced flow and the reduced heat transfer could lead to
(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  
the inability of the CCW system to perform the following safety-related functions:
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.   The licensee initiated CR 2009-23882 to address this concern.  
*   Providing adequate cooling for those safety-related components associated with
Analysis: 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  
    containment and reactor decay heat removal during accident conditions.
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  
*   Providing adequate cooling for those safety-related components associated with
the Phase II pre-solved tables/worksheets.  
    achieving safe shutdown.
27 Enclosure 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  
This event simultaneously affected both redundant trains of the CCW system (i.e.
attached.  
introduced a common cause failure mechanism). FSAR section 9.2.2.3.2, Single Failure
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.  
Analysis, states in part: there is no single failure that could prevent the component
Enforcement: 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  
cooling system from performing its safety function. The licensees evaluation of the air
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  
intrusion event failed to evaluate the operability consequences of the air intrusion on the
a Non-Safety Related System Could Cause a Common Mode Failure of Both Trains of a Safety-Related System.  
CCW flow reduction to the safety-related heat exchangers and failed to consider the
4OA6 Meetings, Including Exit
effect of the air intrusion on the heat exchangers performance. The licensee initiated
  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.  
CR 2009-22929 with actions to evaluate the past operability of the CCW system during
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.  
the air intrusion event.
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.  
Analysis: An original plant design deficiency was revealed by the CCW air intrusion
On December 10, 2009, the NRC presented inspection results in a telephone exit with  
event of October 16, 2008. This design deficiency involved the potential for a non-safety
Mr. Eric Katzman and other members of the licensee's staff.  
system (IA or waste gas) adversely impacting the reliability, capability, and availability of
  ATTACHMENT: SUPPPLEMENTAL INFORMATION
the safety-related CCW system. This design deficiency was identified as a performance
Attachment SUPPLEMENTAL INFORMATION
deficiency. In this case, the design deficiency introduced a common cause failure
  KEY POINTS OF CONTACT  
mechanism. FSAR section 9.2.2.3.2, Single Failure Analysis, states, in part: there is no
Licensee personnel:
single failure that could prevent the CCW system from performing its safety function.
  P. Barnes, Mechanical Engineering Design Supervisor D. Cecchett, Licensing G. Johnston, Site Vice President E. Katzman, Licensing Manager  
This single failure vulnerability existed on Units 1 and 2 from potential failure of the
D. Lany, Operations Senior Reactor Operator J. Porter, Manager Design Engineering S. Short, Electrical Engineering Design Supervisor  
aftercoolers on the waste gas compressors and on Unit 1 from the potential failure of the
NRC personnel
containment IA system.
D. Jones, Acting Chief, Engineering Branch Chief 1, Division of Reactor Safety, RII T. Hoeg, Senior Resident Inspector, St. Lucie  
                                                                                    Enclosure
W. Rogers, Senior Risk Analyst, RII S. Sanchez, Resident Inspector, St. Lucie  
 
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED  
                                            24
The finding was determined to be more than minor because it was associated with the
Opened and Closed  
Mitigating Systems Cornerstone attribute of Equipment Performance. It impacted the
  05000335,389/2009006-01 NCV Failure to Meet the ASME Boiler and Pressure Vessel Code, Section VIII, Division 1 Requirements for the Overpressure Protection  
cornerstone objective because, if left uncorrected, it would affect the availability,
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)  
reliability and capability of a safety system to perform its intended safety function.
 
Specifically, with this vulnerability, a failure of the waste gas aftercooler (either unit) or a
Attachment LIST OF DOCUMENTS REVIEWED Calculations
failure of the containment IA compressors (Unit 1 only) could cause air intrusion into the
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  
CCW system and potentially lead to a loss of CCW event. A loss of CCW could result in
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  
inadequate cooling to essential equipment used to mitigate design bases accidents. The
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  
finding was assessed for significance in accordance with NRC Manual Chapter 0609,
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  
using the Phase I and Phase II SDP worksheets for mitigating systems.
PSL-1-FJE-90-0026, Unit 1 Short Circuit, Voltage Drop and PSB-1 Analysis, Rev 6  
It was determined that a Phase III analysis was required since this finding represented a
Specifications
potential loss of safety system function for multiple trains which was not addressed by
FLO-8770-764, Unit 1 CCW Surge Tank, original issue 10/31/71  
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.
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.
Enforcement: 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: The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI,
Corrective Action, for the licensees 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.
                                                                                      Enclosure
 
