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{{#Wiki_filter:UNITED STATES
{{#Wiki_filter:UNITED STATES  
                              NUCLEAR REGULATORY COMMISSION
NUCLEAR REGULATORY COMMISSION  
                                              REGION II
REGION II  
                                SAM NUNN ATLANTA FEDERAL CENTER
SAM NUNN ATLANTA FEDERAL CENTER
                                61 FORSYTH STREET, SW, SUITE 23T85
61 FORSYTH STREET, SW, SUITE 23T85  
                                    ATLANTA, GEORGIA 30303-8931
ATLANTA, GEORGIA 30303-8931  
                                          January 19, 2010
EA-09-321
Mr. Mano Nazar
January 19, 2010  
Executive Vice President and
   Chief Nuclear Officer
Florida Power & Light Company
EA-09-321  
P.O. Box 14000
Juno Beach, FL 33408-0420
Mr. Mano Nazar  
SUBJECT:       ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES
Executive Vice President and
                INSPECTION - INSPECTION REPORT 05000335/2009006 AND
   Chief Nuclear Officer  
                05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS
Florida Power & Light Company  
Dear Mr. Nazar:
P.O. Box 14000  
On September 4, 2009, U. S. Nuclear Regulatory Commission (NRC) completed an inspection
Juno Beach, FL 33408-0420  
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.
SUBJECT:  
The inspection examined activities conducted under your license as they relate to safety and
ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES  
compliance with the Commissions rules and regulations and with the conditions of your license.
INSPECTION - INSPECTION REPORT 05000335/2009006 AND  
The team reviewed selected procedures and records, observed activities, and interviewed
05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS  
personnel.
Section 4OA5 of the enclosed report discusses an event which occurred in October 2008 when
Dear Mr. Nazar:  
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
On September 4, 2009, U. S. Nuclear Regulatory Commission (NRC) completed an inspection  
CCW system inoperable. Two performance deficiencies were identified with this issue. The
at your St. Lucie Nuclear Plant, Units 1 and 2. The enclosed inspection report documents the  
first performance deficiency involved a common cause failure vulnerability of the CCW system.
preliminary inspection results which were discussed with Mr. Gordon Johnston on September 4,  
Specifically, a non-safety system failure could result in a common cause failure of both trains of
2009 and the final inspection results with Mr. Eric Katzman on December 10, 2009.  
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
The inspection examined activities conducted under your license as they relate to safety and  
source of the air in-leakage into the CCW system and take appropriate corrective actions
compliance with the Commissions rules and regulations and with the conditions of your license.
following the air intrusion event that occurred in October 2008. Further, the licensees corrective
The team reviewed selected procedures and records, observed activities, and interviewed  
action evaluation did not identify the common cause failure vulnerability discussed in the first
personnel.  
performance deficiency.
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
FP&L  
The findings associated with the common cause vulnerability and the inadequate corrective
2  
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 findings associated with the common cause vulnerability and the inadequate corrective  
the Significant Determination Process (SDP). The SDP analysis determined that the two
actions were assessed based on the best available information. The two issues were  
findings are potentially greater than very low safety significance because they potentially
preliminarily determined to be greater than Green findings using influencing assumptions and  
impacted the availability and thus the accident mitigation capability of the CCW system. These
the Significant Determination Process (SDP). The SDP analysis determined that the two  
findings do not represent a current safety concern because the containment IA system has been
findings are potentially greater than very low safety significance because they potentially  
isolated from the CCW system. Additionally, increased station sensitivity exists for recognizing
impacted the availability and thus the accident mitigation capability of the CCW system. These  
and responding in a timely manner if a similar air intrusion event were to occur.
findings do not represent a current safety concern because the containment IA system has been  
The performance deficiencies are documented in the enclosed report as two apparent violations
isolated from the CCW system. Additionally, increased station sensitivity exists for recognizing  
(AVs). The first performance deficiency is an AV of 10 CFR 50, Appendix B, Criterion III,
and responding in a timely manner if a similar air intrusion event were to occur.  
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
The performance deficiencies are documented in the enclosed report as two apparent violations  
system being susceptible to a common cause failure. The second performance deficiency is an
(AVs). The first performance deficiency is an AV of 10 CFR 50, Appendix B, Criterion III,  
AV of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for the failure to identify and
Design Control, for the failure to translate the design basis as specified in the license  
correct a condition adverse to quality following the air intrusion event into the CCW system that
application, into specifications, drawings, procedures, and instructions resulting in the CCW  
occurred in October 2008. These AVs are being considered for escalated enforcement action in
system being susceptible to a common cause failure. The second performance deficiency is an  
accordance with the NRC Enforcement Policy. The current Enforcement Policy is included on
AV of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, for the failure to identify and  
the NRC=s website at http://www.nrc.gov/reading-rm/adams.html.
correct a condition adverse to quality following the air intrusion event into the CCW system that  
In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our
occurred in October 2008. These AVs are being considered for escalated enforcement action in  
evaluation using the best available information and issue our final determination of safety
accordance with the NRC Enforcement Policy. The current Enforcement Policy is included on  
significance within 90 days of this letter. The significance determination process encourages an
the NRC=s website at http://www.nrc.gov/reading-rm/adams.html.  
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
In accordance with Inspection Manual Chapter (IMC) 0609, we intend to complete our  
are providing you an opportunity to: (1) present to the NRC your perspectives on the facts and
evaluation using the best available information and issue our final determination of safety  
assumptions used by the NRC to arrive at the finding and its significance at a Regulatory
significance within 90 days of this letter. The significance determination process encourages an  
Conference or (2) submit your position on the finding to the NRC in writing. If you request a
open dialogue between the staff and the licensee; however, the dialogue should not impact the  
Regulatory Conference, it should be held within 30 days of the receipt of this letter and we
timeliness of the staff=s final determination. Before we make a final decision on this matter, we  
encourage you to submit supporting documentation at least one week prior to the conference in
are providing you an opportunity to: (1) present to the NRC your perspectives on the facts and  
an effort to make the conference more efficient and effective. If a Regulatory Conference is
assumptions used by the NRC to arrive at the finding and its significance at a Regulatory  
held, it will be open for public observation. The NRC will also issue a press release to
Conference or (2) submit your position on the finding to the NRC in writing. If you request a  
announce the conference. If you decide to submit only a written response, such a submittal
Regulatory Conference, it should be held within 30 days of the receipt of this letter and we  
should be sent to the NRC within 30 days of the receipt of this letter.
encourage you to submit supporting documentation at least one week prior to the conference in  
Please contact Mr. Steve Rose at (404) 562-4609 or Mr. Binoy Desai at (404) 562-4519 within
an effort to make the conference more efficient and effective. If a Regulatory Conference is  
10 business days of the date of your receipt of this letter to notify the NRC of your intentions. If
held, it will be open for public observation. The NRC will also issue a press release to  
we have not heard from you within 10 days, we will continue with our significance determination
announce the conference. If you decide to submit only a written response, such a submittal  
and enforcement decisions and you will be advised by separate correspondence of the results
should be sent to the NRC within 30 days of the receipt of this letter.
of our deliberations on this matter.
Since the NRC has not made a final determination in this matter, a Notice of Violation is not
Please contact Mr. Steve Rose at (404) 562-4609 or Mr. Binoy Desai at (404) 562-4519 within  
being issued at this time. In addition, please be advised that the number and characterization of
10 business days of the date of your receipt of this letter to notify the NRC of your intentions. If  
the AVs violations may change as a result of further NRC review.
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
FP&L  
In addition, this report documents two NRC-identified findings of very low safety significance
3  
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
In addition, this report documents two NRC-identified findings of very low safety significance  
Policy because of their very low safety significance and because they were entered into your
which were determined to be violations of NRC requirements. The NRC is treating these two  
corrective action program. If you contest these NCVs, you should provide a response within 30
violations as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement  
days of the date of this inspection report, with the basis for your denial, to the Nuclear
Policy because of their very low safety significance and because they were entered into your  
Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with
corrective action program. If you contest these NCVs, you should provide a response within 30  
copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United
days of the date of this inspection report, with the basis for your denial, to the Nuclear  
States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident
Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with  
inspector at the St. Lucie Nuclear Plant. In addition, if you disagree with the characterization of
copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United  
any finding in this report, you should provide a response within 30 days of the date of this
States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident  
inspection report, with the basis for your disagreement, to the Regional Administrator, Region II,
inspector at the St. Lucie Nuclear Plant. In addition, if you disagree with the characterization of  
and the NRC Resident Inspector at the St. Lucie Nuclear Plant. The information you provide will
any finding in this report, you should provide a response within 30 days of the date of this  
be considered in accordance with the Inspection Manual Chapter 0305.
inspection report, with the basis for your disagreement, to the Regional Administrator, Region II,  
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
and the NRC Resident Inspector at the St. Lucie Nuclear Plant. The information you provide will  
enclosure, and your response (if any) will be available electronically for public inspection in the
be considered in accordance with the Inspection Manual Chapter 0305.  
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
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its  
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
enclosure, and your response (if any) will be available electronically for public inspection in the  
                                              Sincerely,
NRC Public Document Room or from the Publicly Available Records (PARS) component of the  
                                              /RA/
NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at  
                                              Kriss M. Kennedy, Director
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).  
                                              Division of Reactor Safety
Enclosure: Inspection Report 05000335/2009006, 05000389/2009006
Sincerely,  
                w/Attachment: Supplemental Information
Docket Nos.: 50-335, 50-389
License Nos.: DPR-67 and NPF-16
cc w/encl: (See page 4)
/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)


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)
xx  PUBLICLY AVAILABLE
G  NON-PUBLICLY AVAILABLE
G  SENSITIVE
      xx  NON-SENSITIVE
ADAMS:  G Yes
ACCESSION NUMBER:_________________________
xxG  SUNSI REVIEW COMPLETE
OFFICE
RII:DRS
RII:DRS
RII:DRS
RII:DRP
CONTRACTOR
CONTRACTOR
RII:DRP
SIGNATURE
RA
RA
RA
RA
RA
RA
RA
NAME
SROSE
RMOORE
JHAMMAN
RTAYLOR
MSHYLAMBERG NDELIAGRECA MSYKES
DATE
11/30/2009
11/19/2009
1/12/2010
11/20/2009
11/18/2009
11/5/2009
1/13/2010
E-MAIL COPY?
    YES
NO  YES
NO  YES
NO  YES
NO  YES
NO    YES
NO    YES
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
OFFICIAL
RECORD
COPY
DOCUMENT
NAME:
S:\\DRS\\ENG
BRANCH
1\\BRANCH
INSPECTION
FILES\\CDBI
INSPECTIONS\\CDBI INSPECTIONS\\INSP REPORTS\\CDBI FINAL INSPECTION REPORTS\\REV 1 ST LUCIE 2009006 CDBI
REPORT (SDR).DOC


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


FP&L                                     5
FP&L  
Letter to Mano Nazar from Kriss Kennedy dated January 19, 2010.
5  
SUBJECT:       ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES
              INSPECTION - INSPECTION REPORT 05000335/2009006 AND
Letter to Mano Nazar from Kriss Kennedy dated January 19, 2010.  
              05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS
Distribution w/encl:
SUBJECT:  
C. Evans, RII
ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES  
L. Slack, RII
INSPECTION - INSPECTION REPORT 05000335/2009006 AND  
OE Mail
05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS  
RIDSNRRDIRS
PUBLIC
Distribution w/encl:  
RidsNrrPMStLucie Resource
C. Evans, RII
L. Slack, RII
OE Mail
RIDSNRRDIRS  
PUBLIC  
RidsNrrPMStLucie Resource  


          U. S. NUCLEAR REGULATORY COMMISSION
                            REGION II
Enclosure
Docket Nos.: 50-335, 50-389
U. S. NUCLEAR REGULATORY COMMISSION  
License Nos.: DPR-67 and NPF-16
Report Nos.: 05000335/2009006, 05000389/2009006
REGION II  
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)
Docket Nos.: 50-335, 50-389  
                R. Moore, Senior Reactor Inspector
                J. Hamman, Reactor Inspector
                R. Taylor, Senior Reactor Inspector
                M. Shylamberg, Contractor
                N. Della Greca, Contractor
License Nos.: DPR-67 and NPF-16  
Approved by: Binoy Desai, Chief
                Engineering Branch 1
                Division of Reactor Safety
                                                            Enclosure
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  


                              SUMMARY OF FINDINGS
IR 05000335/2009006, 05000389/2009006; 8/3/2009 - 9/4/2009; St. Lucie Nuclear
Enclosure
Plant, Units 1 and 2; NRC Component Design Bases Inspection.
SUMMARY OF FINDINGS  
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)
IR 05000335/2009006, 05000389/2009006; 8/3/2009 - 9/4/2009; St. Lucie Nuclear  
were identified during this inspection and were classified as non-cited violations. Also,
Plant, Units 1 and 2; NRC Component Design Bases Inspection.  
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,
This inspection was conducted by a team of four NRC inspectors from the Region II  
Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for
office, and two NRC contract inspectors. Two findings of very low significance (Green)  
which the SDP does not apply may be Green or be assigned a severity level after NRC
were identified during this inspection and were classified as non-cited violations. Also,  
management review. The NRC's program for overseeing the safe operation of
two apparent violations (AV) with potential safety significance greater than Green were  
commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight
identified. The significance of most findings is indicated by their color (Green, White,  
Process, (ROP) Revision 4, dated December 2006.
Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for  
Cornerstone: Mitigating Systems
which the SDP does not apply may be Green or be assigned a severity level after NRC  
*   Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion
management review. The NRC's program for overseeing the safe operation of  
    III, Design Control, for failure to translate the design basis as specified in the license
commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight  
    application into specifications, drawings, procedures, and instructions. The licensee
Process, (ROP) Revision 4, dated December 2006.  
    did not ensure that the component cooling water (CCW) surge tank design included
    adequate overpressure protection for all procedurally allowed configurations as
Cornerstone: Mitigating Systems  
    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
Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion  
    of discharge. The team concluded that stop valve V6466 was an intervening stop
III, Design Control, for failure to translate the design basis as specified in the license  
    valve for the CCW surge tank vent path to the chemical drain tank (CDT). The issue
application into specifications, drawings, procedures, and instructions. The licensee  
    was entered in the licensees corrective action program as condition report (CR)
did not ensure that the component cooling water (CCW) surge tank design included  
    2009-23473. Immediate licensee corrective actions included verification that the
adequate overpressure protection for all procedurally allowed configurations as  
    valve was in its open position and the implementation of administrative controls to
required by the applicable ASME Boiler and Pressure Vessel Code, Section VIII,  
    maintain the valve open.
Division 1. The code requires that no intervening stop valves be between the vessel  
    This finding is associated with the Mitigating Systems Cornerstone attribute of
and its protective device or devices or between the protective devices and the point  
    Design Control, i.e. initial design, was determined to be more than minor because it
of discharge. The team concluded that stop valve V6466 was an intervening stop  
    impacted the cornerstone objective to ensure the availability, reliability, and capability
valve for the CCW surge tank vent path to the chemical drain tank (CDT). The issue  
    of systems that respond to initiating events to prevent undesirable consequences.
was entered in the licensees corrective action program as condition report (CR)  
    The team determined that if left uncorrected, this design deficiency had the potential
2009-23473. Immediate licensee corrective actions included verification that the  
    to impact the operability of safety-related systems and, thus, become a more
valve was in its open position and the implementation of administrative controls to  
    significant safety concern in that a closed intervening valve had the potential for
maintain the valve open.  
    overpressurizing the CCW surge tank. The team assessed this finding for
    significance in accordance with NRC Manual Chapter 0609, Appendix A, Attachment
This finding is associated with the Mitigating Systems Cornerstone attribute of  
    1, Significance Determination Process (SDP) for Reactor Inspection Findings for At-
Design Control, i.e. initial design, was determined to be more than minor because it  
    Power Situations, and determined that it was of very low safety significance (Green),
impacted the cornerstone objective to ensure the availability, reliability, and capability  
    in that no actual loss of safety system function was identified. The team reviewed
of systems that respond to initiating events to prevent undesirable consequences.
    the finding for cross-cutting aspects and concluded that this finding did not have an
The team determined that if left uncorrected, this design deficiency had the potential  
    associated cross-cutting aspect because the design of the CCW surge tank relief
to impact the operability of safety-related systems and, thus, become a more  
    was established in an original plant design, and therefore, was not representative of
significant safety concern in that a closed intervening valve had the potential for  
    current licensee performance. [Section 1R21.2.2]
overpressurizing the CCW surge tank. The team assessed this finding for  
                                                                                      Enclosure
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]  


                                            3
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
Enclosure
  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
Green. The inspectors identified a finding involving a violation of 10 CFR 50,  
  analysis incorporated incorrect design input specifications. The issue was entered in
Appendix B, Criterion III, Design Control, for the licensees failure to maintain the  
  the licensees corrective action program as CR 2009-24517. Licensee corrective
safety-related 125V DC system design basis information consistent with the plant  
  actions included incorporating the correct design input and specifications by revising
configuration. Specifically, a revision to the Unit 1, safety-related 125V DC system  
  the calculations.
analysis incorporated incorrect design input specifications. The issue was entered in  
  The finding was more than minor because it was associated with the Mitigating
the licensees corrective action program as CR 2009-24517. Licensee corrective  
  Systems Cornerstone attribute of Design Control. It impacted the cornerstone
actions included incorporating the correct design input and specifications by revising  
  objective because if left uncorrected, it had the potential to lead to a more significant
the calculations.  
  safety concern in that future design activity or operability assessments would
  assume the lower voltage (100V DC vs. actual 105V DC) value acceptable for
The finding was more than minor because it was associated with the Mitigating  
  assuring the adequacy of voltage to the safety-related inverters. The team assessed
Systems Cornerstone attribute of Design Control. It impacted the cornerstone  
  this finding for significance in accordance with NRC Manual Chapter 0609, using the
objective because if left uncorrected, it had the potential to lead to a more significant  
  Phase I SDP worksheet for mitigating systems and determined that the finding was
safety concern in that future design activity or operability assessments would  
  of very low safety significance (Green) since it was a design deficiency determined
assume the lower voltage (100V DC vs. actual 105V DC) value acceptable for  
  not to have resulted in a loss of safety function. This finding has a cross-cutting
assuring the adequacy of voltage to the safety-related inverters. The team assessed  
  aspect in the area of human performance because the licensee failed to ensure that
this finding for significance in accordance with NRC Manual Chapter 0609, using the  
  procedures (specifically ENG-QI 1.5) were available and adequate to assure nuclear
Phase I SDP worksheet for mitigating systems and determined that the finding was  
  safety (specifically, complete, accurate and up-to-date design documentation):
of very low safety significance (Green) since it was a design deficiency determined  
  H.2(c). [Section 1R21.2.20]
not to have resulted in a loss of safety function. This finding has a cross-cutting  
* TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design
aspect in the area of human performance because the licensee failed to ensure that  
  Control, for the licensees failure to identify that the CCW system met its license
procedures (specifically ENG-QI 1.5) were available and adequate to assure nuclear  
  specifications related to common cause failure vulnerabilities. Specifically, a non-
safety (specifically, complete, accurate and up-to-date design documentation):  
  safety system failure (i.e. waste gas compressor aftercoolers affecting both units, or
H.2(c). [Section 1R21.2.20]  
  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
TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design  
  evaluate the past operability of the CCW system during the air intrusion event.
Control, for the licensees failure to identify that the CCW system met its license  
  Licensee corrective actions included isolating the CCW system from the containment
specifications related to common cause failure vulnerabilities. Specifically, a non-
  IA compressors.
safety system failure (i.e. waste gas compressor aftercoolers affecting both units, or  
  The finding was determined to be more than minor because if left uncorrected, it
containment IA compressors affecting Unit 1 only) could result in a common cause  
  could affect the availability, reliability and capability of a safety system to perform its
failure of both trains of a safety system (i.e. CCW system). The issue was entered  
  intended safety function. Specifically, with this vulnerability, a failure of the waste
into the licensees corrective action program as CR 2009-22929 with actions to  
  gas aftercooler (both units) or a failure of the containment IA compressors (Unit 1
evaluate the past operability of the CCW system during the air intrusion event.
  only) could cause air intrusion into the CCW system and lead to a loss of CCW
Licensee corrective actions included isolating the CCW system from the containment  
  event, therefore, failing to ensure that adequate cooling would be available or
IA compressors.  
  maintained to essential equipment used to mitigate design bases accidents. The
  finding was assessed for significance in accordance with NRC Manual Chapter 0609,
The finding was determined to be more than minor because if left uncorrected, it  
  using the Phase I and Phase II SDP worksheets for mitigating systems. It was
could affect the availability, reliability and capability of a safety system to perform its  
  determined that a Phase III analysis was required since this finding represented a
intended safety function. Specifically, with this vulnerability, a failure of the waste  
  potential loss of safety system function for multiple trains which was not addressed
gas aftercooler (both units) or a failure of the containment IA compressors (Unit 1  
  by the Phase II pre-solved tables/worksheets. Based on the Phase III SDP, the
only) could cause air intrusion into the CCW system and lead to a loss of CCW  
  finding was preliminarily determined to be greater than Green. The team reviewed
event, therefore, failing to ensure that adequate cooling would be available or  
  the finding for cross-cutting aspect and concluded that this finding did not have an
maintained to essential equipment used to mitigate design bases accidents. The  
                                                                                    Enclosure
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  


                                            4
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
Enclosure
  current licensee performance. [Section 4OA5]
associated cross-cutting aspect because the design of the CCW system was  
* TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI, Corrective
established in an original plant design, and therefore, was not representative of  
  Action, for the licensees failure to implement adequate corrective actions associated
current licensee performance. [Section 4OA5]  
  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
TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI, Corrective  
  CR 2009-25209 to address the ineffective corrective actions for the air intrusion
Action, for the licensees failure to implement adequate corrective actions associated  
  event. Licensee corrective actions included isolating the CCW system from the
with the CCW air intrusion event that occurred in October, 2008. The corrective  
  containment IA compressors.
actions were inadequate in that the licensee failed to identify and correct the cause  
  The finding was determined to be more than minor because it affected the
of air intrusion. The issue was entered in the licensees corrective action program as  
  availability, reliability and capability of a safety system to perform its intended safety
CR 2009-25209 to address the ineffective corrective actions for the air intrusion  
  function. Specifically, without knowing the leak path from the containment IA
event. Licensee corrective actions included isolating the CCW system from the  
  compressors to the CCW system, the licensee could not ensure that adequate
containment IA compressors.  
  cooling would be available or maintained to essential equipment used to mitigate
  design bases accidents. The finding was assessed for significance in accordance
The finding was determined to be more than minor because it affected the  
  with NRC Manual Chapter 0609, using the Phase I and Phase II SDP worksheets for
availability, reliability and capability of a safety system to perform its intended safety  
  mitigating systems. It was determined that a Phase III analysis was required since
function. Specifically, without knowing the leak path from the containment IA  
  this finding represented a loss of safety system function for multiple trains which was
compressors to the CCW system, the licensee could not ensure that adequate  
  not addressed by the Phase II pre-solved tables/worksheets. Based on the Phase III
cooling would be available or maintained to essential equipment used to mitigate  
  SDP, the finding was preliminarily determined to be greater than Green. This finding
design bases accidents. The finding was assessed for significance in accordance  
  was determined to have a cross-cutting aspect in the area of Human Performance,
with NRC Manual Chapter 0609, using the Phase I and Phase II SDP worksheets for  
  Decision Making, specifically H.1(a). [Section 4OA5]
mitigating systems. It was determined that a Phase III analysis was required since  
                                                                                  Enclosure
this finding represented a loss of safety system function for multiple trains which was  
not addressed by the Phase II pre-solved tables/worksheets. Based on the Phase III  
SDP, the finding was preliminarily determined to be greater than Green.   This finding  
was determined to have a cross-cutting aspect in the area of Human Performance,  
Decision Making, specifically H.1(a). [Section 4OA5]  


