ML100210081
| ML100210081 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| 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) | |
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
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.
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.
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)
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.:
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
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
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
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
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
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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
16
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.
17
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
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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
23
Enclosure
compressors run at approximately 160 psig system pressure and the CCW system
pressure is approximately 120 psig. The common cause failure vulnerability of the CCW
system from a failure in the waste compressor units was applicable to both Unit 1 and
Unit 2.
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
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
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
Adequacy of Performance Monitoring of the IA
Compressor Emergency Cooling System.
(1R21.2.3)
05000335, 389/2009006-04
Inadequate Procedure for Restoration of Non-
Essential CCW Flow Following a SIAS
(1R21.3)
05000335, 389/2009006-05
Failure to Translate Design Basis
Specifications to Prevent Single Failure of
CCW (4OA5)
05000335,389/2009006-06
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
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
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
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
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
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
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