                                          25
The licensees 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.
Description: 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 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 teams 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 compressors 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 licensees 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 licensees operations department at the time of
the event. The licensee had not performed an engineering evaluation to support the
                                                                                  Enclosure
 
                                            26
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.
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 teams 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. The
licensee initiated CR 2009-23882 to address this concern.
Analysis: The licensees 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.
                                                                                      Enclosure
 
                                                27
    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.
    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 licensees 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.
    Enforcement: 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.
4OA6 Meetings, Including Exit
    On September 4, 2009, the team presented the preliminary inspection results to Mr.
    Johnston and other members of the licensees 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 licensees staff.
    On December 3, 2009, the NRC presented preliminary inspection results in a telephone
    with Mr. Eric Katzman and other members of the licensees staff.
    On December 10, 2009, the NRC presented inspection results in a telephone exit with
    Mr. Eric Katzman and other members of the licensees staff.
ATTACHMENT: SUPPPLEMENTAL INFORMATION
                                                                                      Enclosure
 
                              SUPPLEMENTAL INFORMATION
                                  KEY POINTS OF CONTACT
Licensee personnel:
P. Barnes, Mechanical Engineering Design Supervisor
D. Cecchett, Licensing
G. Johnston, Site Vice President
E. Katzman, Licensing Manager
D. Lany, Operations Senior Reactor Operator
J. Porter, Manager Design Engineering
S. Short, Electrical Engineering Design Supervisor
NRC personnel
D. Jones, Acting Chief, Engineering Branch Chief 1, Division of Reactor Safety, RII
T. Hoeg, Senior Resident Inspector, St. Lucie
W. Rogers, Senior Risk Analyst, RII
S. Sanchez, Resident Inspector, St. Lucie
                      LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened and Closed
05000335,389/2009006-01           NCV         Failure to Meet the ASME Boiler and Pressure
                                                Vessel Code, Section VIII, Division 1
                                                Requirements for the Overpressure Protection
                                                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)
                                                                                    Attachment
 
                              LIST OF DOCUMENTS REVIEWED
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 CEs 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
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-NOP-14.02, Component Cooling Water System Operation, Rev. 25
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  
2-NOP-14.02, Component Cooling Water System Operation, Rev. 15
3 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  
1-NOP-50.01A, 125V DC Bus 1A (Class 1E) Normal Operation, Rev 2
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
1-NOP-50.01AB, 125V DC Bus 1AB (Class 1E) Normal Operation, Rev 0A
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  
2-GOP-403, Reactor Plant Heatup - Mode 4 to Mode 3, Rev. 32
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  
1-0330020, Turbine Cooling Water System, Rev. 57C
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  
1-0330030, Turbine Cooling Water System, Rev. 16A
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  
1-1010030, Loss Of Instrument Air, Rev. 33a
1-EOP-99, Appendices / Figures / Tables / Data Sheets, Rev. 38
                                                                                    Attachment
 