                                        REPORT DETAILS
1.   REACTOR SAFETY
Enclosure
      Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
REPORT DETAILS  
1R21 Component Design Bases Inspection (71111.21)
.1   Inspection Sample Selection Process
1.  
      The team selected risk significant components and operator actions for review using
REACTOR SAFETY  
      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
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity  
      components, six operator actions, and five operating experience items. Additionally, the
      team reviewed one permanent plant modification by performing activities identified in IP
1R21 Component Design Bases Inspection (71111.21)  
      71111.17, Evaluations of Changes, Tests, or Experiments and Permanent Plant
      Modifications.
.1  
      The team performed a margin assessment and detailed review of the selected risk-
Inspection Sample Selection Process  
      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
The team selected risk significant components and operator actions for review using  
      material condition issues. Equipment reliability issues were also considered in the
information contained in the licensees Probabilistic Risk Assessment (PRA). In general,  
      selection of components for detailed review. These reliability issues included review of
this included components and operator actions that had a risk achievement worth factor  
      items related to performance and surveillance test failures, corrective actions due to
greater than 1.3 or Birnbaum value greater than 1 X10-6. The sample included 20  
      repeat maintenance, maintenance rule (a)1 status, Regulatory Issue Summary (RIS) 05-
components, six operator actions, and five operating experience items. Additionally, the  
      020 (formerly Generic Letter (GL) 91-18) conditions, NRC resident inspector input of
team reviewed one permanent plant modification by performing activities identified in IP  
      problem equipment, system health reports, industry operating experience and licensee
71111.17, Evaluations of Changes, Tests, or Experiments and Permanent Plant  
      problem equipment lists. Consideration was also given to the uniqueness and complexity
Modifications.  
      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.
The team performed a margin assessment and detailed review of the selected risk-
.2   Results of Detailed Reviews
significant components to verify that the design bases had been correctly implemented  
.2.1 Component Cooling Water (CCW) Pumps 1A/1B/1C
and maintained. This design margin assessment considered original design issues,  
   a. Inspection Scope
margin reductions due to modifications, or margin reductions identified as a result of  
      The team reviewed the design bases documents (DBD), related design basis
material condition issues. Equipment reliability issues were also considered in the  
      documentation, drawings, technical specifications (TS), and the final safety analysis
selection of components for detailed review. These reliability issues included review of  
      report (FSAR) to identify design, maintenance, and operational requirements for the
items related to performance and surveillance test failures, corrective actions due to  
      CCW pumps. The team reviewed the system configuration and design calculations to
repeat maintenance, maintenance rule (a)1 status, Regulatory Issue Summary (RIS) 05-
      verify that adequate net positive suction head (NPSH) would be available during
020 (formerly Generic Letter (GL) 91-18) conditions, NRC resident inspector input of  
      accident conditions. Maintenance history, as demonstrated by system health reports,
problem equipment, system health reports, industry operating experience and licensee  
      corrective maintenance documentation, condition reports (CRs), and surveillance test
problem equipment lists. Consideration was also given to the uniqueness and complexity  
      results, were reviewed to verify the design bases had been maintained; potential
of the design, operating experience, and the available defense in depth margins. An  
      degradation was being monitored; and that identified degradation or malfunctions had
overall summary of the reviews performed and the specific inspection findings identified  
      been adequately addressed. The team reviewed normal, abnormal, and emergency
is included in the following sections of the report.  
                                                                                        Enclosure
.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  


                                                6
6  
      operating procedures to verify correct implementation of design bases. The team
      verified that the equipment periodic maintenance performed was consistent with vendor
Enclosure
      recommendations. Additionally, the team conducted a field walkdown of the CCW
operating procedures to verify correct implementation of design bases. The team  
      pumps with the licensee staff to assess observable material condition and to verify that
verified that the equipment periodic maintenance performed was consistent with vendor  
      the installed configuration was consistent with the design basis and plant drawings. The
recommendations. Additionally, the team conducted a field walkdown of the CCW  
      team reviewed voltage drop calculations to confirm that the voltage available at the
pumps with the licensee staff to assess observable material condition and to verify that  
      motor terminals as well as at the circuit breakers was adequate to ensure that the pumps
the installed configuration was consistent with the design basis and plant drawings. The  
      can perform their safety function when called upon. Additionally, the team verified that
team reviewed voltage drop calculations to confirm that the voltage available at the  
      the horsepower rating of the motors were correctly identified in the load flow analysis
motor terminals as well as at the circuit breakers was adequate to ensure that the pumps  
      and that adequate protection was provided for the motors. The team reviewed control
can perform their safety function when called upon. Additionally, the team verified that  
      wiring diagrams to confirm that the operation of the pumps conformed to their intended
the horsepower rating of the motors were correctly identified in the load flow analysis  
      function.
and that adequate protection was provided for the motors. The team reviewed control  
  b. Findings
wiring diagrams to confirm that the operation of the pumps conformed to their intended  
      No findings of significance were identified; however, see section 4OA5 for two findings
function.  
      related to the CCW system.
.2.2 Component Cooling Water Surge Tank
  b.  
  a. Inspection Scope
Findings  
    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
No findings of significance were identified; however, see section 4OA5 for two findings  
    baffle location and height, and tank implosion and overpressure protection were reviewed
related to the CCW system.  
    and compared to as-built configuration. The team also reviewed all CCW system
    operating conditions to verify that design, maintenance, and operational requirements
.2.2  
    were appropriate. The CCW flow assumptions in the FSAR accident analysis were also
Component Cooling Water Surge Tank  
    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
  a.  
    ASME requirements. Maintenance, corrective actions, and design change history were
Inspection Scope  
    reviewed to assess potential component degradation and subsequent impacts on design
    margins.
For the CCW surge tank the team reviewed DBDs, Technical Specifications, FSAR,  
  b. Findings
calculations, and drawings. Specific design requirements for the CCW surge tank levels,  
      Introduction: The inspectors identified a finding of very low safety significance (Green)
tank leakage and make up rate, minimum level vs. NPSH allowed and vortex limits, tank  
      involving a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the
baffle location and height, and tank implosion and overpressure protection were reviewed  
      licensees failure to translate the design basis as specified in the license application, into
and compared to as-built configuration. The team also reviewed all CCW system  
      specifications, drawings, procedures, and instructions. Specifically, the licensees failure
operating conditions to verify that design, maintenance, and operational requirements  
      to assure that the CCW surge tank design included adequate overpressure protection for
were appropriate. The CCW flow assumptions in the FSAR accident analysis were also  
      all configurations allowed by plant procedures, as required by the applicable ASME
reviewed to verify that the surge tank was capable of performing the intended safety  
      Boiler and Pressure Vessel Code, Section VIII, Division 1, was identified by the
functions. Calculations were also reviewed to verify that the surge tank met applicable  
      inspectors as a performance deficiency.
ASME requirements. Maintenance, corrective actions, and design change history were  
      Description: The review of the Unit 1 CCW surge tanks design and operation identified
reviewed to assess potential component degradation and subsequent impacts on design  
      that the tank pressure relief required by the ASME Code (ASME Section VIII) was
margins.  
      provided via a 2-inch vent line. This vent line was routed to a diverting air-operated
                                                                                        Enclosure
  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  


                                            7
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
Enclosure
vent/overflow to the liquid waste management system chemical drain tank (CDT) 1A. A
valve, RCV-14-1. This valve was normally open to atmosphere; however, in the event of  
similar re-alignment would take place on a loss of instrument air. The CDT 1A was a
high radiation, this valve re-aligns the relief path from the atmosphere and diverts the  
closed tank and was vented to a sump pit by a 1-1/2 line. A maintenance valve, V6466,
vent/overflow to the liquid waste management system chemical drain tank (CDT) 1A. A  
was installed between the diverting air-operated valve RCV-14-1 and CDT 1A. A similar
similar re-alignment would take place on a loss of instrument air. The CDT 1A was a  
configuration existed for Unit 2.
closed tank and was vented to a sump pit by a 1-1/2 line. A maintenance valve, V6466,  
ASME Section VIII, Division 1, 1971 Edition, paragraph UG-134(e) states, There shall
was installed between the diverting air-operated valve RCV-14-1 and CDT 1A. A similar  
be no intervening stop valves between the vessel and its protective device or devices or
configuration existed for Unit 2.  
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,
ASME Section VIII, Division 1, 1971 Edition, paragraph UG-134(e) states, There shall  
which states the minimum code requirements for Quality Group C pressure vessels must
be no intervening stop valves between the vessel and its protective device or devices or  
comply with ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. The
between the protective devices and the point of discharge The requirement to  
Quality Group C designation for the safety-related portion of the CCW system was
comply with the ASME Code requirements was based on Unit 1 FSAR Table 3.2-2,  
provided in the Unit 1 DBD for CCW. The Unit 1 CCW Tank was procured per
which states the minimum code requirements for Quality Group C pressure vessels must  
specification FLO-8770-764, originally issued on October 31, 1971. Therefore, the
comply with ASME Boiler and Pressure Vessel Code, Section VIII, Division 1. The  
inspectors concluded that ASME Section VIII, Division 1, 1971 edition applied.
Quality Group C designation for the safety-related portion of the CCW system was  
The team concluded that valve V6466 was an intervening stop valve for the CCW Surge
provided in the Unit 1 DBD for CCW. The Unit 1 CCW Tank was procured per  
Tank vent path to the CDT. The licensee issued CRs 2009-25276 and 2009-23473 to
specification FLO-8770-764, originally issued on October 31, 1971. Therefore, the  
evaluate this condition. The licensees review determined that valve V6466 was a
inspectors concluded that ASME Section VIII, Division 1, 1971 edition applied.  
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
The team concluded that valve V6466 was an intervening stop valve for the CCW Surge  
and V6466 that would provide an alternate relief path should valve V6466 be closed.
Tank vent path to the CDT. The licensee issued CRs 2009-25276 and 2009-23473 to  
The licensees review of records for the past 10 years identified that for Unit 1, valve
evaluate this condition. The licensees review determined that valve V6466 was a  
V6466 was never closed. The licensee identified that for Unit 2, the valve had been
normally open valve. Additionally, there were a number of floor drains (although not  
closed in the past, however, during that time, the drains were rerouted to an alternate
formally maintained clear of blockages) that tie in the header between valves RCV-14-1  
tank, thus providing the required relief path. The team concluded from this information
and V6466 that would provide an alternate relief path should valve V6466 be closed.
that this design deficiency did not represent an actual loss of safety system function.
The licensees review of records for the past 10 years identified that for Unit 1, valve  
The team reviewed the finding for cross-cutting and concluded that this finding did not
V6466 was never closed. The licensee identified that for Unit 2, the valve had been  
have an associated cross-cutting aspect because the design of the CCW surge tank
closed in the past, however, during that time, the drains were rerouted to an alternate  
relief was established in an original plant design, therefore, not representative of current
tank, thus providing the required relief path. The team concluded from this information  
licensee performance.
that this design deficiency did not represent an actual loss of safety system function.  
Analysis: The licensees failure to assure the CCW surge tank design included
adequate overpressure protection as required by the applicable ASME Boiler and
The team reviewed the finding for cross-cutting and concluded that this finding did not  
Pressure Vessel Code was identified as a performance deficiency. This finding,
have an associated cross-cutting aspect because the design of the CCW surge tank  
associated with the Mitigating Systems Cornerstone attribute of Design Control, i.e.
relief was established in an original plant design, therefore, not representative of current  
initial design, was determined to be more than minor because it impacted the
licensee performance.  
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences. The team determined
Analysis: The licensees failure to assure the CCW surge tank design included  
that if left uncorrected, this design deficiency had the potential to impact the operability
adequate overpressure protection as required by the applicable ASME Boiler and  
of safety-related systems and, thus, become a more significant safety concern.
Pressure Vessel Code was identified as a performance deficiency. This finding,  
Specifically, during an overpressure event, if intervening valve V6466 was shut and the
associated with the Mitigating Systems Cornerstone attribute of Design Control, i.e.  
floor drain lines clogged, the CCW surge tank vent path to the CDT would be obstructed
initial design, was determined to be more than minor because it impacted the  
to the point that a loss of CCW surge tank could occur, therefore, increasing the
cornerstone objective to ensure the availability, reliability, and capability of systems that  
likelihood of a loss of CCW. The team assessed this finding for significance in
respond to initiating events to prevent undesirable consequences. The team determined  
                                                                                    Enclosure
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  


                                              8
8  
    accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, Significance
    Determination Process (SDP) for Reactor Inspection Findings for At-Power Situations,
Enclosure
    and determined that it was of very low safety significance (Green), in that no actual loss
accordance with NRC Manual Chapter 0609, Appendix A, Attachment 1, Significance  
    of safety system function was identified. The team concluded that this finding did not
Determination Process (SDP) for Reactor Inspection Findings for At-Power Situations,  
    have an associated cross-cutting aspect because the performance deficiency was not
and determined that it was of very low safety significance (Green), in that no actual loss  
    reflective of current plant performance. The design of the CCW surge tank relief was
of safety system function was identified. The team concluded that this finding did not  
    established during original plant design; and the last design change associated with the
have an associated cross-cutting aspect because the performance deficiency was not  
    CCW surge tank was in 2001.
reflective of current plant performance. The design of the CCW surge tank relief was  
    Enforcement: 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in
established during original plant design; and the last design change associated with the  
    part, that measures shall be established to assure that applicable regulatory
CCW surge tank was in 2001.  
    requirements and the design basis are correctly translated into specifications. Contrary
    to the above, the licensee failed to assure that applicable regulatory requirements and
Enforcement: 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in  
    the design bases were correctly translated into actual plant specifications. The installed
part, that measures shall be established to assure that applicable regulatory  
    CCW surge tank pressure relief protection did not meet the Code requirements
requirements and the design basis are correctly translated into specifications. Contrary  
    described in the Unit 1 FSAR Table 3.2-2. The FSAR required that the minimum code
to the above, the licensee failed to assure that applicable regulatory requirements and  
    requirements for Quality Group C pressure vessels to be ASME Boiler and Pressure
the design bases were correctly translated into actual plant specifications. The installed  
    Vessel Code, Section VIII, Division 1. Specifically, ASME Boiler and Pressure Vessel
CCW surge tank pressure relief protection did not meet the Code requirements  
    Code, Section VIII, Division 1 requirements for the overpressure protection for the CCW
described in the Unit 1 FSAR Table 3.2-2. The FSAR required that the minimum code  
    surge tank were not properly implemented. This design deficiency was an original plant
requirements for Quality Group C pressure vessels to be ASME Boiler and Pressure  
    design and has existed since the operating licenses were issued. Because this violation
Vessel Code, Section VIII, Division 1. Specifically, ASME Boiler and Pressure Vessel  
    was of very low safety significance (Green) and it was entered into the licensees
Code, Section VIII, Division 1 requirements for the overpressure protection for the CCW  
    corrective action program as CR 2009-25276 and CR 2009-23473, this violation is being
surge tank were not properly implemented. This design deficiency was an original plant  
    treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV
design and has existed since the operating licenses were issued. Because this violation  
    05000335,389/2009006-01, Failure to Meet the ASME Boiler and Pressure Vessel
was of very low safety significance (Green) and it was entered into the licensees  
    Code, Section VIII, Division 1 Requirements for the Overpressure Protection for the
corrective action program as CR 2009-25276 and CR 2009-23473, this violation is being  
    CCW Surge Tank.
treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV  
.2.3 Instrument Air Emergency Cooling System
05000335,389/2009006-01, Failure to Meet the ASME Boiler and Pressure Vessel  
  a. Inspection Scope
Code, Section VIII, Division 1 Requirements for the Overpressure Protection for the  
    The team reviewed the drawings, TS, and the FSAR to identify the design, maintenance,
CCW Surge Tank.  
    and operational requirements for the instrument air (IA) emergency cooling system. The
    team reviewed the system configuration and normal, abnormal, and emergency
.2.3  
    operating procedures to verify correct implementation of the design bases. Maintenance
Instrument Air Emergency Cooling System  
    history, as demonstrated by system health reports, corrective maintenance
    documentation, CRs, and surveillance test results, was reviewed to verify that the design
  a.  
    bases had been maintained and correctly implemented; potential degradation was being
Inspection Scope  
    monitored; and that identified degradation or malfunctions had been adequately
    addressed. The team verified that the equipment periodic maintenance performed was
The team reviewed the drawings, TS, and the FSAR to identify the design, maintenance,  
    consistent with vendor recommendations. Additionally, the team conducted a field
and operational requirements for the instrument air (IA) emergency cooling system. The  
    walkdown of the IA emergency cooling system with the licensee staff to assess
team reviewed the system configuration and normal, abnormal, and emergency  
    observable material condition and to verify that the installed configuration was consistent
operating procedures to verify correct implementation of the design bases. Maintenance  
    with the design basis and plant drawings.
history, as demonstrated by system health reports, corrective maintenance  
                                                                                      Enclosure
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
9  
b. Findings
  Introduction: An unresolved item (URI) was identified related to the performance
Enclosure
  monitoring of the IA emergency cooling system. The team determined that the
  b.  
  performance monitoring did not provide reasonable assurance that the system was
Findings  
  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
Introduction: An unresolved item (URI) was identified related to the performance  
  to be in the scope of the maintenance rule (MR), 10 CFR 50.65(a)(1), Requirements for
monitoring of the IA emergency cooling system. The team determined that the  
  Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because it is
performance monitoring did not provide reasonable assurance that the system was  
  included in the St. Lucie emergency operating procedures.
capable of fulfilling its intended function. This failure to monitor the performance of the  
  Description: The IA emergency cooling system is an alternate source of cooling for IA
IA emergency cooling system was a performance deficiency. The system was identified  
  compressors A and B. The system is a small, closed cooling system with a pump, head
to be in the scope of the maintenance rule (MR), 10 CFR 50.65(a)(1), Requirements for  
  tank, fan cooled radiator and connecting piping and valves to the IA compressors. The
Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because it is  
  normal cooling water to the compressors is provided by the turbine cooling water (TCW)
included in the St. Lucie emergency operating procedures.  
  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
Description: The IA emergency cooling system is an alternate source of cooling for IA  
  with vital power so that the compressors can be manually loaded in accordance with
compressors A and B. The system is a small, closed cooling system with a pump, head  
  1[2]-EOP-09, Loss of Offsite Power, Rev. 38.
tank, fan cooled radiator and connecting piping and valves to the IA compressors. The  
  During the inspection, the team requested design, maintenance, or operational
normal cooling water to the compressors is provided by the turbine cooling water (TCW)  
  documentation that would provide reasonable assurance that the emergency cooling
system which does not have power available after a loss of offsite power (LOOP)  
  system could perform its intended function of providing adequate cooling for IA
accident. These IA compressors and the emergency cooling system pump are provided  
  compressors A and B during a LOOP event. There were no documented specifications,
with vital power so that the compressors can be manually loaded in accordance with  
  analysis, or testing available to verify the adequacy of the emergency cooling water
1[2]-EOP-09, Loss of Offsite Power, Rev. 38.  
  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
During the inspection, the team requested design, maintenance, or operational  
  testing did not verify the system adequacy or provide the capability to identify potential
documentation that would provide reasonable assurance that the emergency cooling  
  degradation of the equipment. For example, Procedure OSP-69.13A, ESF-18 Month
system could perform its intended function of providing adequate cooling for IA  
  Surveillance for SIAS/CIS/CSAS, Rev. 2, aligned the IA emergency cooling system to
compressors A and B during a LOOP event. There were no documented specifications,  
  the 2B IA compressor; however, the test configuration was in parallel with the higher
analysis, or testing available to verify the adequacy of the emergency cooling water  
  capacity 2C IA compressor and, therefore, it was not possible to determine if the 2B IA
system to support continued operation of the IA compressors. The team reviewed the  
  compressor was loaded and the emergency cooling system was capable of sustaining
routine testing performed on the emergency cooling system and concluded that this  
  loaded compressor operation. Procedure 2-0330020, Appendix H, Instrument Air
testing did not verify the system adequacy or provide the capability to identify potential  
  Emergency Cooling Test, Rev. 56, required the recirculation pump to be run for 30
degradation of the equipment. For example, Procedure OSP-69.13A, ESF-18 Month  
  minutes but stated that starting the IA compressor was an option. The licensee did not
Surveillance for SIAS/CIS/CSAS, Rev. 2, aligned the IA emergency cooling system to  
  provide past test information that demonstrated the IA compressor was run or loaded
the 2B IA compressor; however, the test configuration was in parallel with the higher  
  during this routine test. The inspectors concluded that the routine testing performed
capacity 2C IA compressor and, therefore, it was not possible to determine if the 2B IA  
  verified the flow path to the unloaded compressor but did not verify that the cooling
compressor was loaded and the emergency cooling system was capable of sustaining  
  system was capable of supporting sustained operation of the compressor. The licensee
loaded compressor operation.   Procedure 2-0330020, Appendix H, Instrument Air  
  documented this issue in CR 2009-22766 and planned to perform a formal test of the
Emergency Cooling Test, Rev. 56, required the recirculation pump to be run for 30  
  system to demonstrate its capabilities.
minutes but stated that starting the IA compressor was an option. The licensee did not  
  The team noted that the IA system at St. Lucie was a non-safety related system. Station
provide past test information that demonstrated the IA compressor was run or loaded  
  design was that air-operated components fail to a safe position or are provided with an
during this routine test. The inspectors concluded that the routine testing performed  
  air accumulator. The emergency cooling system for the IA compressors was identified
verified the flow path to the unloaded compressor but did not verify that the cooling  
  to be in the scope of the MR because it is a non-safety related system that was used in
system was capable of supporting sustained operation of the compressor. The licensee  
  the emergency operating procedures (10 CFR 50.65(b)(2)).
documented this issue in CR 2009-22766 and planned to perform a formal test of the  
                                                                                      Enclosure
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)).  


                                              10
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
Enclosure
    cooling for the IA compressors under conditions comparable to those expected during a
This item will remain unresolved pending the completion of the stations testing, and  
    LOOP event. The item is identified as URI 05000335,389/2009006-02, Adequacy of
NRC review of the results of the IA emergency cooling systems capability to provide  
    Performance Monitoring of the IA Compressor Emergency Cooling System.
cooling for the IA compressors under conditions comparable to those expected during a  
.2.4 GD-1/2 Gravity Damper On HVS-5A/B Outlet
LOOP event. The item is identified as URI 05000335,389/2009006-02, Adequacy of  
  a. Inspection Scope
Performance Monitoring of the IA Compressor Emergency Cooling System.  
    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
.2.4  
    Gravity Damper. The team reviewed the system configuration and normal, abnormal,
GD-1/2 Gravity Damper On HVS-5A/B Outlet  
    and emergency operating procedures to verify correct implementation of design bases.
    Maintenance history, as demonstrated by system health reports, corrective maintenance
  a.  
    documentation, and CRs was reviewed to verify the design bases had been maintained;
Inspection Scope  
    potential degradation was being monitored; and that identified degradation or
    malfunctions had been adequately addressed. The team verified that the equipment
The team reviewed the DBD, related design basis documentation, drawings, TS, and the  
    periodic maintenance performed was consistent with vendor recommendations.
FSAR to identify design, maintenance, and operational requirements for the GD-1/2  
    Additionally, the team conducted a field walkdown of the GD-1/2 Gravity Damper with
Gravity Damper. The team reviewed the system configuration and normal, abnormal,  
    the licensee staff to assess observable material condition and to verify that the installed
and emergency operating procedures to verify correct implementation of design bases.
    configuration was consistent with the design basis and plant drawings.
Maintenance history, as demonstrated by system health reports, corrective maintenance  
  b. Findings
documentation, and CRs was reviewed to verify the design bases had been maintained;  
    No findings of significance were identified.
potential degradation was being monitored; and that identified degradation or  
.2.5 Pressurizer Relief Valve Isolation Valves, V1403 and V1405
malfunctions had been adequately addressed. The team verified that the equipment  
  a. Inspection Scope
periodic maintenance performed was consistent with vendor recommendations.
    The team reviewed the system DBD, related design basis support documentation,
Additionally, the team conducted a field walkdown of the GD-1/2 Gravity Damper with  
    drawings, TS, and the FSAR to identify design, maintenance, and operational
the licensee staff to assess observable material condition and to verify that the installed  
    requirements for these motor operated valves (MOVs). Maintenance history, as
configuration was consistent with the design basis and plant drawings.  
    demonstrated by system health reports, preventive and corrective maintenance, and
    CRs, was reviewed to verify that potential degradation was being monitored and
  b.  
    addressed. The MOV sizing calculations were reviewed to verify that the valves could
Findings  
    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
No findings of significance were identified.  
    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
.2.5  
    reviewed vendor recommendations for preventative maintenance and operation to verify
Pressurizer Relief Valve Isolation Valves, V1403 and V1405  
    that the maintenance practices were consistent with design basis requirements.
  b. Findings
  a.  
    No findings of significance were identified
Inspection Scope  
                                                                                      Enclosure
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  


                                              11
11  
.2.6 Battery Charger 1B
  a. Inspection Scope
Enclosure
    The team reviewed the Class 1E DC electrical distribution system DBD, related design
.2.6  
    basis support documents, drawings, appropriate sections of the TS, and the FSAR to
Battery Charger 1B  
    identify the design bases, maintenance requirements and the operational design
    requirements of the battery charger. The team reviewed the battery charger sizing
  a.  
    calculation, its conformance to the original design, and its capability to support current
Inspection Scope  
    load demands and battery charging requirements. The team also reviewed testing
    requirements and test procedures developed to demonstrate the design capabilities of
The team reviewed the Class 1E DC electrical distribution system DBD, related design  
    the charger under various plant conditions. The review included the vendor manual and
basis support documents, drawings, appropriate sections of the TS, and the FSAR to  
    the procedures that were developed to verify that the installation, operation, and
identify the design bases, maintenance requirements and the operational design  
    maintenance were in accordance with manufacturers recommendations.
requirements of the battery charger. The team reviewed the battery charger sizing  
    The team reviewed the health report and the results of recent tests to verify that the
calculation, its conformance to the original design, and its capability to support current  
    current performance was within accepted limits. Additionally, the team reviewed
load demands and battery charging requirements. The team also reviewed testing  
    selected corrective action reports to verify that anomalies were addressed and
requirements and test procedures developed to demonstrate the design capabilities of  
    corrected. A field walkdown was performed to assess the observable material condition
the charger under various plant conditions. The review included the vendor manual and  
    of the batteries, battery chargers, and inverters.
the procedures that were developed to verify that the installation, operation, and  
  b. Findings
maintenance were in accordance with manufacturers recommendations.  
    No findings of significance were identified.
.2.7 125V DC Bus 1B Power Panel & Cross-Tie Breakers to 125V DC Bus 1AB
The team reviewed the health report and the results of recent tests to verify that the  
  a. Inspection Scope
current performance was within accepted limits. Additionally, the team reviewed  
    The team reviewed the Class 1E DC electrical distribution system DBD, applicable
selected corrective action reports to verify that anomalies were addressed and  
    drawings and documents, including appropriate sections of the FSAR, to identify the
corrected. A field walkdown was performed to assess the observable material condition  
    design bases, maintenance and design requirements and to verify conformance of the
of the batteries, battery chargers, and inverters.  
    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
  b.  
    manufacturers recommendations. The team also addressed short circuit capabilities
Findings  
    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
No findings of significance were identified.  
    conditions. Additionally, the team reviewed recent system modifications and selected
    corrective action reports to verify that anomalies were addressed and corrected. The
.2.7  
    team reviewed operation requirements for the system and the interlocks provided to
125V DC Bus 1B Power Panel & Cross-Tie Breakers to 125V DC Bus 1AB  
    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
  a.  
    protection provided to ensure that the safety-related bus and battery were not
Inspection Scope  
    overloaded beyond calculated limits. A field walkdown of the power panels was
    performed to assess their installation, observable material conditions and to verify the
The team reviewed the Class 1E DC electrical distribution system DBD, applicable  
    current alignment of the buses.
drawings and documents, including appropriate sections of the FSAR, to identify the  
                                                                                        Enclosure
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.  