                                              3
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
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
1-OSP-14.01A, Component Cooling Water Pump Code Run, performed on: 6/12/09, 3/12/09,
4 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  
    12/11/08, 9/12/08, 7/7/08, 3/15/08
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  
1-OSP-14.01B, Component Cooling Water Pump Code Run, performed on: 6/26/09, 3/27/09,
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  
    12/26/08, 9/26/08, 6/26/08, 3/27/08
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
1-OSP-14.01C, Component Cooling Water Pump Code Run, performed on: 6/26/09, 3/27/09,
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  
    12/26/08, 9/26/08, 6/26/08, 3/27/08
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  
1-NOP-14.02, Component Cooling Water System Operation, Appendix I, Essential CCW Load
5  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  
    Flow Balance, performed on: 11/18/08
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  
2-NOP-14.02, Component Cooling Water System Operation, Appendix I, Essential CCW Load
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  
    Flow Balance, performed on: 05/29/09
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  
                                                                                  Attachment
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  
                                          4
6 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  
Drawings
9 8770-B-327, sheet 533, Safeguards Room B Sump Pumps Unit 1 Control Wiring Diagram, Rev  
8770-G-078, Sheet 162A, Flow Diagram, Waste Management System, Rev. 14
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  
8770-G-082, Sheet 1, Flow Diagram, Circulating & Intake Cooling Water System, Rev. 50
14 8770-B-327, sheet 907, Unit 1 Control Wiring Diagram Startup Transformer 1B-2 Breaker, Rev  
8770-G-082, Sheet 2, Flow Diagram, Circulating & Intake Cooling Water System, Rev. 23
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  
8770-G-083, Sheet 1A, Flow Diagram, Component Cooling System, Rev. 59
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-083, Sheet 1B, Flow Diagram, Component Cooling System, Rev. 57
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  
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
 
                                            5
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
 
                                            6
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)
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  
1998-1584, Unit 1 & 2 Charging Pump Surveillance Test Flow Rates Do Not Meet the Design
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  
    Maximum Flow Rates
7 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  
2005-1294, 1C CCW Pump Exceed The Maintenance Rule Unavailability Limit Of 200
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  
    Hours/Year/Pump
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  
2005-2969, Letdown HX CCW Relief Was Found Lifting After 2A CCW Pump Start
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  
2005-30300, Inter-System LOCA Detection Instrumentation for Reactor Pressure Boundary Not
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  
    Prioritize for Deficiency Resolution Prior to Mode 4
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
2007-27048, Incorrect Safety Classification of a DBD Function for Valve TCV-14/4A/4B
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  
2007-28391, Parameter Limits for ICW Operability Performance Curves
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  
2007-35587, PMs Being Changed From Daily to Outage During the Outage
2004-9733, SIT Outlet Valve V3614 Failed to Open.  
2008-31947, Air introduction into CCW System
Completed Work Orders (WOs)
2008-34697, Air introduction into CCW System per CR 2008-31947
38025447, Air Leaked into CCW - Troubleshoot, dated 11/18/08 31023348-01, Unit 1 Replacement of AM507 and AM517, 12/12/01  
2008-35753, Isolate CCW to Containment IA Compressors Aftercoolers
8 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  
2008-37070, St. Lucie Engineering Self-Assessment - Component Design Basis Inspection
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  
2009-19025, Site Glass accidentally broken
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  
                                                                                  Attachment
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  
                                            7
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  
2009-23473, CCW Surge Tank Design Basis Requirements for Code Pressure Relief Capacity
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
    and Design
9 Attachment Modifications
2005-6815, Low Margin Issue - Degraded Grid Action Plan
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  
2006-1885, 1A Battery Charger DC Output Breaker Lead Observed to Be Overheating
Miscellaneous Documents
2006-19927, Develop PMCRs for New SF6 Breakers
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  
2006-20023, During Performance of Breaker PM, 480V Brkr Found with Misaligned Contact
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  
2006-20094, EDG Breaker Closure Failure during Post-Maintenance Testing of EDG
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  
2006-22579, K-600 Breaker found to Have Several Problems
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  
2006-25939, Spare Load Center Breaker Could Not Be Set per Procedure
10 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
2006-30383, UNUSED toc Switch Contacts Do Not Function Properly
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  
2007-1920, 480 V Swgr Breaker Received Refurbishment by ABB in Trip Free Mode
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
2007-23473, EDG Loading Increase Due to Operation at Upper TS Frequency Limit
Nuclear Plant Switchyard Inspection Report (weekly) for breakers 8W23, 8W40, 8W61 and 8W64 Control Room Log, 7/19-21/2009  
2007-4859, K3000 Breaker Refurbishment by ABB Unsatisfactory
CRs and WOs Initiated Due to CDBI Activity:
2007-7456, 480V Swgr Breaker Failed to Trip during Testing
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  
2007-8304, Problem Found on 480V Swgr Breaker after Refurbishment by ABB
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  
2007-9889, Negative Trend in Performance of MCCBs Installed in MCCs
PSL2 2009-24030, A Past Operability Review of an Air Intrusion into CCW event from October 2008  
2007-10302, Hot Connection Observed on 1BB Battery Charger Neutral Lead Connection
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  
2007-13704, Review of IN 2007-09, Equipment Operability Under Degraded Voltage
11 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  
      Conditions.
Attachment PHASE III ANALYSIS  
2007-14099, 1A 125V DC System Swgr Undervoltage Relay out of Adjustment
  SRA Analysis Number: STL0904 Analysis Type: SDP Phase III  
2007-15321, 4.16KV Breaker for 1A LPSI Pump Did Not Charge Spring when Racked in
Inspection Report: 05000335, 389/2009006 Plant Name: St. Lucie   Unit Numbers: 1 & 2 Enforcement Action EA-09-321  
2007-28789, Track & Trend CR for Revision of DC Calculation due to Battery Inter-Cell
  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  
    Resistance.
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  
2007-29985, Jumpering-out of two battery cells
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.  
2007-34306, Medium Voltage Breaker Cluster Finger Problem
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  
2007-36484, 1D Battery Charger Control Board B Found to Be Defective during PM
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  
2007-39837, 480V Swgr 2A3-5B Loose Breaker Power Stab
performance deficiency.  
2008-14926, Adequacy of 1-OSP-50.01 to satisfy NRC Commitment to Test UV Trip
EXPOSURE TIME - One year will be used.  
Feature of DC Cross-Tie Breakers
  DATE OF OCCURRENCE - October 2008  
2008-26139, Hard Ground on 125V DC Bus 1BDefective Masterpact Circuit Breaker Trip Unit
SAFETY IMPACT > Green RISK ANALYSIS/CONSIDERATIONS  
2008-33033, Defective Masterpact Circuit Breaker Trip Unit
Assumptions  
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
 
                                          8
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
 
                                          9
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
 
                                            10
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 Operators
      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 MCCs
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
 
                                          11
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
                                                                                    Attachment
 
                                        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 licensees 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
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: Licensees full scope model
Significant Influence Factor(s) [if any]: How severe the air intrusion was on the CCW systems
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
 
                                                3
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
                                                                                        Attachment


1.  The performance deficiency will be modeled as an increase in the probability of an initiating event, Loss of the CCW system.    
EDITED FULL SCOPE MODEL CUTSETS FOR TOTAL LOSS OF COMPONENT COOLING WATER SYSTEM
2  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.    
      TOP 10,000 Cutsets for PSL1
   
C:\CAFTA32\06R0B\PSL1\Cuts-B\PSL1Top10K.CUT
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.   
    # Cutset Prob      Event Prob   Event      Description
4. With respect to Unit 2 a non-conforming case initiating event frequency will be set at 1/55
4832        8.06E-11      1.00E+00  %ZZCCWU1    LOSS OF CCW IE
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.   
                          1.00E+00   CHFPRCPTRP FAILURE TO TRIP REACTOR COOLANT PUMPS FOLLOWING LOSS OF COMPONENT COOLING WATER
5. No recovery will be considered after air intrusion severe enough to cause CCW pump failure. 
                          3.50E-06  CTKJ1STAIE CCW SURGE TANK RUPTURE FAILS TRAIN A (1 YR EXPOSURE)
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-
                          6.97E-03 CTM1CCWHXB CCW HX B IN TEST OR MAINTENANCE
conforming and conforming cases.
                          1.00E+00  RCPSL      RCP SEAL LOCA FLAG EVENT
  PRA Model used for basis of the risk analysis: Licensee's full scope model 
                          3.30E-03 ZHFPRCPTRP FAILURE TO TRIP RCPS LOSS OF CCW
  Significant Influence Factor(s) [if any]:  How severe the air intrusion was on the CCW system's
EDITED FOR LOSS OF COMPONENT COOLING WATER
ability to perform its numerous risk significant functions.
4832        3.30E-03      1.00E+00  %ZZCCWU1    LOSS OF CCW IE
CALCULATIONS
                          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)
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
                          1.00E+00  CTM1CCWHXB  CCW HX B IN TEST OR MAINTENANCE
comparable to SPAR in the GEM mode.   
                          1.00E+00   RCPSL      RCP SEAL LOCA FLAG EVENT
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
                          3.30E-03  ZHFPRCPTRP FAILURE TO TRIP RCPS LOSS OF CCW
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. 
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 1