                                              12
12  
  b. Findings
    No findings of significance were identified.
Enclosure
.2.8 Engineered Safety Features Actuation System and Diverse Scram System
  b.  
  a. Inspection Scope
Findings  
    The team reviewed the engineered safety features actuation system (ESFAS) and
    diverse scram system (DSS) design basis document and applicable sections of the TS
No findings of significance were identified.  
    and FSAR to identify the design bases and the operational and maintenance
    requirements for the ESFAS and DSS. The team reviewed the DSS components
.2.8  
    including transmitters, logic modules, control and monitoring instrumentation, actuation
Engineered Safety Features Actuation System and Diverse Scram System  
    relays and contactors, selected components, and instrument loops associated with the
    ESFAS. The review included a detailed evaluation of instrument loop diagrams, control
  a.  
    logic, and wiring diagrams to confirm that the design conformed to the intended
Inspection Scope  
    operation of the systems. The review also addressed voltage requirements and voltage
    available at the various components, circuit protection, channel separation, and electrical
The team reviewed the engineered safety features actuation system (ESFAS) and  
    isolation. The team reviewed test procedures and evaluated the tests performed to
diverse scram system (DSS) design basis document and applicable sections of the TS  
    demonstrate the capability of the systems to perform the design basis functions. The
and FSAR to identify the design bases and the operational and maintenance  
    review included instrument and loop calibration procedures, test results, and adequacy
requirements for the ESFAS and DSS. The team reviewed the DSS components  
    of overlapping tests. The team confirmed that system and component maintenance was
including transmitters, logic modules, control and monitoring instrumentation, actuation  
    conducted per vendor recommendations. Additionally, a review of the latest system
relays and contactors, selected components, and instrument loops associated with the  
    health report and recent problem reports was conducted to evaluate whether component
ESFAS. The review included a detailed evaluation of instrument loop diagrams, control  
    concerns were adequately addressed and corrected and that their aging issues were
logic, and wiring diagrams to confirm that the design conformed to the intended  
    appropriately addressed. The team conducted a field verification of selected
operation of the systems. The review also addressed voltage requirements and voltage  
    components to evaluate installation criteria used and to assess their observable material
available at the various components, circuit protection, channel separation, and electrical  
    condition.
isolation. The team reviewed test procedures and evaluated the tests performed to  
  b. Findings
demonstrate the capability of the systems to perform the design basis functions. The  
    No findings of significance were identified.
review included instrument and loop calibration procedures, test results, and adequacy  
.2.9 Pressurizer Pressure Instrumentation
of overlapping tests. The team confirmed that system and component maintenance was  
  a. Inspection Scope
conducted per vendor recommendations. Additionally, a review of the latest system  
    The team reviewed applicable sections of the pressurizer system DBD and applicable
health report and recent problem reports was conducted to evaluate whether component  
    sections of the TS and FSAR to identify the design bases and the operational and
concerns were adequately addressed and corrected and that their aging issues were  
    maintenance requirements for the low range pressure control functions and components,
appropriately addressed. The team conducted a field verification of selected  
    including transmitters, logic modules, control and monitoring instrumentation, and
components to evaluate installation criteria used and to assess their observable material  
    actuation relays. The team conducted a detailed review of instrument loop diagrams
condition.  
    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
  b.  
    requirements and voltage available at the instrument components, circuit protection,
Findings  
    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
No findings of significance were identified.  
    component and loop calibration procedures, test results, and adequacy of overlapping
                                                                                    Enclosure
.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  


                                              13
13  
      tests. The team reviewed the latest system health report and recent corrective action
      reports to evaluate whether component concerns were adequately addressed and
Enclosure
      corrected and that aging issues were appropriately addressed. The team conducted a
tests. The team reviewed the latest system health report and recent corrective action  
      field walkdown of accessible instrument loop components to assess their observable
reports to evaluate whether component concerns were adequately addressed and  
      material condition.
corrected and that aging issues were appropriately addressed. The team conducted a  
  b.  Findings
field walkdown of accessible instrument loop components to assess their observable  
      No findings of significance were identified.
material condition.  
.2.10 Start-Up Transformers 1A and 1B and associated supply and feeder breakers
  a. Inspection Scope
  b.  
      The team reviewed the TS, DBD, FSAR, and alternate current (AC) load flow analysis,
Findings
      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
No findings of significance were identified.  
      and protective schemes in transformer relays to determine if adequate protection was
      provided. The team reviewed maintenance records, system health reports and
.2.10 Start-Up Transformers 1A and 1B and associated supply and feeder breakers  
      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
  a.  
      vulnerability to hazards.
Inspection Scope  
  b.  Findings
      No findings of significance were identified.
The team reviewed the TS, DBD, FSAR, and alternate current (AC) load flow analysis,  
.2.11 480VAC Load Center 1AB Cross-Tie Breaker (to either 480V 1A Load Center or 1B
as well as the Unit 1 computer modeling to assess whether station startup transformers  
      Load Center)
would have sufficient capacity to support required loads in accident/event conditions.
  a. Inspection Scope
The team further reviewed coordination studies to assess the effects of inrush currents  
      The team reviewed the TS, DBD, design drawings, calculations, vendor manuals and
and protective schemes in transformer relays to determine if adequate protection was  
      plant procedures to identify the design, maintenance and operational requirements for
provided. The team reviewed maintenance records, system health reports and  
      the cross-tie breaker. Electrical elementary drawings and wiring diagrams were
corrective action records to assess any adverse operating trends. A walk down of the  
      reviewed to verify that power sources would be available and adequate to power the
Start-Up Transformers 1A and 1B was performed to observe material condition and  
      appropriate safety loads during accident/event conditions. The team reviewed
vulnerability to hazards.  
      preventive maintenance and testing results to determine if the breakers were maintained
      in accordance with industry and vendor standards and recommendations. The team
  b.  
      reviewed short circuit and protection calculations to ensure that the breakers could
Findings
      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
No findings of significance were identified.  
      center 1A was performed to assess installation, configuration, observable material
      condition and vulnerability to hazards.
.2.11 480VAC Load Center 1AB Cross-Tie Breaker (to either 480V 1A Load Center or 1B  
                                                                                      Enclosure
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
14  
  b.  Findings
      No findings of significance were identified.
Enclosure
.2.12 480V Switchgear 1B2 (feeder and supply breakers and transformers)
  b.  
  a. Inspection Scope:
Findings
      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
No findings of significance were identified.  
      verify that power sources would be available and adequate to power the appropriate
      safety loads during accident/event conditions. The team reviewed preventive
.2.12 480V Switchgear 1B2 (feeder and supply breakers and transformers)  
      maintenance and testing procedures and results to determine if the breakers were
      maintained in accordance with industry and vendor standards and recommendations.
  a.  
      The team reviewed short circuit and protection calculations to ensure that the breakers
Inspection Scope:  
      could provide the appropriate interrupting and coordination protection. Selected
      corrective action reports were reviewed to determine if conditions adverse to quality
The team reviewed the TS, DBD, design drawings, calculations, vendor data and  
      were appropriately addressed and corrected. A walk down of the 480V 1B2 Breaker
manuals and plant procedures to identify the design, maintenance and operational  
      panel was performed to assess installation, configuration, observable material condition
requirements. Electrical elementary drawings and wiring diagrams were reviewed to  
      and vulnerability to hazards.
verify that power sources would be available and adequate to power the appropriate  
  b.  Findings
safety loads during accident/event conditions. The team reviewed preventive  
      No findings of significance were identified.
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
maintained in accordance with industry and vendor standards and recommendations.  
      Seal Cooler Heat Exchanger Outlet (TIS-14-32A1/B1/B2/A2)
The team reviewed short circuit and protection calculations to ensure that the breakers  
  a. Inspection Scope
could provide the appropriate interrupting and coordination protection. Selected  
      The team reviewed design and licensing basis documents, drawings and vendor
corrective action reports were reviewed to determine if conditions adverse to quality  
      manuals to identify the design requirements for the temperature indication switches.
were appropriately addressed and corrected. A walk down of the 480V 1B2 Breaker  
      The team reviewed set point calculations to verify that set points were established in
panel was performed to assess installation, configuration, observable material condition  
      accordance with vendor data, equipment capability and system design parameters.
and vulnerability to hazards.  
      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
  b.  
      reviewed calibration records and procedures to verify that instrument accuracy was
Findings
      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.
No findings of significance were identified.  
  b.  Findings
      No findings of significance were identified.
.2.13 Temperature Indication Switches for Reactor Coolant Pump (RCP) 1A and 1B CCW  
                                                                                      Enclosure
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.


                                              15
15  
.2.14 Intersystem Loss of Coolant Accident (LOCA) Instrumentation
  a. Inspection Scope
Enclosure
      The team reviewed design and licensing basis documents, drawings and vendor
.2.14 Intersystem Loss of Coolant Accident (LOCA) Instrumentation  
      manuals to identify the design requirements and capabilities of the intersystem LOCA
      instrumentation. The following instrumentation was included in the review: CCW Surge
  a.  
      Tank Level (LS-14-1A and B; LS-14-5, LG-14-2A and B); CCW System Radiation
Inspection Scope  
      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
The team reviewed design and licensing basis documents, drawings and vendor  
      were established in accordance with vendor data, equipment capability and system
manuals to identify the design requirements and capabilities of the intersystem LOCA  
      design parameters. Appropriate procedures were reviewed to verify set point data and
instrumentation. The following instrumentation was included in the review: CCW Surge  
      alarm points had been consistently translated. The team reviewed calibration records
Tank Level (LS-14-1A and B; LS-14-5, LG-14-2A and B); CCW System Radiation  
      and procedures to verify that instrument accuracy was monitored and maintained.
Monitors; Reactor/Auxiliary Building (RAB) Sump Level; and RAB Radiation Monitors.
      Maintenance history, as demonstrated by work orders and corrective action records, was
The team reviewed set point and level calculations to verify that set points and levels  
      reviewed to note any anomalies in equipment history and to verify corrective actions
were established in accordance with vendor data, equipment capability and system  
      were accomplished in a timely matter.
design parameters. Appropriate procedures were reviewed to verify set point data and  
  b.  Findings
alarm points had been consistently translated. The team reviewed calibration records  
      No findings of significance were identified.
and procedures to verify that instrument accuracy was monitored and maintained.
.2.15 Safety Injection Tank (SIT) Instrumentation
Maintenance history, as demonstrated by work orders and corrective action records, was  
  a. Inspection Scope
reviewed to note any anomalies in equipment history and to verify corrective actions  
      The team reviewed design and licensing basis documents, drawings and vendor
were accomplished in a timely matter.  
      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
  b.  
      levels were established in accordance with vendor data, equipment capability and
Findings
      system design parameters. Appropriate procedures were reviewed to verify alarm levels
   
      and set point data had been consistently translated. The team reviewed calibration
No findings of significance were identified.  
      records and procedures to verify that instrument accuracy was monitored and
      maintained. Maintenance history, as demonstrated by work orders and CRs, was
.2.15 Safety Injection Tank (SIT) Instrumentation  
      reviewed to note any anomalies in equipment history and to verify corrective actions
      were accomplished in a timely matter.
  a.  
  b.  Findings
Inspection Scope  
      No findings of significance were identified.
.2.16 Safety Injection (SI) System Check Valves (V3227, V07174, V07172, V3106, V3107)
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 DBD, related design basis documentation, drawings, TS, and the
instrumentation. The team reviewed set point calculations to verify that set points and  
      FSAR to identify design, maintenance, and operational requirements for selected SI
levels were established in accordance with vendor data, equipment capability and  
      system check valves. Maintenance history, as demonstrated by system health reports,
system design parameters. Appropriate procedures were reviewed to verify alarm levels  
      preventive and corrective maintenance, and CRs, was reviewed to verify that potential
and set point data had been consistently translated. The team reviewed calibration  
                                                                                      Enclosure
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  


                                                16
16  
      degradation was being monitored and addressed. The team conducted interviews with
      the SI System Engineer to obtain additional information and verify the stations
Enclosure
      implementation and analysis of industry operating experience related to check valves.
degradation was being monitored and addressed. The team conducted interviews with  
  b.  Findings
the SI System Engineer to obtain additional information and verify the stations  
      No findings of significance were identified.
implementation and analysis of industry operating experience related to check valves.  
.2.17 Volume Control Tank (VCT) MOVs 2501 & 2504
  a. Inspection Scope
  b.  
      The team reviewed the system DBD, related design basis support documentation,
Findings
      drawings, TS, and the FSAR to identify design, maintenance, and operational
   
      requirements for these MOVs. Maintenance history, as demonstrated by system health
No findings of significance were identified.  
      reports, preventive and corrective maintenance, and CRs, was reviewed to verify that
      potential degradation was being monitored and addressed. The MOV sizing calculations
.2.17 Volume Control Tank (VCT) MOVs 2501 & 2504  
      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
  a.  
      other significant characteristics into the sizing calculations. A review was conducted of
Inspection Scope  
      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
The team reviewed the system DBD, related design basis support documentation,  
      proper operation of the valve. The team reviewed vendor recommendations for
drawings, TS, and the FSAR to identify design, maintenance, and operational  
      preventative maintenance and operation to determine if maintenance practices were
requirements for these MOVs. Maintenance history, as demonstrated by system health  
      consistent with design basis requirements.
reports, preventive and corrective maintenance, and CRs, was reviewed to verify that  
  b.  Findings
potential degradation was being monitored and addressed. The MOV sizing calculations  
      No findings of significance were identified.
were reviewed to verify that the valves could operate during all credited design bases  
.2.18 CCW Control Valves (HCV-14-8A, HCV-14-8B, & HCV-14-9)
events and that the licensee appropriately translated the correct valve dimensions and  
  a. Inspection Scope
other significant characteristics into the sizing calculations. A review was conducted of  
      The team reviewed applicable portions of the FSAR, DBD, and drawings to identify
the licensees testing procedures and results from diagnostic valve testing to verify the  
      design basis requirements for these valves. The air operator sizing calculations were
MOVs were tested in a manner that would detect a malfunctioning valve and verify  
      reviewed to verify inputs were consistent with the most limiting design basis operating
proper operation of the valve. The team reviewed vendor recommendations for  
      conditions. Procurement documentation for the solenoids was reviewed to verify
preventative maintenance and operation to determine if maintenance practices were  
      compliance with environmental qualification (EQ) requirements. Stroke time surveillance
consistent with design basis requirements.  
      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
  b.  
      reviewed to identify recommendations for inspection and maintenance. The CR history
Findings
      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.  
      No findings of significance were identified.
                                                                                      Enclosure
.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.  


                                                17
17  
.2.19 SIT Outlet Valves (V3634, V3614, V3624, & V3644)
  a. Inspection Scope
Enclosure
      The team reviewed the system DBD, related design basis support documentation,
.2.19 SIT Outlet Valves (V3634, V3614, V3624, & V3644)  
      drawings, TS, and the FSAR to identify design, maintenance, and operational
      requirements for these MOVs. Maintenance history, as demonstrated by system health
  a.  
      reports, preventive and corrective maintenance, and CRs, was reviewed to verify that
Inspection Scope  
      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
The team reviewed the system DBD, related design basis support documentation,  
      MOVs were tested in a manner that would detect a malfunctioning valve and verify
drawings, TS, and the FSAR to identify design, maintenance, and operational  
      proper operation of the valve. The team reviewed maintenance practices and vendor
requirements for these MOVs. Maintenance history, as demonstrated by system health  
      recommendations for preventative maintenance and operation to verify that the valves
reports, preventive and corrective maintenance, and CRs, was reviewed to verify that  
      were being maintained consistent with design basis requirements.
potential degradation was being monitored and addressed. A review was conducted of  
  b.  Findings
the licensees testing procedures and results from diagnostic valve testing to verify the  
      No findings of significance were identified.
MOVs were tested in a manner that would detect a malfunctioning valve and verify  
.2.20 Motors and Electrical Components in Inspection Scope
proper operation of the valve. The team reviewed maintenance practices and vendor  
  a. Inspection Scope
recommendations for preventative maintenance and operation to verify that the valves  
      The team reviewed AC and direct current (DC) load flow and voltage (V) drop
were being maintained consistent with design basis requirements.  
      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
  b.  
      was also performed to determine if each component had sufficient voltage to perform its
Findings
      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
No findings of significance were identified.  
      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
.2.20 Motors and Electrical Components in Inspection Scope  
      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
  a.  
      coordination of protective devices to determine if the components were adequately
Inspection Scope  
      protected for overcurrent conditions and the protection was selected to ensure
      satisfactory operation during worst-case bus voltages. The team reviewed the AC and
The team reviewed AC and direct current (DC) load flow and voltage (V) drop  
      DC bus system health reports and recent corrective action reports to determine if circuit
calculations to determine if each motor within the inspection sample had adequate  
      breaker issues were being adequately resolved. Additionally, the team reviewed
terminal voltage to start and operate under worst case design basis events. This review  
      preventive maintenance and testing procedures to verify conformance to manufacturer
was also performed to determine if each component had sufficient voltage to perform its  
      recommendations. For MOVs, the team reviewed the electrical terminal voltages
design function. The review addressed power supply, cable amp capacity, and voltage  
      provided as design inputs to the mechanical torque and thrust calculations to verify the
drop during all modes of operation. For MOVs, the team evaluated valve motor starting  
      values were consistent with analyzed system conditions.
requirements to determine correct modeling in the voltage analysis. The team reviewed  
  b.  Findings
the electrical control schematics associated with the motors to evaluate if the control  
      Introduction: The inspectors identified a finding of very low safety significance (Green)
circuits had adequate voltage to start or stop the motor when required. The team also  
      involving a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the
reviewed the protection provided for each of the inspection sample components and the  
      licensees failure to maintain the safety-related 125V DC system design basis
coordination of protective devices to determine if the components were adequately  
      information consistent with the plant configuration. Specifically, a revision to the Unit 1,
protected for overcurrent conditions and the protection was selected to ensure  
                                                                                        Enclosure
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,  


                                          18
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
Enclosure
basis information. The licensees failure to maintain the vital 125V DC design basis
safety-related 125V DC system analysis (Calculation PSL-1FSE-05-002) incorporated  
information consistent with the plant configuration was identified as a performance
incorrect design input specifications related to the inverter, resulting in inaccurate design  
deficiency.
basis information. The licensees failure to maintain the vital 125V DC design basis  
Description: The current revision of DC Calculation PSL-1FSE-05-002 did not reflect the
information consistent with the plant configuration was identified as a performance  
current configuration of the Unit 1 DC system. In 2006, the licensee prepared two
deficiency.  
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
Description: The current revision of DC Calculation PSL-1FSE-05-002 did not reflect the  
had not occurred at the time of inspection. Based on licensee verbal information, the
current configuration of the Unit 1 DC system. In 2006, the licensee prepared two  
installation of the inverters was scheduled for 2012. The licensee issued Revision 1 of
station modification packages to replace the existing safety-related inverters with new  
the above calculation on December 10, 2008. This revision included the proposed
ones. The replacement of these components, however, did not occur as scheduled and  
replacement inverter equipment specifications as design inputs. The specifications for
had not occurred at the time of inspection. Based on licensee verbal information, the  
the replacement inverters were less limiting than the presently installed inverters. In
installation of the inverters was scheduled for 2012. The licensee issued Revision 1 of  
particular, the installed inverters require a minimum of 105V DC to operate and have an
the above calculation on December 10, 2008. This revision included the proposed  
efficiency of 75 percent. The replacement inverters require 100V DC and have an
replacement inverter equipment specifications as design inputs. The specifications for  
efficiency of 81 percent.
the replacement inverters were less limiting than the presently installed inverters. In  
Through discussions with the licensee pertaining to the discrepancy between the current
particular, the installed inverters require a minimum of 105V DC to operate and have an  
plant configuration and the 125V DC system design analysis, the inspection team
efficiency of 75 percent. The replacement inverters require 100V DC and have an  
determined that such discrepancies are permitted by the stations Quality Assurance
efficiency of 81 percent.  
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
Through discussions with the licensee pertaining to the discrepancy between the current  
modification. Since calculations are issued as-engineered, when a modification is
plant configuration and the 125V DC system design analysis, the inspection team  
cancelled it may be necessary to revise calculations to return them to the correct
determined that such discrepancies are permitted by the stations Quality Assurance  
configuration. Since the QA procedure did not establish a time limit when a discrepancy
Procedure ENG-QI 1.5. Specifically, Section 5.1.C of ENG-QI 1.5 states: Calculations  
was allowed to exist between the design documentation and the configuration of the
may be created or revised to support modifications and issued before completion of the  
plant, such discrepancy could exist for years, as in the case of the postponed
modification. Since calculations are issued as-engineered, when a modification is  
replacement of the inverters. The team was concerned that the existence of official
cancelled it may be necessary to revise calculations to return them to the correct  
design documents that are inconsistent with the configuration of the plant might
configuration. Since the QA procedure did not establish a time limit when a discrepancy  
invalidate conclusions pertaining to the operability and performance of structures,
was allowed to exist between the design documentation and the configuration of the  
systems, and components, particularly if, during the intervening period, other design
plant, such discrepancy could exist for years, as in the case of the postponed  
changes and plant modifications were developed on the assumption that the documents
replacement of the inverters. The team was concerned that the existence of official  
of record reflect the current plant configuration. Regarding the incorrect inverter
design documents that are inconsistent with the configuration of the plant might  
minimum voltage information, the team was concerned that degradation of the battery in
invalidate conclusions pertaining to the operability and performance of structures,  
subsequent years combined with the implementation of other potential modifications
systems, and components, particularly if, during the intervening period, other design  
could result in the nuclear safety-related inverters being unable to perform their design
changes and plant modifications were developed on the assumption that the documents  
safety function.
of record reflect the current plant configuration. Regarding the incorrect inverter  
Analysis: The licensees failure to maintain the vital 125V DC design basis information
minimum voltage information, the team was concerned that degradation of the battery in  
consistent with the plant configuration was identified as a performance deficiency. This
subsequent years combined with the implementation of other potential modifications  
finding, associated with the Mitigating Systems Cornerstone attribute of Design Control
could result in the nuclear safety-related inverters being unable to perform their design  
was more than minor because if left uncorrected, it had the potential to lead to a more
safety function.  
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
Analysis: The licensees failure to maintain the vital 125V DC design basis information  
assuring the adequacy of voltage to the safety-related inverters. The team assessed
consistent with the plant configuration was identified as a performance deficiency. This  
this finding for significance in accordance with NRC Manual Chapter 0609, using the
finding, associated with the Mitigating Systems Cornerstone attribute of Design Control  
Phase I SDP worksheet for mitigating systems and determined that the finding was of
was more than minor because if left uncorrected, it had the potential to lead to a more  
                                                                                    Enclosure
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  


                                                19
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
Enclosure
    configuration discrepancies permitted by procedure ENG-QI 1.5, the team did not
very low safety significance (Green) since it was a design deficiency determined not to  
    identify any other design document that was inconsistent with the current plant
have resulted in a loss of safety function. Regarding the programmatic concern about  
    configuration. This finding reflects current station performance because the identified
configuration discrepancies permitted by procedure ENG-QI 1.5, the team did not  
    performance deficiency occurred in a calculation revision dated December 10, 2008.
identify any other design document that was inconsistent with the current plant  
    The issue was identified to be programmatic because the station procedure for
configuration. This finding reflects current station performance because the identified  
    controlling engineering calculations (ENG-QI-1.5) contributed to the performance
performance deficiency occurred in a calculation revision dated December 10, 2008.
    deficiency. This finding has a cross-cutting aspect in the area of human performance
The issue was identified to be programmatic because the station procedure for  
    because the licensee failed to ensure that procedures (i.e. ENG-QI 1.5) were available
controlling engineering calculations (ENG-QI-1.5) contributed to the performance  
    and adequate to assure nuclear safety (specifically, complete, accurate and up-to-date
deficiency. This finding has a cross-cutting aspect in the area of human performance  
    design documentation). [H.2(c)]
because the licensee failed to ensure that procedures (i.e. ENG-QI 1.5) were available  
    Enforcement: 10 CFR 50 Appendix B, Criterion III, Design Control, requires that design
and adequate to assure nuclear safety (specifically, complete, accurate and up-to-date  
    changes, including field changes, shall be subject to design control measures
design documentation). [H.2(c)]  
    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
Enforcement: 10 CFR 50 Appendix B, Criterion III, Design Control, requires that design  
    to the original design in that a revision to the Unit 1, safety-related 125V DC system
changes, including field changes, shall be subject to design control measures  
    analysis (Calculation PSL-1FSE-05-002) incorporated incorrect design input
commensurate with those applied to the original design. Contrary to the above, design  
    specifications related to the system inverter equipment. As a result, the stations Unit 1,
changes were not subject to design control measures commensurate with those applied  
    safety-related 125V DC system analysis, revised on December 10, 2008, did not reflect
to the original design in that a revision to the Unit 1, safety-related 125V DC system  
    the actual plant configuration and was not conservative in that it concluded that a
analysis (Calculation PSL-1FSE-05-002) incorporated incorrect design input  
    minimum voltage of 100V DC was adequate to assure operation of the safety-related
specifications related to the system inverter equipment. As a result, the stations Unit 1,  
    inverters. Because the finding was of very low safety significance and was entered into
safety-related 125V DC system analysis, revised on December 10, 2008, did not reflect  
    the licensees corrective action program (CR 2009-24517), this violation is being treated
the actual plant configuration and was not conservative in that it concluded that a  
    as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement
minimum voltage of 100V DC was adequate to assure operation of the safety-related  
    Policy: NCV 05000335,389/2009006-03, Failure to Maintain the Safety-Related 125V
inverters. Because the finding was of very low safety significance and was entered into  
    DC System Design Basis Information Consistent with the Plant Configuration.
the licensees corrective action program (CR 2009-24517), this violation is being treated  
.3   Review of Low Margin Operator Actions
as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement  
  a. Inspection Scope
Policy: NCV 05000335,389/2009006-03, Failure to Maintain the Safety-Related 125V  
    The team performed a margin assessment and detailed review of six risk significant and
DC System Design Basis Information Consistent with the Plant Configuration.  
    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
.3  
    actions, the team performed a walkthrough of associated Emergency Operating
Review of Low Margin Operator Actions  
    procedures (EOPs) abnormal operating procedures (AOPs), Normal Operating
    Procedures (OPs), and other operations procedures with appropriate plant operators
  a.  
    and engineers to assess operator knowledge level, adequacy of procedures, availability
Inspection Scope  
    of special equipment when required, and the conditions under which the procedures
    would be performed. The inspection team conducted detailed reviews with operations
The team performed a margin assessment and detailed review of six risk significant and  
    and training department leadership, and observed operator training on the plant
time critical operator actions. Where possible, margins were determined by the review  
    simulator, to assess the procedural rationale and approach to meeting the design basis
of the assumed design basis and FSAR response times. For the selected operator  
    and FSAR response and performance requirements. Operator actions were observed
actions, the team performed a walkthrough of associated Emergency Operating  
    on the plant simulator and during plant walk downs. Additionally, the team reviewed the
procedures (EOPs) abnormal operating procedures (AOPs), Normal Operating  
    station configuration control for risk significant manual valves. This review included field
Procedures (OPs), and other operations procedures with appropriate plant operators  
    verification that the valve positions for a selected sample of risk significant manual
and engineers to assess operator knowledge level, adequacy of procedures, availability  
    valves was consistent with applicable drawings and system operating procedures.
of special equipment when required, and the conditions under which the procedures  
                                                                                        Enclosure
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.  