RECOVERY DEVELOPMENT
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
Two perspectives will be applied to the recovery development, since the time variable could be applied differently. The more
warranted.  
liberal of the two calculations will be applied in the quantification.
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.
            DIAGNOSIS
      
            Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging
3  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
            BASE                      1.0E-02
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
            TIME                      1.0E+01    limited information available as to how much time was left prior to sys failure
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.
            STRESS                    2.0E+00    Unusual condition
 
            COMPLEXITY                1.0E+00     Nominal
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
            EXPERIENCE/TRAI          1.0E+00    Nominal
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:
            N
* an initiator for Loss of the CCW system
            PROCEDURES                1.0E+00    Nominal
* an undetected failure mechanism of any CCW functions while the equipment is in standby     
            ERGONOMICS                1.0E+00    Nominal
 
            FIT FOR DUTY              1.0E+00    Nominal
APPENDICES: 1. Full Scope Model Output  2. Recovery Development
            WORK PROCESS              1.0E+00    Nominal
            DIAGNOSITIC                2.0E-01
            TOTAL
            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/TRAI          1.0E+00    Nominal
            N
            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


    Analyst:  W. Rogers  Date: 10/30/09
DIAGNOSIS
Reviewed By:  G. MacDonald Date: 11/02/2009
Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging
 
BASE                     1.0E-02
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
TIME                   1.0E+00    limited information available as to how much time was left prior to sys failure
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
STRESS                 2.0E+00     Unusual condition
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
COMPLEXITY             1.0E+00     Nominal
Print date: 7/14/2009 2:28 PM
EXPERIENCE/TRAI         1.0E+00     Nominal
Not sorted 
N
  Printed in full
PROCEDURES             1.0E+00     Nominal
 
ERGONOMICS             1.0E+00     Nominal
Appendix 2 RECOVERY DEVELOPMENT    Two perspectives will be applied to the recovery development, since the time variable could be applied differently.  The more
FIT FOR DUTY           1.0E+00     Nominal
liberal of the two calculations will be applied in the quantification.   
WORK PROCESS           1.0E+00     Nominal
    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/TRAI
DIAGNOSITIC              2.0E-02
N 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/TRAI
TOTAL
N 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           
ACTION
 
BASE                     1.0E-03
   
TIME                   1.0E+01     apply time penalty that after diagnosis, time available = time req'd
  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/TRAI
STRESS                 2.0E+00
N 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/TRAI
COMPLEXITY             5.0E+00     numerous actions with multiple sub-tasks outside Main Control Room
N 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
EXPERIENCE/TRAI         1.0E+00     Nominal
N
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
                                                                                                Appendix 2
}}
}}

Revision as of 22:36, 13 November 2019

IR 05000335-09-006, 05000389-09-006, on August 3-14, August 31 - September 4, St. Lucie Units 1 & 2 - NRC Component Design Bases Inspection
ML100210081
Person / Time
Site: Saint Lucie  NextEra Energy icon.png
Issue date: 01/19/2010
From: Kennedy K
Division of Reactor Safety II
To: Nazar M
Florida Power & Light Co
References
IR-09-006
Download: ML100210081 (50)
v • d • e


See also: IR 05000335/2009006

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

SAM NUNN ATLANTA FEDERAL CENTER

61 FORSYTH STREET, SW, SUITE 23T85

ATLANTA, GEORGIA 30303-8931

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: ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES

INSPECTION - INSPECTION REPORT 05000335/2009006 AND

05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS

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 Commissions 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 licensees corrective

action evaluation did not identify the common cause failure vulnerability discussed in the first

performance deficiency.