                                            20
20  
  Operator actions associated with the following events/evolutions were reviewed:
  *   Reactor coolant system feed and bleed and Power Operated Relief Valve (PORV)
Enclosure
        fails open (block valve use)
Operator actions associated with the following events/evolutions were reviewed:  
  *   Inner-system Loss of Coolant Accident (LOCA)
  *   Anticipated Transient Without a Scram (ATWS) - Emergency Boration
*  
  *   Cross-tie 480V 1AB load center
Reactor coolant system feed and bleed and Power Operated Relief Valve (PORV)  
  *   Condensate storage tank makeup from the treated water storage tank
fails open (block valve use)  
  *   Restoration of non-essential CCW following Safety Injection Actuation Signal (SIAS)
*  
b. Findings
Inner-system Loss of Coolant Accident (LOCA)
  Introduction: The team identified a URI related to the licensees failure to provide
*  
  adequate procedures for restoration of non-essential CCW following a SIAS.
Anticipated Transient Without a Scram (ATWS) - Emergency Boration  
  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
Cross-tie 480V 1AB load center  
  not address the potential adverse impact on essential cooling flow required to mitigate a
*  
  LOCA when the non-essential CCW was restored.
Condensate storage tank makeup from the treated water storage tank  
  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
Restoration of non-essential CCW following Safety Injection Actuation Signal (SIAS)  
  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
  b.  
  cooling flow was provided to essential loads for design basis accident conditions.
Findings  
  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
Introduction: The team identified a URI related to the licensees failure to provide  
  on assumptions in the LOCA Containment Analysis, JPN-PSL-SENP-93-001, Rev. 0.
adequate procedures for restoration of non-essential CCW following a SIAS.
  When establishing the essential cooling flow balance per this procedure, the non-
Specifically, emergency operating procedure, 1-EOP-99, Appendix A, Sampling Steam  
  essential portion of the CCW system was isolated. Therefore, adequate essential
Generators, and Appendix J, Restoration of CCW and CBO to the RCPs, Rev. 38, did  
  cooling flow was assured only when the non-essential portion of the system was
not address the potential adverse impact on essential cooling flow required to mitigate a  
  isolated. The EOP assured that CCW train separation was maintained when the non-
LOCA when the non-essential CCW was restored.  
  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
Description: Emergency Operating Procedure 1-EOP-99, Appendix A and J, step 2,  
  essential components, primarily the containment coolers used in containment pressure
directed the operator to restore non-essential CCW if the related isolation valve closed  
  control, the shutdown heat exchanger used for decay heat removal, and cooling for
due to the SIAS. Additionally, an input to isolate non-essential CCW was provided by a  
  emergency core cooling system (ECCS) pumps. The team concluded that the
low CCW surge tank level signal. The purpose of both signals was to assure adequate  
  procedure action to restore non-essential CCW flow after an SIAS signal adversely
cooling flow was provided to essential loads for design basis accident conditions.  
  impacted the licensees capability to assure adequate cooling of essential components
  following a LOCA induced SIAS. In particular, this concern applied to the circumstance
The station CCW flow balance procedure (1-NOP-14.02, Rev. 20, Appendix I) positioned  
  of only one train of CCW being available during LOCA, assuming a single failure event
system flow balance valves to establish cooling flow to the essential components based  
  resulted in the loss of the redundant train.
on assumptions in the LOCA Containment Analysis, JPN-PSL-SENP-93-001, Rev. 0.
  Following identification by the team, the licensee initiated CR 2009-22623 to assess this
When establishing the essential cooling flow balance per this procedure, the non-
  issue. The immediate compensatory action was to issue a standing order to the
essential portion of the CCW system was isolated. Therefore, adequate essential  
  operators related to Emergency Operating Procedure 1, 2-EOP-99 directing them to not
cooling flow was assured only when the non-essential portion of the system was  
  restore the non-essential CCW when responding to a SIAS when only one CCW train
isolated. The EOP assured that CCW train separation was maintained when the non-
  was available. Additionally, the licensee initiated an evaluation to assess the impact on
essential header was restored but did not address that the essential cooling load flow  
  essential CCW flow if non-essential CCW was restored to allow cooling of the RCPs and
would be diverted with the potential adverse impact on cooling capability for the  
                                                                                    Enclosure
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  


                                                21
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
Enclosure
    performance deficiency. The licensees evaluation, and the NRC review of this
the steam generator sample coolers. The licensees failure to provide adequate  
    evaluation, is needed to determine if adequate cooling would be available to essential
procedures for restoration of non-essential CCW following a SIAS was identified as a  
    equipment following the LOCA induced SIAS when the non-essential CCW was
performance deficiency. The licensees evaluation, and the NRC review of this  
    restored. This issue is being documented as URI 05000335, 389/2009006-04,
evaluation, is needed to determine if adequate cooling would be available to essential  
    Inadequate Procedure for Restoration of Non-Essential CCW Flow Following a SIAS.
equipment following the LOCA induced SIAS when the non-essential CCW was  
.4   Review of Industry Operating Experience
restored. This issue is being documented as URI 05000335, 389/2009006-04,  
  a. Inspection Scope
Inadequate Procedure for Restoration of Non-Essential CCW Flow Following a SIAS.  
    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
.4  
    performed an independent applicability review for issues that were identified as
Review of Industry Operating Experience  
    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:
  a.  
    *   Generic Letter 07-01, Inaccessible or Underground Power Cable Failures that
Inspection Scope  
        Disable Accident Mitigation Systems or Cause Plant Transients.
    *   Generic Letter 98-02, Loss of Reactor Coolant Inventory and Associated Potential for
The team reviewed selected operating experience issues that had occurred at domestic  
        Loss of Emergency Mitigation Functions While in a Shutdown Condition.
and foreign nuclear facilities for applicability at the St. Lucie Nuclear Plant. The team  
    *   NRC Information Notice 07-09, Equipment Operability Under Degraded Voltage
performed an independent applicability review for issues that were identified as  
        Conditions.
applicable to the St. Lucie Nuclear Plant and were selected for a detailed review. The  
    *   Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient,
issues that received a detailed review by the team included:  
        dated July 25, 1984
    *   NRC Information Notice 2008-02: Findings Identified During Component Design
*  
        Bases Inspections, March 19, 2008
Generic Letter 07-01, Inaccessible or Underground Power Cable Failures that  
  b. Findings
Disable Accident Mitigation Systems or Cause Plant Transients.
    No findings of significance were identified.
*  
.5   Review of Permanent Plant Modifications
Generic Letter 98-02, Loss of Reactor Coolant Inventory and Associated Potential for  
  a. Inspection Scope
Loss of Emergency Mitigation Functions While in a Shutdown Condition.
    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
NRC Information Notice 07-09, Equipment Operability Under Degraded Voltage  
    capability of the components have not been degraded through modifications. The
Conditions.  
    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
Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient,  
    Experiments and Permanent Plant Modifications. The following modification was
dated July 25, 1984  
    reviewed:
*  
    *   PC/M: 04028, Medium Voltage Switchgear Circuit Breaker Replacement - Phase III
NRC Information Notice 2008-02: Findings Identified During Component Design  
                                                                                        Enclosure
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  


                                                22
22  
  b.  Findings
   
    No findings of significance were identified.
Enclosure
4OA5 Other Activities
  b.  
    CCW Air Intrusion Event
Findings
  a. Inspection Scope
   
    The team performed a detailed review of the condition reports related to the air intrusion
No findings of significance were identified.  
    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.
4OA5 Other Activities  
  b.  Findings
    Introduction: The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design
CCW Air Intrusion Event  
    Control, for the licensees failure to translate the design basis, as specified in the license
   
    application, into specifications, drawings, procedures, and instructions. Specifically, a
  a.  
    non-safety system failure (i.e. containment IA compressors) could cause a common
Inspection Scope  
    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
The team performed a detailed review of the condition reports related to the air intrusion  
    CCW system non-essential header provided cooling and seal makeup to these IA
into the CCW system event that took place from 2:13 a.m. on October 16, 2008, through  
    compressors. On October 16, 2008, an air intrusion event occurred in which air from the
4:02 a.m. on October 17, 2008.  
    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
  b.  
    system check valves V1818A and V18060 to the IA receiver tank combined with the
Findings
    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.
Introduction: The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design  
    The inspectors review of the CCW system CRs identified that the air intrusion event
Control, for the licensees failure to translate the design basis, as specified in the license  
    occurred from 2:13 a.m. on October 16, 2008 through 4:02 a.m. on October 17, 2008.
application, into specifications, drawings, procedures, and instructions. Specifically, a  
    The teams review identified that this event resulted in the degraded performance of both
non-safety system failure (i.e. containment IA compressors) could cause a common  
    trains of the Unit 1 CCW system and a potential loss of the CCW safety function.
cause failure of both trains of a safety system (i.e. CCW system).  
    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
Description: The Unit 1 design included IA compressors inside containment. The Unit 1  
    the air intrusion. Subsequent to this, operators vented a significant amount of air from
CCW system non-essential header provided cooling and seal makeup to these IA  
    the CCW system in order to return the system parameters to normal. The air intrusion
compressors. On October 16, 2008, an air intrusion event occurred in which air from the  
    event demonstrated an original design deficiency on Unit 1 such that a non-safety
IA compressors located inside containment entered into the CCW system. The licensee  
    system (IA) could adversely impact the reliability, capability, and availability of the safety-
determined the air intrusion into the CCW system was caused by the failures of IA  
    related CCW system. In this case, the design deficiency was a common cause failure
system check valves V1818A and V18060 to the IA receiver tank combined with the  
    mechanism.
failure of the IA unloading solenoid SE1814A. Additionally, leakage through the IA seal  
    In addition to the air intrusion source discussed above, the team also determined that
water cooler, which interfaces with the CCW system, created pathways for air to enter  
    this vulnerability potentially existed on the waste gas compressors since non-essential
the CCW system.  
    CCW flow was also used for waste gas compressor aftercooler cooling. The waste gas
                                                                                          Enclosure
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  


                                          23
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
Enclosure
system from a failure in the waste compressor units was applicable to both Unit 1 and
compressors run at approximately 160 psig system pressure and the CCW system  
Unit 2.
pressure is approximately 120 psig. The common cause failure vulnerability of the CCW  
The CCW system essential header cools the containment fan coolers (CFCs), shutdown
system from a failure in the waste compressor units was applicable to both Unit 1 and  
cooling heat exchanger, and bearing/seal coolers for the containment spray, high
Unit 2.  
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
The CCW system essential header cools the containment fan coolers (CFCs), shutdown  
intrusion affecting both CCW trains could have prevented the CCW system from
cooling heat exchanger, and bearing/seal coolers for the containment spray, high  
delivering the flow specified by the TS Surveillance Requirement 4.6.2.1.1 (1,200 gpm to
pressure safety injection, and low pressure safety injection pumps. The CCW trains are  
each cooling train fan unit), and reduced flow to the remaining safety-related heat
normally cross-connected during normal operation. The team concluded that the air  
exchangers below the analyzed/required values. An additional impact of the air intrusion
intrusion affecting both CCW trains could have prevented the CCW system from  
into the CCW system was potential degradation of the safety-related heat exchangers
delivering the flow specified by the TS Surveillance Requirement 4.6.2.1.1 (1,200 gpm to  
performance. The team concluded that given enough air introduction, the possibility
each cooling train fan unit), and reduced flow to the remaining safety-related heat  
existed that the heat exchangers could become fully or partially air bound (e.g., upper
exchangers below the analyzed/required values. An additional impact of the air intrusion  
tube regions), thus significantly decreasing the heat transfer capability.
into the CCW system was potential degradation of the safety-related heat exchangers  
The combined effects of the reduced flow and the reduced heat transfer could lead to
performance. The team concluded that given enough air introduction, the possibility  
the inability of the CCW system to perform the following safety-related functions:
existed that the heat exchangers could become fully or partially air bound (e.g., upper  
*   Providing adequate cooling for those safety-related components associated with
tube regions), thus significantly decreasing the heat transfer capability.  
    containment and reactor decay heat removal during accident conditions.
*   Providing adequate cooling for those safety-related components associated with
The combined effects of the reduced flow and the reduced heat transfer could lead to  
    achieving safe shutdown.
the inability of the CCW system to perform the following safety-related functions:  
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
Providing adequate cooling for those safety-related components associated with  
cooling system from performing its safety function. The licensees evaluation of the air
containment and reactor decay heat removal during accident conditions.  
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
Providing adequate cooling for those safety-related components associated with  
effect of the air intrusion on the heat exchangers performance. The licensee initiated
achieving safe shutdown.  
CR 2009-22929 with actions to evaluate the past operability of the CCW system during
the air intrusion event.
This event simultaneously affected both redundant trains of the CCW system (i.e.  
Analysis: An original plant design deficiency was revealed by the CCW air intrusion
introduced a common cause failure mechanism). FSAR section 9.2.2.3.2, Single Failure  
event of October 16, 2008. This design deficiency involved the potential for a non-safety
Analysis, states in part: there is no single failure that could prevent the component  
system (IA or waste gas) adversely impacting the reliability, capability, and availability of
cooling system from performing its safety function. The licensees evaluation of the air  
the safety-related CCW system. This design deficiency was identified as a performance
intrusion event failed to evaluate the operability consequences of the air intrusion on the  
deficiency. In this case, the design deficiency introduced a common cause failure
CCW flow reduction to the safety-related heat exchangers and failed to consider the  
mechanism. FSAR section 9.2.2.3.2, Single Failure Analysis, states, in part: there is no
effect of the air intrusion on the heat exchangers performance. The licensee initiated  
single failure that could prevent the CCW system from performing its safety function.
CR 2009-22929 with actions to evaluate the past operability of the CCW system during  
This single failure vulnerability existed on Units 1 and 2 from potential failure of the
the air intrusion event.  
aftercoolers on the waste gas compressors and on Unit 1 from the potential failure of the
containment IA system.
Analysis: An original plant design deficiency was revealed by the CCW air intrusion  
                                                                                    Enclosure
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.


                                            24
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
Enclosure
cornerstone objective because, if left uncorrected, it would affect the availability,
The finding was determined to be more than minor because it was associated with the  
reliability and capability of a safety system to perform its intended safety function.
Mitigating Systems Cornerstone attribute of Equipment Performance. It impacted the  
Specifically, with this vulnerability, a failure of the waste gas aftercooler (either unit) or a
cornerstone objective because, if left uncorrected, it would affect the availability,  
failure of the containment IA compressors (Unit 1 only) could cause air intrusion into the
reliability and capability of a safety system to perform its intended safety function.
CCW system and potentially lead to a loss of CCW event. A loss of CCW could result in
Specifically, with this vulnerability, a failure of the waste gas aftercooler (either unit) or a  
inadequate cooling to essential equipment used to mitigate design bases accidents. The
failure of the containment IA compressors (Unit 1 only) could cause air intrusion into the  
finding was assessed for significance in accordance with NRC Manual Chapter 0609,
CCW system and potentially lead to a loss of CCW event. A loss of CCW could result in  
using the Phase I and Phase II SDP worksheets for mitigating systems.
inadequate cooling to essential equipment used to mitigate design bases accidents. The  
It was determined that a Phase III analysis was required since this finding represented a
finding was assessed for significance in accordance with NRC Manual Chapter 0609,  
potential loss of safety system function for multiple trains which was not addressed by
using the Phase I and Phase II SDP worksheets for mitigating systems.  
the Phase II pre-solved tables/worksheets.
The preliminary Phase III analysis determined that for the air intrusion event of October
It was determined that a Phase III analysis was required since this finding represented a  
2008, it was reasonable to assume the initiating event frequency increased from the
potential loss of safety system function for multiple trains which was not addressed by  
baseline by at least one magnitude and therefore the performance deficiency was
the Phase II pre-solved tables/worksheets.  
preliminarily characterized as greater than Green. The preliminary Phase III analysis is
attached.
The preliminary Phase III analysis determined that for the air intrusion event of October  
The team concluded that this finding did not have an associated cross-cutting aspect
2008, it was reasonable to assume the initiating event frequency increased from the  
because the design of the CCW system was established in an original plant design, and
baseline by at least one magnitude and therefore the performance deficiency was  
therefore, was not representative of current licensee performance.
preliminarily characterized as greater than Green. The preliminary Phase III analysis is  
Enforcement: 10 CFR 50, Appendix B, Criterion III, Design Control, requires that the
attached.    
design basis specified in the license application be correctly translated into
specifications, drawings, procedures, and instructions. FSAR section 9.2.2.3.2, Single
The team concluded that this finding did not have an associated cross-cutting aspect  
Failure Analysis, states in part: there is no single failure that could prevent the
because the design of the CCW system was established in an original plant design, and  
component cooling system from performing its safety function. Contrary to the above,
therefore, was not representative of current licensee performance.  
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
Enforcement: 10 CFR 50, Appendix B, Criterion III, Design Control, requires that the  
compressor aftercoolers, both units, or containment IA compressors, Unit 1 only) could
design basis specified in the license application be correctly translated into  
result in a common cause failure of both trains of a safety system (i.e. CCW system).
specifications, drawings, procedures, and instructions. FSAR section 9.2.2.3.2, Single  
The air intrusion event revealed an original design deficiency that a non-safety system
Failure Analysis, states in part: there is no single failure that could prevent the  
(IA) could adversely impact the reliability, capability, and availability of safety related
component cooling system from performing its safety function. Contrary to the above,  
CCW system. In this case, the design deficiency was a common cause failure
the licensee failed to correctly translate the original design basis into specifications for  
mechanism. This design deficiency was established in the original plant design and has
the design of the CCW system. Specifically, a non-safety system failure (i.e. waste gas  
existed since the operating licenses were issued. This issue is being documented as AV
compressor aftercoolers, both units, or containment IA compressors, Unit 1 only) could  
05000335, 389/2009006-05, Failure to Translate Design Basis Specifications to Prevent
result in a common cause failure of both trains of a safety system (i.e. CCW system).
Single Failure of CCW.
The air intrusion event revealed an original design deficiency that a non-safety system  
Introduction: The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI,
(IA) could adversely impact the reliability, capability, and availability of safety related  
Corrective Action, for the licensees failure to identify a condition adverse to quality
CCW system. In this case, the design deficiency was a common cause failure  
associated with the CCW air intrusion event that occurred in October 2008. Following
mechanism. This design deficiency was established in the original plant design and has  
the October 2008 event, the licensee failed to properly identify and correct the source of
existed since the operating licenses were issued. This issue is being documented as AV  
the air intrusion into the CCW system prior to closing the associated Condition Report.
05000335, 389/2009006-05, Failure to Translate Design Basis Specifications to Prevent  
                                                                                      Enclosure
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.  


                                          25
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
Enclosure
a performance deficiency.
The licensees failure to identify the source (i.e. leak path from the containment IA  
Description: The team reviewed CRs for the CCW system air intrusion event that took
compressors to the CCW system) of air intrusion into the CCW system was identified as  
place from October 16, 2008 through October 17, 2008. Review of the control room
a performance deficiency.  
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.
Description: The team reviewed CRs for the CCW system air intrusion event that took  
Subsequent to this, operators vented a significant amount of air from the CCW pumps
place from October 16, 2008 through October 17, 2008. Review of the control room  
and heat exchangers in order to return the system parameters to normal. As discussed
operational logs, CR 2008-31947, CR 2008-34697 and, CR2008-35753 identified that  
in section 4OA5 b.1, the licensee identified that the containment IA compressors
both CCW pumps exhibited motor amp fluctuations due to the air in the system.
provided a pathway for which air intrusion into the CCW system could occur.
Subsequent to this, operators vented a significant amount of air from the CCW pumps  
The teams review of the station data identified that the indicated maximum containment
and heat exchangers in order to return the system parameters to normal. As discussed  
IA pressure was approximately 113 psig during normal operation of the compressor.
in section 4OA5 b.1, the licensee identified that the containment IA compressors  
The maximum identified pressure during the air intrusion event was 129 psig (CCW
provided a pathway for which air intrusion into the CCW system could occur.  
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
The teams review of the station data identified that the indicated maximum containment  
solenoid or the solenoid itself, such that it remained closed keeping the unit loaded and
IA pressure was approximately 113 psig during normal operation of the compressor.
allowing header pressure to reach 129 psig.
The maximum identified pressure during the air intrusion event was 129 psig (CCW  
The licensee determined that the most likely path for air intrusion into the CCW system
system pressure is approximately 120 psig). The licensee identified that the elevated IA  
to be through the 1A containment IA compressors aftercooler (as documented in CR
pressure was attributed to a failure of the pressure switch that activates the unloader  
2008-34697). Listed below is a summary of actions taken by the licensee:
solenoid or the solenoid itself, such that it remained closed keeping the unit loaded and  
*   Initial troubleshooting performed on November 10, 2008, under CR 2008-31947,
allowing header pressure to reach 129 psig.  
    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 licensee determined that the most likely path for air intrusion into the CCW system  
*   CR 2008-34697 identified that CCW to the IA compressor aftercoolers was not
to be through the 1A containment IA compressors aftercooler (as documented in CR  
    needed and should remain isolated. TSA-1-08-013 was developed to accomplish
2008-34697). Listed below is a summary of actions taken by the licensee:  
    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
Initial troubleshooting performed on November 10, 2008, under CR 2008-31947,  
    gas, also did not leak.
determined that IA aftercoolers, when tested to 100 psig with compressed air, did not  
*   CR 2008-35753 was closed on November 19, 2008. The closure was based on
leak. CR 2008-31947 was subsequently closed to CR 2008-34697.  
    isolation of the CCW from the aftercoolers to remove the risk of compressed air
*  
    entering the CCW System from this high pressure source.
CR 2008-34697 identified that CCW to the IA compressor aftercoolers was not  
*   The licensee performed an operability review of the CCW system and determined
needed and should remain isolated. TSA-1-08-013 was developed to accomplish  
    the system was operable (CR 2008-31947). The corrective action documents did not
this task and CR 2008-34697 was closed to CR 2008-35753.  
    provide a basis for this determination.
*  
*   The 1A compressor unloading solenoid valve body and internals were replaced on
Subsequent troubleshooting was performed on November 18, 2008, under WO  
    November 21, 2008 (after the event). The licensees decision-making at the time of
38025447 and determined that IA aftercoolers, when tested to 120 psig with argon  
    the event resulted in the isolation of CCW cooling to both aftercoolers.
gas, also did not leak.  
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
CR 2008-35753 was closed on November 19, 2008. The closure was based on  
intrusion path. CR 2009-24030 was initiated to evaluate why a prompt operability
isolation of the CCW from the aftercoolers to remove the risk of compressed air  
determination was not requested by the licensees operations department at the time of
entering the CCW System from this high pressure source.  
the event. The licensee had not performed an engineering evaluation to support the
*  
                                                                                  Enclosure
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  


                                            26
26  
operability determination. Consequently, the licensee had not evaluated if the air
intrusion was significant enough to block cooling flow to safety-related components
Enclosure
during an accident. CR 2009-22929 was initiated to perform a past operability review to
operability determination. Consequently, the licensee had not evaluated if the air  
address this concern.
intrusion was significant enough to block cooling flow to safety-related components  
The team identified to the licensee an additional air intrusion path, not previously
during an accident. CR 2009-22929 was initiated to perform a past operability review to  
identified by the licensee. The team concluded that the most likely source for the air
address this concern.  
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
The team identified to the licensee an additional air intrusion path, not previously  
event. The potential source of air intrusion into the CCW system from the containment
identified by the licensee. The team concluded that the most likely source for the air  
IA system was re-reviewed and re-evaluated by the licensee.
intrusion was the CCW seal makeup interface with the IA compressor. The licensee  
The licensee documented, in CR 2009-25209, that the most probable cause of the air
issued CR 2009-25209 to address the ineffective corrective actions for the air intrusion  
intrusion into the CCW system was the failure of 1A IA compressor unloader solenoid
event. The potential source of air intrusion into the CCW system from the containment  
(SE-18-14A) in conjunction with failure of check valves V1818A and V18060 to fully seat,
IA system was re-reviewed and re-evaluated by the licensee.  
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
The licensee documented, in CR 2009-25209, that the most probable cause of the air  
cooler for containment IA compressor did not identify any leaks. The original evaluation
intrusion into the CCW system was the failure of 1A IA compressor unloader solenoid  
documented in CR 2008-31947 failed to identify or address this susceptibility. As
(SE-18-14A) in conjunction with failure of check valves V1818A and V18060 to fully seat,  
detailed above, the teams review of the troubleshooting and corrective actions
which could have allowed instrument air to enter the CCW system via the make-up line.
documented in CR 2008-31947, CR 2008-34697, CR 2008-35753, and Work Order
This failure mechanism explained why leak testing of the aftercoolers and seal water  
(WO) 38025447 determined that the licensee did not correctly identify the source of the
cooler for containment IA compressor did not identify any leaks. The original evaluation  
air intrusion. This vulnerability also potentially exists on both units should the
documented in CR 2008-31947 failed to identify or address this susceptibility. As  
aftercoolers on the waste gas compressors fail. The waste gas compressors run at
detailed above, the teams review of the troubleshooting and corrective actions  
approximately 160 psig pressure and the CCW system pressure is approximately 120
documented in CR 2008-31947, CR 2008-34697, CR 2008-35753, and Work Order  
psig. The team concluded that the failure of a non-safety system (i.e. containment IA or
(WO) 38025447 determined that the licensee did not correctly identify the source of the  
waste gas compressor) that could cause a common cause failure of both trains of a
air intrusion. This vulnerability also potentially exists on both units should the  
safety-related system (i.e. CCW system) was a condition adverse to quality. The
aftercoolers on the waste gas compressors fail. The waste gas compressors run at  
licensee initiated CR 2009-23882 to address this concern.
approximately 160 psig pressure and the CCW system pressure is approximately 120  
Analysis: The licensees failure to identify and correct the source (i.e. leak path from the
psig. The team concluded that the failure of a non-safety system (i.e. containment IA or  
containment IA compressors to the CCW system) of air intrusion into the CCW system
waste gas compressor) that could cause a common cause failure of both trains of a  
was identified as a performance deficiency. The finding was determined to be more than
safety-related system (i.e. CCW system) was a condition adverse to quality.   The  
minor because it was associated with the Mitigating Systems Cornerstone attribute of
licensee initiated CR 2009-23882 to address this concern.  
Equipment Performance. It impacted the cornerstone objective because it affected the
availability, reliability and capability of a safety system to perform its intended safety
Analysis: The licensees failure to identify and correct the source (i.e. leak path from the  
function. Specifically, the failure to identify and correct the source of air intrusion into the
containment IA compressors to the CCW system) of air intrusion into the CCW system  
CCW system affected the ability of the system to ensure that adequate cooling would be
was identified as a performance deficiency. The finding was determined to be more than  
available or maintained to essential equipment used to mitigate design bases accidents.
minor because it was associated with the Mitigating Systems Cornerstone attribute of  
The finding was assessed for significance in accordance with NRC Manual Chapter
Equipment Performance. It impacted the cornerstone objective because it affected the  
0609, using the Phase I and Phase II SDP worksheets for mitigating systems. It also
availability, reliability and capability of a safety system to perform its intended safety  
was determined that a Phase III analysis was required since this finding represented a
function. Specifically, the failure to identify and correct the source of air intrusion into the  
potential loss of safety system function for multiple trains which was not addressed by
CCW system affected the ability of the system to ensure that adequate cooling would be  
the Phase II pre-solved tables/worksheets.
available or maintained to essential equipment used to mitigate design bases accidents.
                                                                                      Enclosure
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.  


                                                27
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
Enclosure
    baseline by at least one magnitude and therefore the performance deficiency was
The preliminary Phase III analysis determined that for the air intrusion event of October  
    preliminarily characterized as greater than Green. The preliminary Phase III analysis is
2008, it was reasonable to assume the initiating event frequency increased from the  
    attached.
baseline by at least one magnitude and therefore the performance deficiency was  
    This finding was determined to have a cross-cutting aspect in the area of Human
preliminarily characterized as greater than Green. The preliminary Phase III analysis is  
    Performance, Decision Making, specifically, H.1(a), which states, the licensee makes
attached.  
    safety-significant or risk-significant decisions using a systematic process, especially
    when faced with uncertain or unexpected plant conditions, to ensure safety is
This finding was determined to have a cross-cutting aspect in the area of Human  
    maintained. The inspectors determined that the licensees decision to close the
Performance, Decision Making, specifically, H.1(a), which states, the licensee makes  
    associated corrective action documents without finding the cause of the air intrusion
safety-significant or risk-significant decisions using a systematic process, especially  
    contributed to extending the length of time that the CCW system was susceptible to this
when faced with uncertain or unexpected plant conditions, to ensure safety is  
    common cause failure mode.
maintained. The inspectors determined that the licensees decision to close the  
    Enforcement: 10 CFR 50 Appendix B Criterion XVI, Corrective Action, requires, in part,
associated corrective action documents without finding the cause of the air intrusion  
    that measures shall be established to assure that conditions adverse to quality, such as
contributed to extending the length of time that the CCW system was susceptible to this  
    failures, malfunctions, deficiencies, deviations, defective material and equipment, and
common cause failure mode.  
    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
Enforcement: 10 CFR 50 Appendix B Criterion XVI, Corrective Action, requires, in part,  
    failed to identify and correct the source of the air intrusion into the CCW system and
that measures shall be established to assure that conditions adverse to quality, such as  
    closed the associated Condition Report. As a result, the plant remained susceptible to a
failures, malfunctions, deficiencies, deviations, defective material and equipment, and  
    non-safety system failure (i.e. containment IA compressors), which could cause a
nonconformances are promptly identified and corrected. Contrary to the above,  
    common cause failure of both trains of a safety system (i.e. CCW System), for
following the discovery of air in the CCW system on October 16, 2008, the licensee  
    approximately one year. This issue is being documented as AV 05000335,
failed to identify and correct the source of the air intrusion into the CCW system and  
    389/2009006-06, Failure to Identify and Correct a Condition Adverse to Quality such that
closed the associated Condition Report. As a result, the plant remained susceptible to a  
    a Non-Safety Related System Could Cause a Common Mode Failure of Both Trains of a
non-safety system failure (i.e. containment IA compressors), which could cause a  
    Safety-Related System.
common cause failure of both trains of a safety system (i.e. CCW System), for  
4OA6 Meetings, Including Exit
approximately one year. This issue is being documented as AV 05000335,  
    On September 4, 2009, the team presented the preliminary inspection results to Mr.
389/2009006-06, Failure to Identify and Correct a Condition Adverse to Quality such that  
    Johnston and other members of the licensees staff. Although proprietary information
a Non-Safety Related System Could Cause a Common Mode Failure of Both Trains of a  
    was reviewed as part of this inspection, all proprietary information was returned and no
Safety-Related System.  
    proprietary information is documented in the report.
    On October 19, 2009, the NRC presented preliminary inspection results in a telephone
4OA6 Meetings, Including Exit  
    with Mr. Jim Porter and other members of the licensees staff.
    On December 3, 2009, the NRC presented preliminary inspection results in a telephone
On September 4, 2009, the team presented the preliminary inspection results to Mr.  
    with Mr. Eric Katzman and other members of the licensees staff.
Johnston and other members of the licensees staff. Although proprietary information  
    On December 10, 2009, the NRC presented inspection results in a telephone exit with
was reviewed as part of this inspection, all proprietary information was returned and no  
    Mr. Eric Katzman and other members of the licensees staff.
proprietary information is documented in the report.  
ATTACHMENT: SUPPPLEMENTAL INFORMATION
                                                                                      Enclosure
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  


                              SUPPLEMENTAL INFORMATION
                                  KEY POINTS OF CONTACT
Attachment
Licensee personnel:
SUPPLEMENTAL INFORMATION  
P. Barnes, Mechanical Engineering Design Supervisor
D. Cecchett, Licensing
KEY POINTS OF CONTACT  
G. Johnston, Site Vice President
E. Katzman, Licensing Manager
Licensee personnel:  
D. Lany, Operations Senior Reactor Operator
J. Porter, Manager Design Engineering
P. Barnes, Mechanical Engineering Design Supervisor  
S. Short, Electrical Engineering Design Supervisor
D. Cecchett, Licensing  
NRC personnel
G. Johnston, Site Vice President  
D. Jones, Acting Chief, Engineering Branch Chief 1, Division of Reactor Safety, RII
E. Katzman, Licensing Manager  
T. Hoeg, Senior Resident Inspector, St. Lucie
D. Lany, Operations Senior Reactor Operator
W. Rogers, Senior Risk Analyst, RII
J. Porter, Manager Design Engineering  
S. Sanchez, Resident Inspector, St. Lucie
S. Short, Electrical Engineering Design Supervisor  
                      LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened and Closed
NRC personnel  
05000335,389/2009006-01           NCV         Failure to Meet the ASME Boiler and Pressure
D. Jones, Acting Chief, Engineering Branch Chief 1, Division of Reactor Safety, RII  
                                                Vessel Code, Section VIII, Division 1
T. Hoeg, Senior Resident Inspector, St. Lucie  
                                                Requirements for the Overpressure Protection
W. Rogers, Senior Risk Analyst, RII  
                                                for the CCW Surge Tank (1R21.2.2)
S. Sanchez, Resident Inspector, St. Lucie  
05000335,389/2009006-03           NCV         Failure to Maintain the Safety-Related 125V
                                                DC System Design Basis Information
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED  
                                                Consistent with the Plant Configuration
                                                (1R21.2.20)
Opened and Closed
Opened
05000335,389/2009006-02           URI         Adequacy of Performance Monitoring of the IA
05000335,389/2009006-01  
                                                Compressor Emergency Cooling System.
NCV  
                                                (1R21.2.3)
Failure to Meet the ASME Boiler and Pressure  
05000335, 389/2009006-04           URI         Inadequate Procedure for Restoration of Non-
Vessel Code, Section VIII, Division 1  
                                                Essential CCW Flow Following a SIAS
Requirements for the Overpressure Protection  
                                                (1R21.3)
for the CCW Surge Tank (1R21.2.2)  
05000335, 389/2009006-05           AV           Failure to Translate Design Basis
05000335,389/2009006-03  
                                                Specifications to Prevent Single Failure of
NCV  
                                                CCW (4OA5)
Failure to Maintain the Safety-Related 125V  
05000335,389/2009006-06           AV           Failure to Identify and Correct a Condition
DC System Design Basis Information  
                                                Adverse to Quality such that Non-Safety
Consistent with the Plant Configuration  
                                                Related System Could Cause a Common
(1R21.2.20)  
                                                Mode Failure of Both Trains of a Safety-
Opened  
                                                Related System (4OA5)
                                                                                    Attachment
05000335,389/2009006-02  
URI  
Adequacy of Performance Monitoring of the IA  
Compressor Emergency Cooling System.  
(1R21.2.3)  
05000335, 389/2009006-04  
URI  
Inadequate Procedure for Restoration of Non-
Essential CCW Flow Following a SIAS  
(1R21.3)  
05000335, 389/2009006-05  
AV  
Failure to Translate Design Basis  
Specifications to Prevent Single Failure of  
CCW (4OA5)  
05000335,389/2009006-06  
AV  
Failure to Identify and Correct a Condition  
Adverse to Quality such that Non-Safety  
Related System Could Cause a Common  
Mode Failure of Both Trains of a Safety-
Related System (4OA5)  


                              LIST OF DOCUMENTS REVIEWED
Calculations
Attachment
128-42A.6002, Component Cooling Water (CCW) System SIAS Operation, Rev. 0,
LIST OF DOCUMENTS REVIEWED  
    CRN 07127-17201
Calculations  
PSL-1FJM-93-06, Intake Cooling Water System Performance, Rev. 2
128-42A.6002, Component Cooling Water (CCW) System SIAS Operation, Rev. 0,  
007-AS93-C-004 PSL-1CHN-93-002A, Unit 1 LOCA Containment Pressure/Temperature (P/T)
CRN 07127-17201  
    Analysis for 102% Power (2754 MWt), Rev. 0
PSL-1FJM-93-06, Intake Cooling Water System Performance, Rev. 2  
NSSS-040, Component Cooling Water System, Rev. 3
007-AS93-C-004 PSL-1CHN-93-002A, Unit 1 LOCA Containment Pressure/Temperature (P/T)  
PSL-1FJI-91-006, FIS-14-12A, B, C, & D Setpoints, Rev. 1
Analysis for 102% Power (2754 MWt), Rev. 0  
PSL-BSFM-01-014, Acceptable Corrosion Allowance on the Units 1 and 2 CCW Surge Tank for
NSSS-040, Component Cooling Water System, Rev. 3  
    a 50 psi Design Pressure, Rev. 0
PSL-1FJI-91-006, FIS-14-12A, B, C, & D Setpoints, Rev. 1  
PSL-BSFM-01-019, Component Cooling Water Surge Tank Pressure Analysis, Rev. 0
PSL-BSFM-01-014, Acceptable Corrosion Allowance on the Units 1 and 2 CCW Surge Tank for  
32-82-6001, HVAC System HVS 5A &5B, RAB El. 43- 0, Heat Load Verification, Rev 0
a 50 psi Design Pressure, Rev. 0  
C2-B-9, HVAC System HVS 5A &5B, RAB El. 43- 0, Heat Load Verification, Rev 0
PSL-BSFM-01-019, Component Cooling Water Surge Tank Pressure Analysis, Rev. 0  
JPN-PSL-SEIP-92-025, Evaluation of CEs PPS Setpoint Calculation, Rev 4
32-82-6001, HVAC System HVS 5A &5B, RAB El. 43- 0, Heat Load Verification, Rev 0  
JPN-PSL-SEMP-91-007, Safety Evaluation for RAB Electrical Equipment and Battery Room,
C2-B-9, HVAC System HVS 5A &5B, RAB El. 43- 0, Heat Load Verification, Rev 0  
    Rev 1
JPN-PSL-SEIP-92-025, Evaluation of CEs PPS Setpoint Calculation, Rev 4  
PSL-1FJE-90-013, St. Lucie Unit 1 Emergency Diesel Generator 1A and 1B Electrical Loads,
JPN-PSL-SEMP-91-007, Safety Evaluation for RAB Electrical Equipment and Battery Room,  
    Rev 6
Rev 1  
PSL-1FJE-90-0014, Unit 1 Battery Charger 1A, 1AA, 1B, 1BB, and 1AB Sizing, Rev 1
PSL-1FJE-90-013, St. Lucie Unit 1 Emergency Diesel Generator 1A and 1B Electrical Loads,  
PSL-1FJE-90-026, St. Lucie - Unit 1 Short Circuit, Voltage Drop and PSB-1 Analysis, Rev 6
Rev 6  
PSL-1-FJE-91-002, Instrument Inverters 1A, 1B, 1C, & 1D AC Output Loading, Rev 05
PSL-1FJE-90-0014, Unit 1 Battery Charger 1A, 1AA, 1B, 1BB, and 1AB Sizing, Rev 1  
PSL-1FSE-03-009, Unit 1 ELECTRICAL System Computer Model (ETAP) Documentation,
PSL-1FJE-90-026, St. Lucie - Unit 1 Short Circuit, Voltage Drop and PSB-1 Analysis, Rev 6  
    Rev 1
PSL-1-FJE-91-002, Instrument Inverters 1A, 1B, 1C, & 1D AC Output Loading, Rev 05  
PSL-1FSE-05-002, Unit 1 125V DC System ETAP Model & Analysis, Rev 1
PSL-1FSE-03-009, Unit 1 ELECTRICAL System Computer Model (ETAP) Documentation,  
PSL-1FSE-05-002, Unit 1 -125V DC System ETAP Model & Analysis, Rev. 1
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
PSL-1FSE-05-002, Unit 1 125V DC System ETAP Model & Analysis, Rev 1  
IC.0004, Safety Injection Tank Level Instrumentation, Rev 4
PSL-1FSE-05-002, Unit 1 -125V DC System ETAP Model & Analysis, Rev. 1  
PSL-1-FJE-98-001, Review of Selective Coordination for the Electrical Circuits on the St. Lucie
PSL-1FJI-92-035, Unit 1 Pressurizer NR Pressure Uncertainty Determination, Rev 1  
    Unit 1 Essential Equipment List, rev 5, 9/26/02
PSL-1FJI-92-047, St. Lucie Unit 1 Safety Injection Tank Pressure Setpoints, 2/7/94  
PSL-1FSE-03-009, Unit 1 Electrical System Computer Model (ETAP) Documentation, Rev 1
IC.0004, Safety Injection Tank Level Instrumentation, Rev 4  
PSL-1-FJE-90-0026, Unit 1 Short Circuit, Voltage Drop and PSB-1 Analysis, Rev 6
PSL-1-FJE-98-001, Review of Selective Coordination for the Electrical Circuits on the St. Lucie  
Specifications
Unit 1 Essential Equipment List, rev 5, 9/26/02  
FLO-8770-764, Unit 1 CCW Surge Tank, original issue 10/31/71
PSL-1FSE-03-009, Unit 1 Electrical System Computer Model (ETAP) Documentation, Rev 1  
Procedures
PSL-1-FJE-90-0026, Unit 1 Short Circuit, Voltage Drop and PSB-1 Analysis, Rev 6  
1-NOP-14.02, Component Cooling Water System Operation, Rev. 25
2-NOP-14.02, Component Cooling Water System Operation, Rev. 15
Specifications  
1-NOP-50.01A, 125V DC Bus 1A (Class 1E) Normal Operation, Rev 2
FLO-8770-764, Unit 1 CCW Surge Tank, original issue 10/31/71  
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
Procedures  
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
2-NOP-14.02, Component Cooling Water System Operation, Rev. 15  
1-1010030, Loss Of Instrument Air, Rev. 33a
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
1-EOP-99, Appendices / Figures / Tables / Data Sheets, Rev. 38
                                                                                    Attachment


                                              3
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
Attachment
1-OSP-14.01C, Component Cooling Water Pump Code Run, Rev. 0B
1-OSP-14.01A, Component Cooling Water Pump Code Run, Rev. 0B  
1-OSP-100.01, Schedule of Periodic Tests, Checks and Calibrations Week 1, Rev. 34B
1-OSP-14.01B, Component Cooling Water Pump Code Run, Rev. 0B  
0-EMP-50.01, 125V DC System Battery Charger 18 Month Operability Testing, Rev 8D
1-OSP-14.01C, Component Cooling Water Pump Code Run, Rev. 0B  
0-EMP-50.01A, 125V DC Bus 1A (Class 1E) Normal Operation, Rev 2
1-OSP-100.01, Schedule of Periodic Tests, Checks and Calibrations Week 1, Rev. 34B  
0-EMP-50.05, Safety Battery Performance Test, Rev 4A
0-EMP-50.01, 125V DC System Battery Charger 18 Month Operability Testing, Rev 8D  
0-EMP-50.05, Safety Battery Performance Test, Rev 6
0-EMP-50.01A, 125V DC Bus 1A (Class 1E) Normal Operation, Rev 2  
0-EMP-50.08, Safety Battery Emergency Load Profile Test (Service Test), Rev 11
0-EMP-50.05, Safety Battery Performance Test, Rev 4A  
0-EMP-80.11, Votes Testing of Globe and Gate Valves, Rev 6
0-EMP-50.05, Safety Battery Performance Test, Rev 6  
0-EMP-80.06, Preventative Maintenance of Limitorque MOV Actuators, Rev 17B
0-EMP-50.08, Safety Battery Emergency Load Profile Test (Service Test), Rev 11
0-CME-50.21, Safety Related Battery Cell Charging and Replacement, Rev 1A
0-EMP-80.11, Votes Testing of Globe and Gate Valves, Rev 6  
0-PMI-69-01, Anticipated Transient Without a Scram (ATWS) Functional Test, Rev 2
0-EMP-80.06, Preventative Maintenance of Limitorque MOV Actuators, Rev 17B  
1-EMP-50.01, Safety Battery 18 Month Maintenance, Rev 4E
0-CME-50.21, Safety Related Battery Cell Charging and Replacement, Rev 1A
1-IMP-01.37L, Pressurizer Pressure Low Range Loop Calibration, Rev 4D
0-PMI-69-01, Anticipated Transient Without a Scram (ATWS) Functional Test, Rev 2  
1-IMP-01.37T, Pressurizer Pressure Low Range Transmitter Calibration, Rev 2B
1-EMP-50.01, Safety Battery 18 Month Maintenance, Rev 4E  
1-IMP-01.39T, Pressurizer Pressure Safety Channel Transmitter Calibration, Rev 4
1-IMP-01.37L, Pressurizer Pressure Low Range Loop Calibration, Rev 4D  
1-IMP-26.14, Containment Atmosphere Process Monitor Functional and Calibration Instruction,
1-IMP-01.37T, Pressurizer Pressure Low Range Transmitter Calibration, Rev 2B  
    Rev 12A
1-IMP-01.39T, Pressurizer Pressure Safety Channel Transmitter Calibration, Rev 4  
1-IMP-26.19, Component Cooling Water Process Monitor Functional and Secondary Calibration
1-IMP-26.14, Containment Atmosphere Process Monitor Functional and Calibration Instruction,  
    Instruction, Rev 8
Rev 12A  
1-IMP-69.01, Safeguards Group Actuation Procedure, Rev 0B
1-IMP-26.19, Component Cooling Water Process Monitor Functional and Secondary Calibration  
1-IMP-69.02, ESFAS Monthly Channel Functional Test, Rev 11
Instruction, Rev 8  
1-OSP-50.01, 125V DC Bus 1AB Crosstie Breaker In-place Undervoltage Testing, Rev 1B
1-IMP-69.01, Safeguards Group Actuation Procedure, Rev 0B  
1-1400052, Engineered Safeguards Actuation System, Channel Functional Test, Rev 54
1-IMP-69.02, ESFAS Monthly Channel Functional Test, Rev 11  
1-IMP-14.02, CCW to RCP Seal Temperature Switch Calibration, Rev 3
1-OSP-50.01, 125V DC Bus 1AB Crosstie Breaker In-place Undervoltage Testing, Rev 1B  
OP-1-0010125, Schedule of Periodic Tests, Checks and Calibrations St. Lucie Unit 1, Rev 79
1-1400052, Engineered Safeguards Actuation System, Channel Functional Test, Rev 54  
OP-1-0010125A, Surveillance Data Sheets, Rev. 125
1-IMP-14.02, CCW to RCP Seal Temperature Switch Calibration, Rev 3  
1-1400064L, Installed Plant Instrumentation Calibration (Level), Rev 47
OP-1-0010125, Schedule of Periodic Tests, Checks and Calibrations St. Lucie Unit 1, Rev 79  
1-PTP-21, Bus 1A1 SF6 Breakers Pre-Operational Testing, Rev 0A
OP-1-0010125A, Surveillance Data Sheets, Rev. 125  
1-PTP-24, Bus 1B1 SF6 Breakers Pre-Operational Testing, Rev 0A
1-1400064L, Installed Plant Instrumentation Calibration (Level), Rev 47  
0310080, Preoperational Test Procedure, Component Cooling Water Functional Test, Rev. 2
1-PTP-21, Bus 1A1 SF6 Breakers Pre-Operational Testing, Rev 0A  
ENG-QI 1.5, Quality Instruction Nuclear Engineering Calculations, Rev 8
1-PTP-24, Bus 1B1 SF6 Breakers Pre-Operational Testing, Rev 0A  
IMP-76.01, Rosemount Transmitter Repair & Calibration (Model 1153 & 1154)
0310080, Preoperational Test Procedure, Component Cooling Water Functional Test, Rev. 2  
IMG-.04, Magnetrol Level Switch Calibration, Rev 10A
ENG-QI 1.5, Quality Instruction Nuclear Engineering Calculations, Rev 8  
Completed Procedures
IMP-76.01, Rosemount Transmitter Repair & Calibration (Model 1153 & 1154)  
1-OSP-14.01A, Component Cooling Water Pump Code Run, performed on: 6/12/09, 3/12/09,
IMG-.04, Magnetrol Level Switch Calibration, Rev 10A  
    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,
Completed Procedures  
    12/26/08, 9/26/08, 6/26/08, 3/27/08
1-OSP-14.01A, Component Cooling Water Pump Code Run, performed on: 6/12/09, 3/12/09,  
1-OSP-14.01C, Component Cooling Water Pump Code Run, performed on: 6/26/09, 3/27/09,
12/11/08, 9/12/08, 7/7/08, 3/15/08  
    12/26/08, 9/26/08, 6/26/08, 3/27/08
1-OSP-14.01B, Component Cooling Water Pump Code Run, performed on: 6/26/09, 3/27/09,  
1-NOP-14.02, Component Cooling Water System Operation, Appendix I, Essential CCW Load
12/26/08, 9/26/08, 6/26/08, 3/27/08  
    Flow Balance, performed on: 11/18/08
1-OSP-14.01C, Component Cooling Water Pump Code Run, performed on: 6/26/09, 3/27/09,  
2-NOP-14.02, Component Cooling Water System Operation, Appendix I, Essential CCW Load
12/26/08, 9/26/08, 6/26/08, 3/27/08  
    Flow Balance, performed on: 05/29/09
1-NOP-14.02, Component Cooling Water System Operation, Appendix I, Essential CCW Load  
                                                                                  Attachment
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  