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 NRCs 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

NRCs 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: Inspection Report 05000335/2009006, 05000389/2009006

w/Attachment: Supplemental Information

Docket Nos.: 50-335, 50-389

License Nos.: DPR-67 and NPF-16

cc w/encl: (See page 4)

_________________________ 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 NO

OFFICE RII:DRS RII:OE

SIGNATURE RA RA

NAME BDESAI CEVANS

DATE 1/11/2010 1/13/2010

E-MAIL COPY? YES NO YES NO

FP&L 4

cc w/encl: Mitch S. Ross

Richard L. Anderson Vice President and Associate General

Site Vice President Counsel

St. Lucie Nuclear Plant Florida Power & Light Company

Electronic Mail Distribution Electronic Mail Distribution

Robert J. Hughes Marjan Mashhadi

Plant General Manager Senior Attorney

St. Lucie Nuclear Plant Florida Power & Light Company

Electronic Mail Distribution Electronic Mail Distribution

Mark Hicks William A. Passetti

Operations Manager Chief

St. Lucie Nuclear Plant Florida Bureau of Radiation Control

Electronic Mail Distribution Department of Health

Electronic Mail Distribution

Rajiv S. Kundalkar

Vice President - Fleet Organizational Ruben D. Almaguer

Support Director

Florida Power & Light Company Division of Emergency Preparedness

Electronic Mail Distribution Department of Community Affairs

Electronic Mail Distribution

Eric Katzman

Licensing Manager J. Kammel

St. Lucie Nuclear Plant Radiological Emergency Planning

Electronic Mail Distribution Administrator

Department of Public Safety

Abdy Khanpour Electronic Mail Distribution

Vice President

Engineering Support Mano Nazar

Florida Power and Light Company Executive Vice President and Chief Nuclear

P.O. Box 14000 Officer

Juno Beach, FL 33408-0420 Florida Power & Light Company

Electronic Mail Distribution

McHenry Cornell

Director (Vacant)

Licensing and Performance Improvement Vice President

Florida Power & Light Company Nuclear Plant Support

Electronic Mail Distribution Florida Power & Light Company

Electronic Mail Distribution

Alison Brown

Nuclear Licensing Jack Southard

Florida Power & Light Company Director

Electronic Mail Distribution Public Safety Department

St. Lucie County

Faye Outlaw Electronic Mail Distribution

County Administrator

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

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

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 licensees 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

3

  • Green. The inspectors identified a finding involving a violation of 10 CFR 50,

Appendix B, Criterion III, Design Control, for the licensees 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 licensees 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]

Control, for the licensees 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 licensees 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

4

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]

Action, for the licensees 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 licensees 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]

Enclosure

REPORT DETAILS

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R21 Component Design Bases Inspection (71111.21)

.1 Inspection Sample Selection Process

The team selected risk significant components and operator actions for review using

information contained in the licensees 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

Enclosure

6

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 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

licensees failure to translate the design basis as specified in the license application, into

specifications, drawings, procedures, and instructions. Specifically, the licensees 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 tanks 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

Enclosure

7

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 licensees 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 licensees 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 licensees 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

Enclosure

8

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 licensees

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.

Enclosure

9

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)).

Enclosure

10

This item will remain unresolved pending the completion of the stations testing, and

NRC review of the results of the IA emergency cooling systems 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 licensees 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

Enclosure

11

.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 manufacturers 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

manufacturers 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.

Enclosure

12

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

Enclosure

13

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.

Enclosure

14

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.

Enclosure

15

.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 CRs, 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

Enclosure

16

degradation was being monitored and addressed. The team conducted interviews with

the SI System Engineer to obtain additional information and verify the stations

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 licensees 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.