                                          4
Drawings
4  
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
Attachment
8770-G-082, Sheet 2, Flow Diagram, Circulating & Intake Cooling Water System, Rev. 23
Drawings  
8770-G-083, Sheet 1A, Flow Diagram, Component Cooling System, Rev. 59
8770-G-078, Sheet 162A, Flow Diagram, Waste Management System, Rev. 14
8770-G-083, Sheet 1B, Flow Diagram, Component Cooling System, Rev. 57
8770-G-082, Sheet 1, Flow Diagram, Circulating & Intake Cooling Water System, Rev. 50  
8770-G-083, Sheet 2, Flow Diagram, Component Cooling System, Rev. 4
8770-G-082, Sheet 2, Flow Diagram, Circulating & Intake Cooling Water System, Rev. 23  
8770-G-085, Sheet 2A, Instrument Air System, Rev. 39
8770-G-083, Sheet 1A, Flow Diagram, Component Cooling System, Rev. 59  
8770-G-085, Sheet 4B, Instrument Air System, Rev. 31
8770-G-083, Sheet 1B, Flow Diagram, Component Cooling System, Rev. 57  
8770-G-089, Sheet 1A, Flow Diagram, Turbine Cooling Water System, Rev. 26
8770-G-083, Sheet 2, Flow Diagram, Component Cooling System, Rev. 4  
8770-G-089, Sheet 1B, Flow Diagram, Turbine Cooling Water System, Rev. 26
8770-G-085, Sheet 2A, Instrument Air System, Rev. 39  
8770-G-089, Sheet 2, Flow Diagram, Turbine Cooling Water System, Rev. 25
8770-G-085, Sheet 4B, Instrument Air System, Rev. 31  
8770-G-100, Flow Diagram Symbols, Rev. 10
8770-G-089, Sheet 1A, Flow Diagram, Turbine Cooling Water System, Rev. 26  
8770-G-125, Sheet CC-H-5, Large Bore Piping Isometric, Component Cooling Piping, Rev. 5
8770-G-089, Sheet 1B, Flow Diagram, Turbine Cooling Water System, Rev. 26  
8770-G-125, Sheet CC-H-7, Large Bore Piping Isometric, Component Cooling Piping, Rev. 5
8770-G-089, Sheet 2, Flow Diagram, Turbine Cooling Water System, Rev. 25  
8770-G-862, HVAC - Air Flow Diagram, Rev. 31
8770-G-100, Flow Diagram Symbols, Rev. 10  
8770-G-879, HVAC - Control Diagrams - Sheet 2, Rev. 39
8770-G-125, Sheet CC-H-5, Large Bore Piping Isometric, Component Cooling Piping, Rev. 5  
8770-16336, Bettis Actuator, Spring Return, Rev. 1
8770-G-125, Sheet CC-H-7, Large Bore Piping Isometric, Component Cooling Piping, Rev. 5  
8770-5624, Component Cooling Water Surge Tank, Rev. 4
8770-G-862, HVAC - Air Flow Diagram, Rev. 31  
8770-B-326, Sh. 269, Schematic Diagram Safety Injection Tank Isolation Valve V-3626, Rev 8
8770-G-879, HVAC - Control Diagrams - Sheet 2, Rev. 39  
8770-B-327, Sh. 118, Control Wiring Diagram Pressurizer Relief Isolation Valve V-1403, Rev 16
8770-16336, Bettis Actuator, Spring Return, Rev. 1  
8770-B-327, Sh. 120, Control Wiring Diagram Pressurizer Relief Isolation Valve V-1405, Rev 15
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-
8770-B-327, Sh. 140, Control Wiring Diagram Measurement Channels L-1103, L-1116, & P-
    1103, Rev 15
1103, Rev 15  
8770-B-327, Sh. 141, Control Wiring Diagram Measurement Channels PS-1118, PT-1116, &
8770-B-327, Sh. 141, Control Wiring Diagram Measurement Channels PS-1118, PT-1116, &  
    PT-1104, Rev 24
PT-1104, Rev 24  
8770-B-327, Sh. 161, Control Wiring Diagram Volume Control Tank Discharge Valve V-2501,
8770-B-327, Sh. 161, Control Wiring Diagram Volume Control Tank Discharge Valve V-2501,  
    Rev 8
Rev 8  
8770-B-327, Sh. 162, Control Wiring Diagram Refueling Water to Discharge Pumps V-2504,
8770-B-327, Sh. 162, Control Wiring Diagram Refueling Water to Discharge Pumps V-2504,  
    Rev 7
Rev 7  
8770-B-327, Sh. 201, Control Wiring Diagram Component Cooling Water Pump 1A, Rev 16
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. 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. 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 &
8770-B-327, Sh. 211, Control Wiring Diagram Component Cool Wtr Shutdown Heat Exch &  
    Surge Tank Fill Valves, Rev 13
Surge Tank Fill Valves, Rev 13  
8770-B-327, Sh. 250, Control Wiring Diagram Shutdown Cooling Isolation Valve V-3481,
8770-B-327, Sh. 250, Control Wiring Diagram Shutdown Cooling Isolation Valve V-3481,    
    Rev 12
Rev 12  
8770-B-327, Sh. 253, Control Wiring Diagram Shutdown Cooling Isolation Valve V-3651,
8770-B-327, Sh. 253, Control Wiring Diagram Shutdown Cooling Isolation Valve V-3651,    
    Rev 15
Rev 15  
8770-B-327, Sh. 269, Control Wiring Diagram Injection Tank 1A1 Isolation Valve V-3624, Rev 8
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,
8770-B-327, Sh. 270, Control Wiring Diagram Injection Tank 1A2 Isolation Valve V-3614,    
    Rev 13
Rev 13  
8770-B-327, Sh. 271, Control Wiring Diagram Injection Tank 1B1 Isolation Valve V-3634,
8770-B-327, Sh. 271, Control Wiring Diagram Injection Tank 1B1 Isolation Valve V-3634,    
    Rev 13
Rev 13  
8770-B-327, Sh. 272, Control Wiring Diagram Injection Tank 1B2 Isolation Valve V-3644, Rev 8
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. 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
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,
5  
    Rev 7
8770-B-327, Sh. 477, Control Wiring Diagram Electrical Equipment Room Supply Fan HVS-5B,
Attachment
    Rev 7
8770-B-327, Sh. 476, Control Wiring Diagram Electrical Equipment Room Supply Fan HVS-5A,  
8770-B-327, Sh. 532, Control Wiring Diagram Safeguards Room A Sump Pumps, Rev 9
Rev 7  
8770-B-327, Sh. 533, Control Wiring Diagram Safeguards Room B Sump Pumps, Rev 10
8770-B-327, Sh. 477, Control Wiring Diagram Electrical Equipment Room Supply Fan HVS-5B,  
8770-B-327, Sh. 583, Control Wiring Diagram Equipment Drain Sump Pump, Rev 6
Rev 7  
8770-B-327, Sh. 600, Control Wiring Diagram Instrument Air Compressor Emergency Cooling
8770-B-327, Sh. 532, Control Wiring Diagram Safeguards Room A Sump Pumps, Rev 9  
    System, Rev 2
8770-B-327, Sh. 533, Control Wiring Diagram Safeguards Room B Sump Pumps, Rev 10  
8770-B-327, Sh. 934, Control Wiring Diagram 4160V Swgr 1A2 Fdr to Bus 1A3, Rev 13
8770-B-327, Sh. 583, Control Wiring Diagram Equipment Drain Sump Pump, Rev 6  
8770-B-327, Sh. 935, Control Wiring Diagram 4160V Swgr 1B2 Fdr to Bus 1B3, Rev 13
8770-B-327, Sh. 600, Control Wiring Diagram Instrument Air Compressor Emergency Cooling  
8770-B-327, Sh. 978, Control Wiring Diagram 480V Switchgear 1A2 - 1AB Tie, Rev 9
System, Rev 2  
8770-B-327, Sh. 979, Control Wiring Diagram 480V Switchgear 1AB - 1A2 Tie, Rev 10
8770-B-327, Sh. 934, Control Wiring Diagram 4160V Swgr 1A2 Fdr to Bus 1A3, Rev 13  
8770-B-327, Sh. 980, Control Wiring Diagram 480V Switchgear 1B2 Fdr, Rev 11
8770-B-327, Sh. 935, Control Wiring Diagram 4160V Swgr 1B2 Fdr to Bus 1B3, Rev 13  
8770-B-327, Sh. 1002, Control Wiring Diagram Battery 1B & Battery Charger 1B, Rev 24
8770-B-327, Sh. 978, Control Wiring Diagram 480V Switchgear 1A2 - 1AB Tie, Rev 9  
8770-B-327, Sh. 1003, Control Wiring Diagram Battery Charger 1AB, Rev 16
8770-B-327, Sh. 979, Control Wiring Diagram 480V Switchgear 1AB - 1A2 Tie, Rev 10  
8770-B-327, Sh. 1601, Control Wiring Diagram Battery Charger 1BB, Rev 6
8770-B-327, Sh. 980, Control Wiring Diagram 480V Switchgear 1B2 Fdr, Rev 11  
8770-G-272, Main One Line Wiring Diagram, Rev 25
8770-B-327, Sh. 1002, Control Wiring Diagram Battery 1B & Battery Charger 1B, Rev 24  
8770-G-274, Auxiliary One Line Diagram, Rev 16
8770-B-327, Sh. 1003, Control Wiring Diagram Battery Charger 1AB, Rev 16  
8770-G-275, 6.9KV Swgr. & 4.16 KV Swgr. One Line Wiring Diagram, Rev 17
8770-B-327, Sh. 1601, Control Wiring Diagram Battery Charger 1BB, Rev 6  
8770-G-275, Sh.2, 480V Swgr. & Pressurizer Htr. Bus One Line Wiring Diagram, Rev 20
8770-G-272, Main One Line Wiring Diagram, Rev 25  
8770-G-332, Sh. 1, 480V Miscellaneous, 125V DC and Vital AC One Line, Rev 23
8770-G-274, Auxiliary One Line Diagram, Rev 16  
8770-G-332, Sh. 2, 480V Miscellaneous, 125V DC and Vital AC One Line, Rev 6
8770-G-275, 6.9KV Swgr. & 4.16 KV Swgr. One Line Wiring Diagram, Rev 17  
8770-3639, CEA Drive MG Sets Elementary Connection Diagram, Rev 11
8770-G-275, Sh.2, 480V Swgr. & Pressurizer Htr. Bus One Line Wiring Diagram, Rev 20  
8770-5515, Sh. 3, Electrical Schematic, Safety Features Actuation System SB, Rev 20
8770-G-332, Sh. 1, 480V Miscellaneous, 125V DC and Vital AC One Line, Rev 23  
8770-5516, Sh. 3, Electrical Schematic, Safety Features Actuation System SA, Rev 19
8770-G-332, Sh. 2, 480V Miscellaneous, 125V DC and Vital AC One Line, Rev 6  
8770-5517, Sh. 1, Electrical Schematic, Safety Features Actuation System MA, Rev 14
8770-3639, CEA Drive MG Sets Elementary Connection Diagram, Rev 11  
8770-5518, Sh. 1, Electrical Schematic, Safety Features Actuation System MC, Rev 15
8770-5515, Sh. 3, Electrical Schematic, Safety Features Actuation System SB, Rev 20  
8770-5519, Sh. 1, Electrical Schematic, Safety Features Actuation System MB, Rev 15
8770-5516, Sh. 3, Electrical Schematic, Safety Features Actuation System SA, Rev 19  
8770-5520, Sh. 1, Electrical Schematic, Safety Features Actuation System MD, Rev 14
8770-5517, Sh. 1, Electrical Schematic, Safety Features Actuation System MA, Rev 14  
8770-12315, Sh. 2, Electrical Schematic, Safety Features Actuation System MD, Rev 0
8770-5518, Sh. 1, Electrical Schematic, Safety Features Actuation System MC, Rev 15  
8770-12316, Sh. 3, Electrical Schematic, Safety Features Actuation System MA, Rev 0
8770-5519, Sh. 1, Electrical Schematic, Safety Features Actuation System MB, Rev 15  
8770-12317, Sh. 2, Electrical Schematic, Safety Features Actuation System MB, Rev 0
8770-5520, Sh. 1, Electrical Schematic, Safety Features Actuation System MD, Rev 14  
8770-12318, Sh. 2, Electrical Schematic, Safety Features Actuation System MC, Rev 0
8770-12315, Sh. 2, Electrical Schematic, Safety Features Actuation System MD, Rev 0  
8770-G-083, sheet 1A, Flow Diagram Component Cooling System - Unit 1, Rev 59
8770-12316, Sh. 3, Electrical Schematic, Safety Features Actuation System MA, Rev 0  
8770-G-083, sheet 1B, Flow Diagram Component Cooling System - Unit 1, Rev 57
8770-12317, Sh. 2, Electrical Schematic, Safety Features Actuation System MB, Rev 0  
8770-G-227, sheet 1, Reactor Auxiliary Building Instrument Arrangement, Rev 21
8770-12318, Sh. 2, Electrical Schematic, Safety Features Actuation System MC, Rev 0  
8770-G-272, Unit 1 Main One Line Wiring Diagram, Rev 25
8770-G-083, sheet 1A, Flow Diagram Component Cooling System - Unit 1, Rev 59  
8770-G-274, Unit 1 Auxiliary One Line Diagram, Rev 17
8770-G-083, sheet 1B, Flow Diagram Component Cooling System - Unit 1, Rev 57  
8770-G275, sheet 1, 6.9KV Switchgear & 4.16KV Switchgear One Line Wiring Diagram, Rev 19
8770-G-227, sheet 1, Reactor Auxiliary Building Instrument Arrangement, Rev 21  
8770-B-327, sheet 202, Control Wiring Diagram Normal Supply Header & Normal Return
8770-G-272, Unit 1 Main One Line Wiring Diagram, Rev 25  
    Header Isolation Valves - Unit 1, Rev 6
8770-G-274, Unit 1 Auxiliary One Line Diagram, Rev 17  
8778-B-327, sheet 211, Component Cooling Water Shutdown Heat Exchanger & Surge Tank
8770-G275, sheet 1, 6.9KV Switchgear & 4.16KV Switchgear One Line Wiring Diagram, Rev 19  
    Fill Valves Unit 1 Control Wiring Diagrams, Rev 13
8770-B-327, sheet 202, Control Wiring Diagram Normal Supply Header & Normal Return  
8770-B-327, sheet 280, Safety Injection Tank 1A-2 Instrument and Check Valve Leakage Drain
Header Isolation Valves - Unit 1, Rev 6  
    to RWT HCV-3618 Unit 1 Control Wiring Diagram, Rev 19, 5/30/07
8778-B-327, sheet 211, Component Cooling Water Shutdown Heat Exchanger & Surge Tank  
8770-B-327, sheet 353, CCW Rad Mon Channels 56 and 57 Unit 1 Control Wiring Diagram,
Fill Valves Unit 1 Control Wiring Diagrams, Rev 13  
    Rev 6
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
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
6  
    Pumping System control, Rev 5
8770-B-327, sheet 532, Safeguards Room A Sump Pumps Unit 1 Control Wiring Diagram, Rev
Attachment
    9
8770-B-327, sheet 450, Control Wiring Diagram Process Radiation Channels 31 &32 & Iodine  
8770-B-327, sheet 533, Safeguards Room B Sump Pumps Unit 1 Control Wiring Diagram, Rev
Pumping System control, Rev 5  
    10
8770-B-327, sheet 532, Safeguards Room A Sump Pumps Unit 1 Control Wiring Diagram, Rev  
8770-B-327, sheet 583, Equipment Drain Sump Pump Unit 1 Control Wiring Diagram, Rev 6
9  
8770-B-327, sheet 906, Unit 1 Control Wiring Diagram Startup Transformer 1A-2 Breaker, Rev
8770-B-327, sheet 533, Safeguards Room B Sump Pumps Unit 1 Control Wiring Diagram, Rev  
    14
10  
8770-B-327, sheet 907, Unit 1 Control Wiring Diagram Startup Transformer 1B-2 Breaker, Rev
8770-B-327, sheet 583, Equipment Drain Sump Pump Unit 1 Control Wiring Diagram, Rev 6  
    17
8770-B-327, sheet 906, Unit 1 Control Wiring Diagram Startup Transformer 1A-2 Breaker, Rev  
8770-B-327, sheet 934, Unit 1 Control Wiring Diagram 4160 Switchgear 1A2 Feeder to Bus
14  
    1A3, Rev 13
8770-B-327, sheet 907, Unit 1 Control Wiring Diagram Startup Transformer 1B-2 Breaker, Rev  
8770-B-327, sheet 935, Unit 1 Control Wiring Diagram 4160 Switchgear 1B2 Feeder to Bus
17  
    1B3, Rev 17
8770-B-327, sheet 934, Unit 1 Control Wiring Diagram 4160 Switchgear 1A2 Feeder to Bus  
8770-B-327, sheet 948, Unit 1 Control Wiring Diagram 480 V Station Service Transformer 1B2
1A3, Rev 13  
    4160V Feeder Breaker, Rev 13
8770-B-327, sheet 935, Unit 1 Control Wiring Diagram 4160 Switchgear 1B2 Feeder to Bus  
8770-B-327, sheet 978, 480 V Switchgear 1A2-1AB Tie, Unit 1, Rev 9
1B3, Rev 17  
8770-B-327, sheet 979, 480 V Switchgear 1A2-1AB Tie, Unit 1, Rev 10
8770-B-327, sheet 948, Unit 1 Control Wiring Diagram 480 V Station Service Transformer 1B2  
8770-B-327, sheet 980, Control Wiring Diagram, 480 V Switchgear 1B2 Feeder Breaker, Rev 11
4160V Feeder Breaker, Rev 13  
2998-G-083, sheet 1, Flow Diagram Component Cooling System - Unit 2, Rev 41
8770-B-327, sheet 978, 480 V Switchgear 1A2-1AB Tie, Unit 1, Rev 9  
2998-B-327, sheet 202, Control Wiring Diagram Normal Supply Header & Normal Return
8770-B-327, sheet 979, 480 V Switchgear 1A2-1AB Tie, Unit 1, Rev 10  
    Header Isolation Valves - Unit 2, Rev 11
8770-B-327, sheet 980, Control Wiring Diagram, 480 V Switchgear 1B2 Feeder Breaker, Rev 11  
T/RCO/0711502-F1-R10, Unit 1 Main Power Distribution
2998-G-083, sheet 1, Flow Diagram Component Cooling System - Unit 2, Rev 41  
E-57953, 230KV Switchyard Operating Diagram, Rev 49
2998-B-327, sheet 202, Control Wiring Diagram Normal Supply Header & Normal Return  
8770-G-078SH.131, Flow Diagram Safety Injection System, Rev 19
Header Isolation Valves - Unit 2, Rev 11  
8770-G-088SH.2, Flow Diagram Containment Spray and Refueling Water Systems, Rev 52
T/RCO/0711502-F1-R10, Unit 1 Main Power Distribution  
8770-G-078SH.110, Flow Diagram Reactor Coolant System, Rev 30
E-57953, 230KV Switchyard Operating Diagram, Rev 49  
8770-G-083SH.1A, Flow Diagram Component Cooling System, Rev 59
8770-G-078SH.131, Flow Diagram Safety Injection System, Rev 19  
8770-G-078SH.121, Flow Diagram CVCS, Rev 39
8770-G-088SH.2, Flow Diagram Containment Spray and Refueling Water Systems, Rev 52  
Condition Reports (CRs)
8770-G-078SH.110, Flow Diagram Reactor Coolant System, Rev 30  
1998-1584, Unit 1 & 2 Charging Pump Surveillance Test Flow Rates Do Not Meet the Design
8770-G-083SH.1A, Flow Diagram Component Cooling System, Rev 59  
    Maximum Flow Rates
8770-G-078SH.121, Flow Diagram CVCS, Rev 39  
2005-1294, 1C CCW Pump Exceed The Maintenance Rule Unavailability Limit Of 200
    Hours/Year/Pump
Condition Reports (CRs)  
2005-2969, Letdown HX CCW Relief Was Found Lifting After 2A CCW Pump Start
1998-1584, Unit 1 & 2 Charging Pump Surveillance Test Flow Rates Do Not Meet the Design  
2005-30300, Inter-System LOCA Detection Instrumentation for Reactor Pressure Boundary Not
Maximum Flow Rates  
    Prioritize for Deficiency Resolution Prior to Mode 4
2005-1294, 1C CCW Pump Exceed The Maintenance Rule Unavailability Limit Of 200  
2007-27048, Incorrect Safety Classification of a DBD Function for Valve TCV-14/4A/4B
Hours/Year/Pump  
2007-28391, Parameter Limits for ICW Operability Performance Curves
2005-2969, Letdown HX CCW Relief Was Found Lifting After 2A CCW Pump Start  
2007-35587, PMs Being Changed From Daily to Outage During the Outage
2005-30300, Inter-System LOCA Detection Instrumentation for Reactor Pressure Boundary Not  
2008-31947, Air introduction into CCW System
Prioritize for Deficiency Resolution Prior to Mode 4  
2008-34697, Air introduction into CCW System per CR 2008-31947
2007-27048, Incorrect Safety Classification of a DBD Function for Valve TCV-14/4A/4B  
2008-35753, Isolate CCW to Containment IA Compressors Aftercoolers
2007-28391, Parameter Limits for ICW Operability Performance Curves  
2008-37070, St. Lucie Engineering Self-Assessment - Component Design Basis Inspection
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
2009-19025, Site Glass accidentally broken
                                                                                  Attachment


                                            7
2009-23473, CCW Surge Tank Design Basis Requirements for Code Pressure Relief Capacity
7  
    and Design
2005-6815, Low Margin Issue - Degraded Grid Action Plan
Attachment
2006-1885, 1A Battery Charger DC Output Breaker Lead Observed to Be Overheating
2009-23473, CCW Surge Tank Design Basis Requirements for Code Pressure Relief Capacity  
2006-19927, Develop PMCRs for New SF6 Breakers
and Design  
2006-20023, During Performance of Breaker PM, 480V Brkr Found with Misaligned Contact
2005-6815, Low Margin Issue - Degraded Grid Action Plan  
2006-20094, EDG Breaker Closure Failure during Post-Maintenance Testing of EDG
2006-1885, 1A Battery Charger DC Output Breaker Lead Observed to Be Overheating  
2006-22579, K-600 Breaker found to Have Several Problems
2006-19927, Develop PMCRs for New SF6 Breakers  
2006-25939, Spare Load Center Breaker Could Not Be Set per Procedure
2006-20023, During Performance of Breaker PM, 480V Brkr Found with Misaligned Contact  
2006-30383, UNUSED toc Switch Contacts Do Not Function Properly
2006-20094, EDG Breaker Closure Failure during Post-Maintenance Testing of EDG  
2007-1920, 480 V Swgr Breaker Received Refurbishment by ABB in Trip Free Mode
2006-22579, K-600 Breaker found to Have Several Problems  
2007-23473, EDG Loading Increase Due to Operation at Upper TS Frequency Limit
2006-25939, Spare Load Center Breaker Could Not Be Set per Procedure  
2007-4859, K3000 Breaker Refurbishment by ABB Unsatisfactory
2006-30383, UNUSED toc Switch Contacts Do Not Function Properly  
2007-7456, 480V Swgr Breaker Failed to Trip during Testing
2007-1920, 480 V Swgr Breaker Received Refurbishment by ABB in Trip Free Mode  
2007-8304, Problem Found on 480V Swgr Breaker after Refurbishment by ABB
2007-23473, EDG Loading Increase Due to Operation at Upper TS Frequency Limit  
2007-9889, Negative Trend in Performance of MCCBs Installed in MCCs
2007-4859, K3000 Breaker Refurbishment by ABB Unsatisfactory  
2007-10302, Hot Connection Observed on 1BB Battery Charger Neutral Lead Connection
2007-7456, 480V Swgr Breaker Failed to Trip during Testing  
2007-13704, Review of IN 2007-09, Equipment Operability Under Degraded Voltage
2007-8304, Problem Found on 480V Swgr Breaker after Refurbishment by ABB  
      Conditions.
2007-9889, Negative Trend in Performance of MCCBs Installed in MCCs  
2007-14099, 1A 125V DC System Swgr Undervoltage Relay out of Adjustment
2007-10302, Hot Connection Observed on 1BB Battery Charger Neutral Lead Connection  
2007-15321, 4.16KV Breaker for 1A LPSI Pump Did Not Charge Spring when Racked in
2007-13704, Review of IN 2007-09, Equipment Operability Under Degraded Voltage  
2007-28789, Track & Trend CR for Revision of DC Calculation due to Battery Inter-Cell
Conditions.  
    Resistance.
2007-14099, 1A 125V DC System Swgr Undervoltage Relay out of Adjustment  
2007-29985, Jumpering-out of two battery cells
2007-15321, 4.16KV Breaker for 1A LPSI Pump Did Not Charge Spring when Racked in  
2007-34306, Medium Voltage Breaker Cluster Finger Problem
2007-28789, Track & Trend CR for Revision of DC Calculation due to Battery Inter-Cell  
2007-36484, 1D Battery Charger Control Board B Found to Be Defective during PM
Resistance.  
2007-39837, 480V Swgr 2A3-5B Loose Breaker Power Stab
2007-29985, Jumpering-out of two battery cells  
2008-14926, Adequacy of 1-OSP-50.01 to satisfy NRC Commitment to Test UV Trip
2007-34306, Medium Voltage Breaker Cluster Finger Problem  
Feature of DC Cross-Tie Breakers
2007-36484, 1D Battery Charger Control Board B Found to Be Defective during PM  
2008-26139, Hard Ground on 125V DC Bus 1BDefective Masterpact Circuit Breaker Trip Unit
2007-39837, 480V Swgr 2A3-5B Loose Breaker Power Stab  
2008-33033, Defective Masterpact Circuit Breaker Trip Unit
2008-14926, Adequacy of 1-OSP-50.01 to satisfy NRC Commitment to Test UV Trip
2008-35540, 1A EDG Output Breaker Failed to Open during Engineered Safeguard Test
Feature of DC Cross-Tie Breakers  
2009-8276, Potential Part 21 Notification for ABB K-Line Circuit Breakers
2008-26139, Hard Ground on 125V DC Bus 1BDefective Masterpact Circuit Breaker Trip Unit
2009-9055, Potential Part 21 from ABB Due to Possible Tension Spring Failure
2008-33033, Defective Masterpact Circuit Breaker Trip Unit
2009-15659, 480V Swgr Breaker Had Trip Indication Illuminated
2008-35540, 1A EDG Output Breaker Failed to Open during Engineered Safeguard Test  
2009-15807, During the ESF Testing, the CEA MG Set Breaker Failed to Trip as Expected
2009-8276, Potential Part 21 Notification for ABB K-Line Circuit Breakers  
2007-12838, HVS-1C field cables megger readings were identified out of spec low, 4/27/2007
2009-9055, Potential Part 21 from ABB Due to Possible Tension Spring Failure  
2005-10351,Potentail for Motor Degradation, 4/11/2005
2009-15659, 480V Swgr Breaker Had Trip Indication Illuminated  
2008-21053, Leakage into the 1A2 Safety Injection Tank, 6/26/2008
2009-15807, During the ESF Testing, the CEA MG Set Breaker Failed to Trip as Expected  
2006-17344, V1403 Did Not Stoke Closed as Expected, 6/4/2005
2007-12838, HVS-1C field cables megger readings were identified out of spec low, 4/27/2007  
2009-20554, SIT Outlet Valves V3614 and V3624 Failed to Open, 7/20/2009
2005-10351,Potentail for Motor Degradation, 4/11/2005  
2007-42630, 2A1 SIT Iso Valve, V3624 Breaker Trip, 12/26/2007
2008-21053, Leakage into the 1A2 Safety Injection Tank, 6/26/2008  
2005-1469, 2A2 SIT Iso Valve Failed to Open, 1/23/2005LOL
2006-17344, V1403 Did Not Stoke Closed as Expected, 6/4/2005  
2004-9733, SIT Outlet Valve V3614 Failed to Open.
2009-20554, SIT Outlet Valves V3614 and V3624 Failed to Open, 7/20/2009  
Completed Work Orders (WOs)
2007-42630, 2A1 SIT Iso Valve, V3624 Breaker Trip, 12/26/2007  
38025447, Air Leaked into CCW - Troubleshoot, dated 11/18/08
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
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
8  
34013031-01, Pressurizer Pressure P-1103/1104 Loop Calibration, 11/5/05
34013455-01, 125V DC Battery 1B Performance Maintenance, 11/22/05
Attachment
35027963-01, 125V DC Battery 1B Capacity Test, 11/22/05
33001113-01, 125V DC Battery 1B Capacity Test, 4/9/04  
36001038-01,125V DC Battery 1B Battery Profile Test, 4/2/07
34013031-01, Pressurizer Pressure P-1103/1104 Loop Calibration, 11/5/05  
36001576-01, 125V DC Battery 1B Performance Maintenance, 4/9/07
34013455-01, 125V DC Battery 1B Performance Maintenance, 11/22/05  
36006266-01,125V DC System Battery Charger 1B 18-Month Operability Test, 7/3/07
35027963-01, 125V DC Battery 1B Capacity Test, 11/22/05  
36008706-01, PT-1103 EQ Rosemount Replacement, 4/18/07
36001038-01,125V DC Battery 1B Battery Profile Test, 4/2/07  
36008707-01, PT-1104 EQ Rosemount Replacement, 4/15/07
36001576-01, 125V DC Battery 1B Performance Maintenance, 4/9/07  
37011755-01,125V DC Battery 1B Battery Profile Test, 11/13/08
36006266-01,125V DC System Battery Charger 1B 18-Month Operability Test, 7/3/07  
37011878-01, 125V DC Battery 1B Performance Maintenance, 11/14/08
36008706-01, PT-1103 EQ Rosemount Replacement, 4/18/07  
37019185-01,125V DC System Battery Charger 1B 18-Month Operability Test, 7/17/08
36008707-01, PT-1104 EQ Rosemount Replacement, 4/15/07  
37024229-01, Pressurizer Pressure PT1103 Transmitter Calibration, 11/6/08
37011755-01,125V DC Battery 1B Battery Profile Test, 11/13/08  
37024231-01, Pressurizer Pressure PT1104 Transmitter Calibration, 11/5/08
37011878-01, 125V DC Battery 1B Performance Maintenance, 11/14/08  
38003350-01, ATWS Functional Test, 6/6/09
37019185-01,125V DC System Battery Charger 1B 18-Month Operability Test, 7/17/08  
38005460-01, Unit 1 Battery Charger 1AA Charger PM, 7/2/08
37024229-01, Pressurizer Pressure PT1103 Transmitter Calibration, 11/6/08  
38008615-01, Pressurizer Pressure P-1103/1104 Loop Calibration, 12/5/08
37024231-01, Pressurizer Pressure PT1104 Transmitter Calibration, 11/5/08  
38013076-01, 125V DC Battery 1B Quarterly PM, 1/6/09
38003350-01, ATWS Functional Test, 6/6/09  
38025268-01, 125V DC Battery 1B Quarterly PM, 3/19/09
38005460-01, Unit 1 Battery Charger 1AA Charger PM, 7/2/08  
39001705-01, Engineered Safeguards Monthly, 6/21/09
38008615-01, Pressurizer Pressure P-1103/1104 Loop Calibration, 12/5/08  
39001717-01, 125V DC Battery 1B Quarterly PM, 5/28/09
38013076-01, 125V DC Battery 1B Quarterly PM, 1/6/09  
39003272-01, ESFAS Monthly PM, 6/9/2009
38025268-01, 125V DC Battery 1B Quarterly PM, 3/19/09  
W/R 39005930, Replace relay 27-4, 8/6/09
39001705-01, Engineered Safeguards Monthly, 6/21/09  
W/R 37008352, Oil leak by north oil pump on Startup Transformer 1B, 7/30/07
39001717-01, 125V DC Battery 1B Quarterly PM, 5/28/09  
W/R 38013946, Breaker binding when racking in or out, 11/12/08
39003272-01, ESFAS Monthly PM, 6/9/2009  
W/O 34020981, Calibration of Safeguards Room B Level Alarm Switch LS-06-1B and High-High
W/R 39005930, Replace relay 27-4, 8/6/09  
    Alarm Level Switch LS-06-41
W/R 37008352, Oil leak by north oil pump on Startup Transformer 1B, 7/30/07  
W/O 38005217, Calibration of Safeguards Room B Level Alarm Switch LS-06-1B and High-High
W/R 38013946, Breaker binding when racking in or out, 11/12/08  
    Alarm Level Switch LS-06-41
W/O 34020981, Calibration of Safeguards Room B Level Alarm Switch LS-06-1B and High-High  
W/O 34020329, Calibration of Safeguards Room A Level Alarm Switch LS-06-1A and High-High
Alarm Level Switch LS-06-41  
    Alarm Level Switch LS-06-40
W/O 38005217, Calibration of Safeguards Room B Level Alarm Switch LS-06-1B and High-High  
W/O 38011496, Channel 31 and 21 18 month Calibration, 7/9/08
Alarm Level Switch LS-06-41  
W/O 37017925, RE 26-56 & 57 Calibration, 8/16/07
W/O 34020329, Calibration of Safeguards Room A Level Alarm Switch LS-06-1A and High-High  
W/O 32013060, Spare Breaker PM, 12/18/03
Alarm Level Switch LS-06-40  
W/O 33016593, Breaker 1B2-7B 54 Month PM, 9/10/04
W/O 38011496, Channel 31 and 21 18 month Calibration, 7/9/08  
W/O 33022130, Breaker 1B2-2C 54 Month PM, 7/27/04
W/O 37017925, RE 26-56 & 57 Calibration, 8/16/07  
W/O 34018962, Breaker 1A1-7B 54 Month PM, 3/9/06
W/O 32013060, Spare Breaker PM, 12/18/03  
W/O 35002739, 4/16KV SWGR 1B3-2 Breaker Replacement and Testing , 6/3/05
W/O 33016593, Breaker 1B2-7B 54 Month PM, 9/10/04  
W/O 35009733, Breaker 1A1-5D PM and Swap, 7/15/05
W/O 33022130, Breaker 1B2-2C 54 Month PM, 7/27/04  
W/O 35020492, Breaker 1B2-6B 54 Month PM, 2/13/06
W/O 34018962, Breaker 1A1-7B 54 Month PM, 3/9/06  
W/O 36008492, Breaker 1B2-7A PM and Swap, 10/04/06
W/O 35002739, 4/16KV SWGR 1B3-2 Breaker Replacement and Testing , 6/3/05  
WO31022173-01, V3106 Check Valve Inspection
W/O 35009733, Breaker 1A1-5D PM and Swap, 7/15/05  
WO31022495-01, V07174 IST Check Valve Inspection
W/O 35020492, Breaker 1B2-6B 54 Month PM, 2/13/06  
WO33003927-01, V07172 IST Check Valve Inspection
W/O 36008492, Breaker 1B2-7A PM and Swap, 10/04/06  
WO34019438-01, V07174 IST Check Valve Inspection
WO31022173-01, V3106 Check Valve Inspection  
WO36000672-01, V07172 IST Check Valve Inspection
WO31022495-01, V07174 IST Check Valve Inspection  
WO37015831-01, V07174 IST Check Valve Inspection
WO33003927-01, V07172 IST Check Valve Inspection  
WO38018501-01, V07174 IST Check Valve Inspection
WO34019438-01, V07174 IST Check Valve Inspection  
                                                                              Attachment
WO36000672-01, V07172 IST Check Valve Inspection  
WO37015831-01, V07174 IST Check Valve Inspection  
WO38018501-01, V07174 IST Check Valve Inspection  


                                          9
Modifications
9  
Change Request Notice CRN 03167-13230, Vendor Manual for Shutdown Cooling Heat
    Exchanger Update to Show the Correct Tube Plugs, Rev. 0
Attachment
Change Request Notice CRN 18362, Install Temporary Protection on LG-14-2A and LG-14-2B,
Modifications  
    Rev. 0
Change Request Notice CRN 03167-13230, Vendor Manual for Shutdown Cooling Heat  
Change Request Notice CRN 00048-9446, Permanent Removal of Gravity Damper Cover
Exchanger Update to Show the Correct Tube Plugs, Rev. 0  
    Plates on GD-1 and GD-2, Rev. 0
Change Request Notice CRN 18362, Install Temporary Protection on LG-14-2A and LG-14-2B,  
Miscellaneous Documents
Rev. 0  
DBD-CCW-1, Component Cooling Water System, Rev. 2
Change Request Notice CRN 00048-9446, Permanent Removal of Gravity Damper Cover  
DBD-ICW-1, Intake Cooling Water System, Rev. 2
Plates on GD-1 and GD-2, Rev. 0  
DBD-HVAC-1, Safety Related HVAC Systems, Rev. 2
DBD-120V-AC-1, Class 1E 120 V AC Power System, Rev 2
Miscellaneous Documents  
DBD-480V-AC-1, 480 VAC Distribution System, Rev 2A
DBD-CCW-1, Component Cooling Water System, Rev. 2  
DBD-4160 VAC-1, 4160 VAC Distribution System, Rev 2
DBD-ICW-1, Intake Cooling Water System, Rev. 2  
DBD-EDG-1, Emergency Diesel Generatoor System, Rev 3
DBD-HVAC-1, Safety Related HVAC Systems, Rev. 2  
DBD-ESF-1, Engineered Safety Features Actuation System, Rev 2
DBD-120V-AC-1, Class 1E 120 V AC Power System, Rev 2  
DBD-HVAC-1, Safety Related HVAC Systems Design Basis Document, Rev 2
DBD-480V-AC-1, 480 VAC Distribution System, Rev 2A  
DBD-PZR-1, Pressurizer System, Rev 2
DBD-4160 VAC-1, 4160 VAC Distribution System, Rev 2  
DBD-VDC-1, Class 1E DC Electrical Distribution System, Rev 2
DBD-EDG-1, Emergency Diesel Generatoor System, Rev 3  
8770-5756, Component Cooling Water Pump, Rev. 6
DBD-ESF-1, Engineered Safety Features Actuation System, Rev 2  
8770-7248, I/M Centrifugal Fans HVS-4A, 4B, 5A, 5B, HVE-1, 2, 4, 5, 7A, 7B, 8A, 8B, 9A, 9B,
DBD-HVAC-1, Safety Related HVAC Systems Design Basis Document, Rev 2  
    10A, 10B, 15, 16A, 16B, 21A, 21B, 13A, 13B, 14 & 33, Rev. 5
DBD-PZR-1, Pressurizer System, Rev 2  
0711209, Component Cooling Water System, Rev. 12
DBD-VDC-1, Class 1E DC Electrical Distribution System, Rev 2  
0702209, Component Cooling Water System, Rev. 8
8770-5756, Component Cooling Water Pump, Rev. 6  
Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient, Dated July 25,
8770-7248, I/M Centrifugal Fans HVS-4A, 4B, 5A, 5B, HVE-1, 2, 4, 5, 7A, 7B, 8A, 8B, 9A, 9B,  
    1984
10A, 10B, 15, 16A, 16B, 21A, 21B, 13A, 13B, 14 & 33, Rev. 5  
EPO-84-1662, CCW Surge Tank Overpressurization, Dated, August 20, 1984
0711209, Component Cooling Water System, Rev. 12  
SLN-88-021-10-20, JPN-PSL-SEICP-92-28, Evaluation of the Design basis for Fisher & Porter
0702209, Component Cooling Water System, Rev. 8  
    Indicating Controllers for Temperature Control Valves TCV-14-4A and TCV-14-4B.
Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient, Dated July 25,  
JPN-PSL-SENP-93-001, Inputs for the LOCA Containment Re-Analysis, Rev. 0
1984  
NRC NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, April 1995
EPO-84-1662, CCW Surge Tank Overpressurization, Dated, August 20, 1984  
NRC Information Notice 2008-02: Findings Identified During Component Design Bases
SLN-88-021-10-20, JPN-PSL-SEICP-92-28, Evaluation of the Design basis for Fisher & Porter  
    Inspections, March 19, 2008
Indicating Controllers for Temperature Control Valves TCV-14-4A and TCV-14-4B.  
FPL-09-366, Westinghouse Letter to Phil Barnes, CCW Flowrates Used in Containment
JPN-PSL-SENP-93-001, Inputs for the LOCA Containment Re-Analysis, Rev. 0  
    Analysis for St. Lucie, September 2, 2009
NRC NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, April 1995  
JPN-PSL-SEMP-91-007, Safety Evaluation for RAB Electrical Equipment and Battery Room
NRC Information Notice 2008-02: Findings Identified During Component Design Bases  
    HVAC, St. Lucie Unit 1, Rev 1
Inspections, March 19, 2008  
0711401, Engineered Safety Features Actuation System, Rev 1
FPL-09-366, Westinghouse Letter to Phil Barnes, CCW Flowrates Used in Containment  
00809-0100-4388, Rosemount 1153 Series D Alphaline Nuclear Pressure Transmitter, Rev BA
Analysis for St. Lucie, September 2, 2009  
2998-15662, Instruction Manual for Safety Features Actuation System (SFAS) Vol. 1 & Vol. 2,
JPN-PSL-SEMP-91-007, Safety Evaluation for RAB Electrical Equipment and Battery Room  
    Manual No. TM9N38, Rev 8
HVAC, St. Lucie Unit 1, Rev 1  
8770-7423, Instruction Manual Turbine Building Battery Charger, C&D Battery, Manual No.
0711401, Engineered Safety Features Actuation System, Rev 1  
    MCB-2010, Rev 5
00809-0100-4388, Rosemount 1153 Series D Alphaline Nuclear Pressure Transmitter, Rev BA  
8770-10459, Instruction Manual for Battery Chargers 1AA, 1BB, and 1D, C&D Battery, Manual
2998-15662, Instruction Manual for Safety Features Actuation System (SFAS) Vol. 1 & Vol. 2,  
    No. RS-421, Rev 5
Manual No. TM9N38, Rev 8  
                                                                                  Attachment
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


                                            10
8770-15662, Instruction Manual for Safety Features Actuation System (SFAS) Vol 1 & Vol. 2,
10  
      Manual No. TM9N38, Rev 8
Unit 1 System 47, 480 VAC System Health Report, 6/30/2009
Attachment
Unit 1 System 50, 125V DC System Health Report, 6/30/2009
8770-15662, Instruction Manual for Safety Features Actuation System (SFAS) Vol 1 & Vol. 2,  
Unit 1 System 52, 4.16 KV System Health Report, 6/30/2009
Manual No. TM9N38, Rev 8  
Unit 1 System 63, Reactor Protection System Health Report, 6/30/09
Unit 1 System 47, 480 VAC System Health Report, 6/30/2009  
IEEE Standard 450-1995, IEEE Recommended Practice for Maintenance, Testing, and
Unit 1 System 50, 125V DC System Health Report, 6/30/2009
Replacement of Vented Lead-Acid for Stationary Applications.
Unit 1 System 52, 4.16 KV System Health Report, 6/30/2009  
Vendor Manual 8770-15227, OTEK HI-Q2000 Instruction Manual, Rev 1, 5/11/.06
Unit 1 System 63, Reactor Protection System Health Report, 6/30/09  
Vendor Manual 8770-12474, Beckman Industrial Model 500T Digital Panel Indicator Operators
IEEE Standard 450-1995, IEEE Recommended Practice for Maintenance, Testing, and  
      Manual, Rev 0, 2/13/91
Replacement of Vented Lead-Acid for Stationary Applications.  
L-2007-067, Response to Generic Letter 2007-01, 5/8/2007
Vendor Manual 8770-15227, OTEK HI-Q2000 Instruction Manual, Rev 1, 5/11/.06  
Component Evaluation Sheet, page 2, Rosemount 1153-GD-7 Series D Pressure Transmitter,
Vendor Manual 8770-12474, Beckman Industrial Model 500T Digital Panel Indicator Operators  
      Rev 11
Manual, Rev 0, 2/13/91  
Maintenance Rule Scoping for Switchyard System, Rev 3
L-2007-067, Response to Generic Letter 2007-01, 5/8/2007  
Maintenance Rule Scoping for 480V Switchgear, Breakers and MCCs
Component Evaluation Sheet, page 2, Rosemount 1153-GD-7 Series D Pressure Transmitter,  
Nuclear Plant Switchyard Inspection Report (weekly) for breakers 8W23, 8W40, 8W61 and
Rev 11  
      8W64
Maintenance Rule Scoping for Switchyard System, Rev 3  
Control Room Log, 7/19-21/2009
Maintenance Rule Scoping for 480V Switchgear, Breakers and MCCs  
CRs and WOs Initiated Due to CDBI Activity:
Nuclear Plant Switchyard Inspection Report (weekly) for breakers 8W23, 8W40, 8W61 and  
2009-22430, Inadequate housekeeping in the PSL1 CCW Surge Tank Room
8W64  
2009-22556, Lid on Head Tank for Instrument Air Compressor Cooling Water Fan Cooler Was
Control Room Log, 7/19-21/2009  
      Rusted Shut
2009-22623, Alignment of the Non-Essential CCW Header to the Only Remaining Essential
CRs and WOs Initiated Due to CDBI Activity:  
      CCW Header Under Certain Accident Scenarios Could Potentially Place the Plant in the
2009-22430, Inadequate housekeeping in the PSL1 CCW Surge Tank Room
      Unanalyzed Condition
2009-22556, Lid on Head Tank for Instrument Air Compressor Cooling Water Fan Cooler Was  
2009-22766, Instrument Air Compressor Cooling Water Cooler Original Design Documentation
Rusted Shut  
      Can Not Be Located
2009-22623, Alignment of the Non-Essential CCW Header to the Only Remaining Essential  
2009-22811, Two Steel Angles (Support on HVAC duct) Extend Down Into The Walk Path
CCW Header Under Certain Accident Scenarios Could Potentially Place the Plant in the  
      Around The East Side Of The Surge Tank In The Unit 1 CCW Surge Tank Room
Unanalyzed Condition  
2009-22892, 1(2)A and 1(2)B Instrument Air Compressor Emergency Cooling Lineup Issues
2009-22766, Instrument Air Compressor Cooling Water Cooler Original Design Documentation  
2009-22929, A NRC inspector for the CBDI team has questioned the operability determination
Can Not Be Located  
      previously done for Air Intrusion into CCW Event from October, 2008
2009-22811, Two Steel Angles (Support on HVAC duct) Extend Down Into The Walk Path  
2009-22959, Missing Information from Calculation PSL-1CHN-93-002, Rev. 0 about 3 Plugged
Around The East Side Of The Surge Tank In The Unit 1 CCW Surge Tank Room  
      Tubes in the 1A Shutdown Cooling Heat Exchanger
2009-22892, 1(2)A and 1(2)B Instrument Air Compressor Emergency Cooling Lineup Issues  
2009-23011, CCW Pump IST Procedure (1-OSP-14.01A/B/C) does not address affect (sic) of
2009-22929, A NRC inspector for the CBDI team has questioned the operability determination  
      pump degradation on SIAS CCW System flow rates
previously done for Air Intrusion into CCW Event from October, 2008  
2009-23473, CCW surge tank design basis requirements for code pressure relief capacity and
2009-22959, Missing Information from Calculation PSL-1CHN-93-002, Rev. 0 about 3 Plugged  
      design
Tubes in the 1A Shutdown Cooling Heat Exchanger  
2009-23882, Investigate possible sources of air ingress into the CCW System on PSL1 and
2009-23011, CCW Pump IST Procedure (1-OSP-14.01A/B/C) does not address affect (sic) of  
      PSL2
pump degradation on SIAS CCW System flow rates  
2009-24030, A Past Operability Review of an Air Intrusion into CCW event from October 2008
2009-23473, CCW surge tank design basis requirements for code pressure relief capacity and  
2009-25209, Evaluation of the Air/Gas Intrusion into CCW event from 10/16/08 which was
design  
      documented in 3/C CR 2008-34697
2009-23882, Investigate possible sources of air ingress into the CCW System on PSL1 and  
2009-25276, Unit 1 CCW Surge Tank Overpressure Protection configuration is not in
PSL2  
      compliance with the ASME Code
2009-24030, A Past Operability Review of an Air Intrusion into CCW event from October 2008  
                                                                                  Attachment
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


                                          11
2009-17349, Calculation PSL-1-FSE-002, Rev. 1, Transmitted to Document Control as the
11  
    Calculation of Record Without an FPL Acceptance Signature
2009-22338, Loose Tools Identified by NRC at 2A4/2B4 Switchgear and 1AB Load Center.
Attachment
2009-22998, Technical Specification Battery Inter-Cell Connection Resistance Limit of 150
2009-17349, Calculation PSL-1-FSE-002, Rev. 1, Transmitted to Document Control as the  
    Micro-Ohms not Used in DC System Analysis.
Calculation of Record Without an FPL Acceptance Signature
2009-22999, Possible Calculation Procedure Enhancement.
2009-22338, Loose Tools Identified by NRC at 2A4/2B4 Switchgear and 1AB Load Center.  
2009-24649, Current Revision of Calculation PSL-1FSE-05-002 Does not Reflect the As-Built
2009-22998, Technical Specification Battery Inter-Cell Connection Resistance Limit of 150  
    status of Unit 1.
Micro-Ohms not Used in DC System Analysis.  
2009-25088, During SL1-22 Both Narrow Range Pressurizer Pressure Transmitters Found out
2009-22999, Possible Calculation Procedure Enhancement.  
    of Calibration High.
2009-24649, Current Revision of Calculation PSL-1FSE-05-002 Does not Reflect the As-Built  
2009-25178, Battery Profile (Service) Test Procedure Enhancement
status of Unit 1.  
2009-22338, Loose Tools Identified by NRC at 2A4/2B4 Switchgear and 1AB Load Center,
2009-25088, During SL1-22 Both Narrow Range Pressurizer Pressure Transmitters Found out  
    8/5/2009
of Calibration High.  
                                                                                    Attachment
2009-25178, Battery Profile (Service) Test Procedure Enhancement  
2009-22338, Loose Tools Identified by NRC at 2A4/2B4 Switchgear and 1AB Load Center,  
8/5/2009