Enclosure

17

.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 licensees 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

licensees failure to maintain the safety-related 125V DC system design basis

information consistent with the plant configuration. Specifically, a revision to the Unit 1,

Enclosure

18

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 licensees 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 stations 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 licensees 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

Enclosure

19

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 stations 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 licensees 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.

Enclosure

20

Operator actions associated with the following events/evolutions were reviewed:

fails open (block valve use)

  • Inner-system Loss of Coolant Accident (LOCA)
  • 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 licensees 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 licensees 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

Enclosure

21

the steam generator sample coolers. The licensees failure to provide adequate

procedures for restoration of non-essential CCW following a SIAS was identified as a

performance deficiency. The licensees 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:

Disable Accident Mitigation Systems or Cause Plant Transients.

Loss of Emergency Mitigation Functions While in a Shutdown Condition.

Conditions.

dated July 25, 1984

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

Enclosure

22

b. Findings

No findings of significance were identified.

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 licensees 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 inspectors 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 teams 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

Enclosure

23

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.

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:

  • 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 licensees 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: 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.

Enclosure

24

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.

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.

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.

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.

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.

Enforcement: 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: The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI,

Corrective Action, for the licensees 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.

Enclosure

25

The licensees 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.

Description: 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 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 teams 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 compressors 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 licensees 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 licensees operations department at the time of

the event. The licensee had not performed an engineering evaluation to support the

Enclosure

26

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.

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 teams 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. The

licensee initiated CR 2009-23882 to address this concern.

Analysis: The licensees 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.

Enclosure

27

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.

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 licensees 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.

Enforcement: 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.

4OA6 Meetings, Including Exit

On September 4, 2009, the team presented the preliminary inspection results to Mr.

Johnston and other members of the licensees 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 licensees staff.

On December 3, 2009, the NRC presented preliminary inspection results in a telephone

with Mr. Eric Katzman and other members of the licensees staff.

On December 10, 2009, the NRC presented inspection results in a telephone exit with

Mr. Eric Katzman and other members of the licensees staff.

ATTACHMENT: SUPPPLEMENTAL INFORMATION

Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel:

P. Barnes, Mechanical Engineering Design Supervisor

D. Cecchett, Licensing

G. Johnston, Site Vice President

E. Katzman, Licensing Manager

D. Lany, Operations Senior Reactor Operator

J. Porter, Manager Design Engineering

S. Short, Electrical Engineering Design Supervisor

NRC personnel

D. Jones, Acting Chief, Engineering Branch Chief 1, Division of Reactor Safety, RII

T. Hoeg, Senior Resident Inspector, St. Lucie

W. Rogers, Senior Risk Analyst, RII

S. Sanchez, Resident Inspector, St. Lucie

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened and Closed

05000335,389/2009006-01 NCV Failure to Meet the ASME Boiler and Pressure

Vessel Code,Section VIII, Division 1

Requirements for the Overpressure Protection

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)

Attachment

LIST OF DOCUMENTS REVIEWED

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 CEs 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

3

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

4

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

5

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

6

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

7

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

8

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

9

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

10

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 Operators

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 MCCs

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

11

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

Attachment

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 licensees 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

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: Licensees full scope model

Significant Influence Factor(s) [if any]: How severe the air intrusion was on the CCW systems

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

3

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

Attachment

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

  1. Cutset Prob Event Prob Event Description

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 1

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/TRAI 1.0E+00 Nominal

N

PROCEDURES 1.0E+00 Nominal

ERGONOMICS 1.0E+00 Nominal

FIT FOR DUTY 1.0E+00 Nominal

WORK PROCESS 1.0E+00 Nominal

DIAGNOSITIC 2.0E-01

TOTAL

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/TRAI 1.0E+00 Nominal

N

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/TRAI 1.0E+00 Nominal

N

PROCEDURES 1.0E+00 Nominal

ERGONOMICS 1.0E+00 Nominal

FIT FOR DUTY 1.0E+00 Nominal

WORK PROCESS 1.0E+00 Nominal

DIAGNOSITIC 2.0E-02

TOTAL

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/TRAI 1.0E+00 Nominal

N

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

Appendix 2

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