                                        PHASE III ANALYSIS
SRA Analysis Number: STL0904
Attachment
Analysis Type: SDP Phase III
PHASE III ANALYSIS  
Inspection Report: 05000335, 389/2009006
Plant Name: St. Lucie
Unit Numbers: 1 & 2
SRA Analysis Number: STL0904  
Enforcement Action EA-09-321
Analysis Type: SDP Phase III  
BACKGROUND - Air intrusion into the CCW system occurred on October 16, 2008, and was
Inspection Report: 05000335, 389/2009006
originally documented in CR 2008-31947. This air intrusion event on Unit 1 affected the CCW
Plant Name: St. Lucie  
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
Unit Numbers: 1 & 2  
compressors provided a pathway for which air intrusion into the system occurred. This
Enforcement Action EA-09-321  
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
BACKGROUND - Air intrusion into the CCW system occurred on October 16, 2008, and was  
is approximately 120 psig. Original design deficiency: Non-safety related instrument air
originally documented in CR 2008-31947. This air intrusion event on Unit 1 affected the CCW  
compressor inside containment (Unit 1 only) and waste gas air compressor (both units) provide
system to the extent that both operating CCW pumps, one in each train, were cavitating as  
a common vulnerability for safety related component cooling water (CCW) system. FSAR
evidenced by fluctuating amp indication. It was identified that the containment instrument air  
section 9.2.2.3.2, Single Failure Analysis for the CCW system, states in part: there is no single
compressors provided a pathway for which air intrusion into the system occurred. This  
failure that could prevent the component cooling system from performing its safety function.
vulnerability also exists, on both units, should the aftercoolers on the waste gas compressors  
Therefore, the air intrusion that affected both trains of the CCW system was a significant
fail. The waste gas compressors run at approximately 160 psig and the CCW system pressure  
condition adverse to quality.
is approximately 120 psig. Original design deficiency: Non-safety related instrument air  
PERFORMANCE DEFICIENCY - Section 4OA5 of the report discusses the air intrusion in
compressor inside containment (Unit 1 only) and waste gas air compressor (both units) provide  
detail. The air intrusion potentially rendered both trains of the safety-related CCW system
a common vulnerability for safety related component cooling water (CCW) system. FSAR  
inoperable. Two performance deficiencies were identified associated with this issue. The first
section 9.2.2.3.2, Single Failure Analysis for the CCW system, states in part: there is no single  
performance deficiency involved a common cause failure vulnerability of the CCW system.
failure that could prevent the component cooling system from performing its safety function.
Specifically, a non-safety system failure could result in a common cause failure of both trains of
Therefore, the air intrusion that affected both trains of the CCW system was a significant  
the CCW system. The second performance deficiency involved the failure to identify and
condition adverse to quality.  
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
PERFORMANCE DEFICIENCY - Section 4OA5 of the report discusses the air intrusion in  
following the air intrusion event that occurred in October 2008. Further, the licensees corrective
detail. The air intrusion potentially rendered both trains of the safety-related CCW system  
action evaluation did not identify the common cause failure vulnerability discussed in the first
inoperable. Two performance deficiencies were identified associated with this issue. The first  
performance deficiency.
performance deficiency involved a common cause failure vulnerability of the CCW system.
EXPOSURE TIME - One year will be used.
Specifically, a non-safety system failure could result in a common cause failure of both trains of  
DATE OF OCCURRENCE - October 2008
the CCW system. The second performance deficiency involved the failure to identify and  
SAFETY IMPACT > Green
correct a condition adverse to quality. Specifically, the licensee failed to properly determine the  
RISK ANALYSIS/CONSIDERATIONS
source of the air in-leakage into the CCW system and take appropriate corrective actions  
Assumptions
following the air intrusion event that occurred in October 2008. Further, the licensees corrective  
1. The performance deficiency will be modeled as an increase in the probability of an initiating
action evaluation did not identify the common cause failure vulnerability discussed in the first  
event, Loss of the CCW system.
performance deficiency.  
                                                                                        Attachment
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.


                                                2
2. With respect to Unit 1 the performance deficiency caused a failure or an imminent failure of
2  
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.
Attachment
3. Given the response of the operators to the abnormal condition of the CCW system, recovery
2. With respect to Unit 1 the performance deficiency caused a failure or an imminent failure of  
credit is appropriate. A 0.1 failure probability will be assigned to operators failing to recognize
the CCW system. Given the condition of the pumps and the surge tank level perturbations, the  
and mitigate the air intrusion before air binding of the pumps happens.
probability of failure will be set at 1.0 for the one year exposure time.  
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
3. Given the response of the operators to the abnormal condition of the CCW system, recovery  
their operating licenses were issued. Recovery will be applied here also.
credit is appropriate. A 0.1 failure probability will be assigned to operators failing to recognize  
5. No recovery will be considered after air intrusion severe enough to cause CCW pump failure.
and mitigate the air intrusion before air binding of the pumps happens.  
6. The non-conforming case will be considered the delta core damage frequency case. This is
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-
due to at least a magnitude shift in the core damage frequency results between the non-
conforming and conforming cases.
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
PRA Model used for basis of the risk analysis: Licensees full scope model  
ability to perform its numerous risk significant functions.
CALCULATIONS
Significant Influence Factor(s) [if any]: How severe the air intrusion was on the CCW systems  
The top 10,000 cutsets from the full scope model were screened for a loss of CCW system
ability to perform its numerous risk significant functions.  
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
CALCULATIONS  
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
The top 10,000 cutsets from the full scope model were screened for a loss of CCW system  
and represents the conditional core damage probability given a Loss of CCW. This CCDP was
initiator. A loss of Train A Surge Tank and Train B in test and maintenance was selected.
comparable to SPAR in the GEM mode.
Those cutsets with these events were extracted and are shown in Appendix 2. Once the  
Applying the Unit 1 non-conforming case initiating event frequency of 1.0 yields a core damage
initiating event is removed, only one basic event remained in the accident sequence, operators  
frequency of 3.3E-3 for the exposure period. Applying the non-recovery term (see Attachment 3
fail to trip the operating Reactor Coolant Pumps. This basic event failure probability was 3.3E-3  
for its detailed development) of 0.1 yields a core damage frequency of 3.3E-4 for the exposure
and represents the conditional core damage probability given a Loss of CCW. This CCDP was  
period.
comparable to SPAR in the GEM mode.  
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
Applying the Unit 1 non-conforming case initiating event frequency of 1.0 yields a core damage  
final core damage frequency of 6E-6 for the exposure period.
frequency of 3.3E-3 for the exposure period. Applying the non-recovery term (see Attachment 3  
EXTERNAL EVENTS CONSIDERATIONS - Due to the nature of the performance deficiency
for its detailed development) of 0.1 yields a core damage frequency of 3.3E-4 for the exposure  
which increases the frequency of an internal events initiator, external events consideration is not
period.
warranted.
LARGE EARLY RELEASE FREQUENCY IMPACT - Since there is not an increase in SGTR or
Applying the Unit 2 non-conforming case initiating event frequency of 1.8E-2/yr to the CCDP of  
ISLOCA accident sequences, LERF is not the appropriate decision making metric.
3.3E-3 yields a core damage frequency of 6E-5. Applying the non-recovery term of 0.1 yields a  
                                                                                          Attachment
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.


                                                3
RECONCILIATION BETWEEN PHASE III AND PLANT NOTEBOOK/ PHASE II RESULTS -
3  
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
Attachment
sequence is assigned a nominal value of 6 - four for the initiating event frequency and two for
RECONCILIATION BETWEEN PHASE III AND PLANT NOTEBOOK/ PHASE II RESULTS -  
the operator error. Phase III results in a lower probability of operators tripping the RCP of 3E-3.
The dominant accident sequence from the Phase II Notebook is Loss of CCW followed by  
Therefore, the color is the same in both phases but, numerically a magnitude higher in the
operators failing to trip the RCP leading directly to a large seal LOCA and core damage. The  
Phase II result. This shows reconciliation between the two phases.
sequence is assigned a nominal value of 6 - four for the initiating event frequency and two for  
CONCLUSIONS/RECOMMENDATIONS - Given the present information associated with the air
the operator error. Phase III results in a lower probability of operators tripping the RCP of 3E-3.
intrusion of October 2008, it is reasonable to assume the initiating event frequency increased by
Therefore, the color is the same in both phases but, numerically a magnitude higher in the  
at least a magnitude. Such a shift with recovery is in the White zone of safety characterization.
Phase II result. This shows reconciliation between the two phases.  
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
CONCLUSIONS/RECOMMENDATIONS - Given the present information associated with the air  
characterized as >Green with the intent to acquire as much information about air intrusion into
intrusion of October 2008, it is reasonable to assume the initiating event frequency increased by  
the CCW system as:
at least a magnitude. Such a shift with recovery is in the White zone of safety characterization.
            *   an initiator for Loss of the CCW system
Assuming that CCW was in imminent failure the safety characterization shifts into the red zone,  
            *   an undetected failure mechanism of any CCW functions while the equipment is in
even with recovery. Therefore, this performance deficiency should be preliminarily  
                standby
characterized as >Green with the intent to acquire as much information about air intrusion into  
APPENDICES: 1. Full Scope Model Output
the CCW system as:  
                  2. Recovery Development
      Analyst: W. Rogers               Date: 10/30/09
*  
      Reviewed By: G. MacDonald Date: 11/02/2009
an initiator for Loss of the CCW system  
                                                                                        Attachment
*  
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  


EDITED FULL SCOPE MODEL CUTSETS FOR TOTAL LOSS OF COMPONENT COOLING WATER SYSTEM
      TOP 10,000 Cutsets for PSL1
Appendix 1
C:\CAFTA32\06R0B\PSL1\Cuts-B\PSL1Top10K.CUT
EDITED FULL SCOPE MODEL CUTSETS FOR TOTAL LOSS OF COMPONENT COOLING WATER SYSTEM  
    # 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
TOP 10,000 Cutsets for PSL1  
                          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)
C:\\CAFTA32\\06R0B\\PSL1\\Cuts-B\\PSL1Top10K.CUT  
                          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
Cutset Prob  
      Not sorted
Event Prob  
      Printed in full
Event
                                                                                                              Appendix 1
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  


RECOVERY DEVELOPMENT
Two perspectives will be applied to the recovery development, since the time variable could be applied differently. The more
Appendix 2
liberal of the two calculations will be applied in the quantification.
RECOVERY DEVELOPMENT  
            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
Two perspectives will be applied to the recovery development, since the time variable could be applied differently. The more  
            STRESS                   2.0E+00     Unusual condition
liberal of the two calculations will be applied in the quantification.  
            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
DIAGNOSIS  
            DIAGNOSITIC               2.0E-01
            TOTAL
            ACTION
            BASE                       1.0E-03
            TIME                     1.0E+00     although limited information available time penalty applied to diagnosis
Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging  
            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
BASE  
            PROCEDURES                1.0E+00     Nominal
1.0E-02  
            ERGONOMICS               1.0E+00     Nominal
TIME  
            FIT FOR DUTY             1.0E+00     Nominal
1.0E+01  
            WORK PROCESS             1.0E+00     Nominal
limited information available as to how much time was left prior to sys failure  
            ACTION TOTAL               1.0E-02
STRESS  
            TOTAL                     2.1E-01
2.0E+00  
                                                                                                              Appendix 2
Unusual condition  
COMPLEXITY  
1.0E+00  
Nominal  
EXPERIENCE/TRAI
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
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  


DIAGNOSIS
Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging
BASE                     1.0E-02
Appendix 2
TIME                   1.0E+00     limited information available as to how much time was left prior to sys failure
DIAGNOSIS  
STRESS                 2.0E+00     Unusual condition
COMPLEXITY             1.0E+00     Nominal
EXPERIENCE/TRAI         1.0E+00     Nominal
N
PROCEDURES              1.0E+00     Nominal
Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging  
ERGONOMICS             1.0E+00     Nominal
FIT FOR DUTY           1.0E+00     Nominal
WORK PROCESS           1.0E+00     Nominal
DIAGNOSITIC             2.0E-02
BASE  
TOTAL
1.0E-02  
ACTION
TIME  
BASE                     1.0E-03
1.0E+00  
TIME                   1.0E+01     apply time penalty that after diagnosis, time available = time req'd
limited information available as to how much time was left prior to sys failure  
STRESS                 2.0E+00
STRESS  
COMPLEXITY             5.0E+00     numerous actions with multiple sub-tasks outside Main Control Room
2.0E+00  
EXPERIENCE/TRAI         1.0E+00     Nominal
Unusual condition  
N
COMPLEXITY  
PROCEDURES              1.0E+00     Nominal
1.0E+00  
ERGONOMICS             1.0E+00     Nominal
Nominal  
FIT FOR DUTY           1.0E+00     Nominal
EXPERIENCE/TRAI
WORK PROCESS           1.0E+00     Nominal
N
ACTION TOTAL             1.0E-01
1.0E+00  
TOTAL                   1.2E-01
Nominal  
                                                                                                Appendix 2
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
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
}}
}}

Latest revision as of 07:07, 14 January 2025

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)

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)

xx PUBLICLY AVAILABLE

G NON-PUBLICLY AVAILABLE

G SENSITIVE

xx NON-SENSITIVE

ADAMS: G Yes

ACCESSION NUMBER:_________________________

xxG SUNSI REVIEW COMPLETE

OFFICE

RII:DRS

RII:DRS

RII:DRS

RII:DRP

CONTRACTOR

CONTRACTOR

RII:DRP

SIGNATURE

RA

RA

RA

RA

RA

RA

RA

NAME

SROSE

RMOORE

JHAMMAN

RTAYLOR

MSHYLAMBERG NDELIAGRECA MSYKES

DATE

11/30/2009

11/19/2009

1/12/2010

11/20/2009

11/18/2009

11/5/2009

1/13/2010

E-MAIL COPY?

YES

NO YES

NO YES

NO YES

NO YES

NO YES

NO YES

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

OFFICIAL

RECORD

COPY

DOCUMENT

NAME:

S:\\DRS\\ENG

BRANCH

1\\BRANCH

INSPECTION

FILES\\CDBI

INSPECTIONS\\CDBI INSPECTIONS\\INSP REPORTS\\CDBI FINAL INSPECTION REPORTS\\REV 1 ST LUCIE 2009006 CDBI

REPORT (SDR).DOC

FP&L

4

cc w/encl:

Richard L. Anderson

Site Vice President

St. Lucie Nuclear Plant

Electronic Mail Distribution

Robert J. Hughes

Plant General Manager

St. Lucie Nuclear Plant

Electronic Mail Distribution

Mark Hicks

Operations Manager

St. Lucie Nuclear Plant

Electronic Mail Distribution

Rajiv S. Kundalkar

Vice President - Fleet Organizational

Support

Florida Power & Light Company

Electronic Mail Distribution

Eric Katzman

Licensing Manager

St. Lucie Nuclear Plant

Electronic Mail Distribution

Abdy Khanpour

Vice President

Engineering Support

Florida Power and Light Company

P.O. Box 14000

Juno Beach, FL 33408-0420

McHenry Cornell

Director

Licensing and Performance Improvement

Florida Power & Light Company

Electronic Mail Distribution

Alison Brown

Nuclear Licensing

Florida Power & Light Company

Electronic Mail Distribution

Faye Outlaw

County Administrator

St. Lucie County

Electronic Mail Distribution

Mitch S. Ross

Vice President and Associate General

Counsel

Florida Power & Light Company

Electronic Mail Distribution

Marjan Mashhadi

Senior Attorney

Florida Power & Light Company

Electronic Mail Distribution

William A. Passetti

Chief

Florida Bureau of Radiation Control

Department of Health

Electronic Mail Distribution

Ruben D. Almaguer

Director

Division of Emergency Preparedness

Department of Community Affairs

Electronic Mail Distribution

J. Kammel

Radiological Emergency Planning

Administrator

Department of Public Safety

Electronic Mail Distribution

Mano Nazar

Executive Vice President and Chief Nuclear

Officer

Florida Power & Light Company

Electronic Mail Distribution

(Vacant)

Vice President

Nuclear Plant Support

Florida Power & Light Company

Electronic Mail Distribution

Jack Southard

Director

Public Safety Department

St. Lucie County

Electronic Mail Distribution

FP&L

5

Letter to Mano Nazar from Kriss Kennedy dated January 19, 2010.

SUBJECT:

ST. LUCIE NUCLEAR PLANT - NRC COMPONENT DESIGN BASES

INSPECTION - INSPECTION REPORT 05000335/2009006 AND

05000389/2009006; PRELIMINARY GREATER THAN GREEN FINDINGS

Distribution w/encl:

C. Evans, RII

L. Slack, RII

OE Mail

RIDSNRRDIRS

PUBLIC

RidsNrrPMStLucie Resource

Enclosure

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket Nos.: 50-335, 50-389

License Nos.: DPR-67 and NPF-16

Report Nos.: 05000335/2009006, 05000389/2009006

Licensee:

Florida Power & Light Company (FP&L)

Facility:

St. Lucie Nuclear Plant, Units 1 & 2

Location:

Jensen Beach, FL 34957

Dates:

August 3-14 (Weeks 1 & 2)

August 31-September 4 (Week 3)

Inspectors:

S. Rose, Senior Operations Inspector (Lead)

R. Moore, Senior Reactor Inspector

J. Hamman, Reactor Inspector

R. Taylor, Senior Reactor Inspector

M. Shylamberg, Contractor

N. Della Greca, Contractor

Approved by: Binoy Desai, Chief

Engineering Branch 1

Division of Reactor Safety

Enclosure

SUMMARY OF FINDINGS

IR 05000335/2009006, 05000389/2009006; 8/3/2009 - 9/4/2009; St. Lucie Nuclear

Plant, Units 1 and 2; NRC Component Design Bases Inspection.

This inspection was conducted by a team of four NRC inspectors from the Region II

office, and two NRC contract inspectors. Two findings of very low significance (Green)

were identified during this inspection and were classified as non-cited violations. Also,

two apparent violations (AV) with potential safety significance greater than Green were

identified. The significance of most findings is indicated by their color (Green, White,

Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for

which the SDP does not apply may be Green or be assigned a severity level after NRC

management review. The NRC's program for overseeing the safe operation of

commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight

Process, (ROP) Revision 4, dated December 2006.

Cornerstone: Mitigating Systems

Green. The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion

III, Design Control, for failure to translate the design basis as specified in the license

application into specifications, drawings, procedures, and instructions. The licensee

did not ensure that the component cooling water (CCW) surge tank design included

adequate overpressure protection for all procedurally allowed configurations as

required by the applicable ASME Boiler and Pressure Vessel Code,Section VIII,

Division 1. The code requires that no intervening stop valves be between the vessel

and its protective device or devices or between the protective devices and the point

of discharge. The team concluded that stop valve V6466 was an intervening stop

valve for the CCW surge tank vent path to the chemical drain tank (CDT). The issue

was entered in the 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]

3

Enclosure

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]

TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion III, Design

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

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]

TBD. The team identified an AV of 10 CFR 50, Appendix B, Criterion XVI, Corrective

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

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

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

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

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

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

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.

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)).

10

Enclosure

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

11

Enclosure

.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.

12

Enclosure

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

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

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.

14

Enclosure

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.

15

Enclosure

.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

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Enclosure

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.

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Enclosure

.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,

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 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

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

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.

20

Enclosure

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

21

Enclosure

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

22

Enclosure

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

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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.

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.

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.

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.

25

Enclosure

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

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.

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.

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

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

Attachment

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

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

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

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

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

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

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

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

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

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

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

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

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

9

Attachment

Modifications

Change Request Notice CRN 03167-13230, Vendor Manual for Shutdown Cooling Heat

Exchanger Update to Show the Correct Tube Plugs, Rev. 0

Change Request Notice CRN 18362, Install Temporary Protection on LG-14-2A and LG-14-2B,

Rev. 0

Change Request Notice CRN 00048-9446, Permanent Removal of Gravity Damper Cover

Plates on GD-1 and GD-2, Rev. 0

Miscellaneous Documents

DBD-CCW-1, Component Cooling Water System, Rev. 2

DBD-ICW-1, Intake Cooling Water System, Rev. 2

DBD-HVAC-1, Safety Related HVAC Systems, Rev. 2

DBD-120V-AC-1, Class 1E 120 V AC Power System, Rev 2

DBD-480V-AC-1, 480 VAC Distribution System, Rev 2A

DBD-4160 VAC-1, 4160 VAC Distribution System, Rev 2

DBD-EDG-1, Emergency Diesel Generatoor System, Rev 3

DBD-ESF-1, Engineered Safety Features Actuation System, Rev 2

DBD-HVAC-1, Safety Related HVAC Systems Design Basis Document, Rev 2

DBD-PZR-1, Pressurizer System, Rev 2

DBD-VDC-1, Class 1E DC Electrical Distribution System, Rev 2

8770-5756, Component Cooling Water Pump, Rev. 6

8770-7248, I/M Centrifugal Fans HVS-4A, 4B, 5A, 5B, HVE-1, 2, 4, 5, 7A, 7B, 8A, 8B, 9A, 9B,

10A, 10B, 15, 16A, 16B, 21A, 21B, 13A, 13B, 14 & 33, Rev. 5

0711209, Component Cooling Water System, Rev. 12

0702209, Component Cooling Water System, Rev. 8

Westinghouse, 10CFR21, Component Cooling Water - Overpressure Transient, Dated July 25,

1984

EPO-84-1662, CCW Surge Tank Overpressurization, Dated, August 20, 1984

SLN-88-021-10-20, JPN-PSL-SEICP-92-28, Evaluation of the Design basis for Fisher & Porter

Indicating Controllers for Temperature Control Valves TCV-14-4A and TCV-14-4B.

JPN-PSL-SENP-93-001, Inputs for the LOCA Containment Re-Analysis, Rev. 0

NRC NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, April 1995

NRC Information Notice 2008-02: Findings Identified During Component Design Bases

Inspections, March 19, 2008

FPL-09-366, Westinghouse Letter to Phil Barnes, CCW Flowrates Used in Containment

Analysis for St. Lucie, September 2, 2009

JPN-PSL-SEMP-91-007, Safety Evaluation for RAB Electrical Equipment and Battery Room

HVAC, St. Lucie Unit 1, Rev 1

0711401, Engineered Safety Features Actuation System, Rev 1

00809-0100-4388, Rosemount 1153 Series D Alphaline Nuclear Pressure Transmitter, Rev BA

2998-15662, Instruction Manual for Safety Features Actuation System (SFAS) Vol. 1 & Vol. 2,

Manual No. TM9N38, Rev 8

8770-7423, Instruction Manual Turbine Building Battery Charger, C&D Battery, Manual No.

MCB-2010, Rev 5

8770-10459, Instruction Manual for Battery Chargers 1AA, 1BB, and 1D, C&D Battery, Manual

No. RS-421, Rev 5

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

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

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

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.

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.

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.

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

operators failing to trip the RCP leading directly to a large seal LOCA and core damage. The

sequence is assigned a nominal value of 6 - four for the initiating event frequency and two for

the operator error. Phase III results in a lower probability of operators tripping the RCP of 3E-3.

Therefore, the color is the same in both phases but, numerically a magnitude higher in the

Phase II result. This shows reconciliation between the two phases.

CONCLUSIONS/RECOMMENDATIONS - Given the present information associated with the air

intrusion of October 2008, it is reasonable to assume the initiating event frequency increased by

at least a magnitude. Such a shift with recovery is in the White zone of safety characterization.

Assuming that CCW was in imminent failure the safety characterization shifts into the red zone,

even with recovery. Therefore, this performance deficiency should be preliminarily

characterized as >Green with the intent to acquire as much information about air intrusion into

the CCW system as:

an initiator for Loss of the CCW system

an undetected failure mechanism of any CCW functions while the equipment is in

standby

APPENDICES: 1. Full Scope Model Output

2. Recovery Development

Analyst: W. Rogers

Date: 10/30/09

Reviewed By: G. MacDonald Date: 11/02/2009

Appendix 1

EDITED FULL SCOPE MODEL CUTSETS FOR TOTAL LOSS OF COMPONENT COOLING WATER SYSTEM

TOP 10,000 Cutsets for PSL1

C:\\CAFTA32\\06R0B\\PSL1\\Cuts-B\\PSL1Top10K.CUT

Cutset Prob

Event Prob

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 2

RECOVERY DEVELOPMENT

Two perspectives will be applied to the recovery development, since the time variable could be applied differently. The more

liberal of the two calculations will be applied in the quantification.

DIAGNOSIS

Operators recognize air intrusion via surge tank annunicators and pump ampmeter indicators swinging

BASE

1.0E-02

TIME

1.0E+01

limited information available as to how much time was left prior to sys failure

STRESS

2.0E+00

Unusual condition

COMPLEXITY

1.0E+00

Nominal

EXPERIENCE/TRAI

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

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

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

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

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

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