ML033020375
| ML033020375 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/27/2003 |
| From: | Joel Munday NRC/RGN-II/DRP/RPB3 |
| To: | Stall J Florida Power & Light Co |
| References | |
| IR-03-004 | |
| Download: ML033020375 (24) | |
See also: IR 05000250/2003004
Text
October 27, 2003
Florida Power & Light Company
ATTN: Mr. J. A. Stall
Senior Vice President of Nuclear Operations
PO Box 14000
Juno Beach, FL 33408-0420
SUBJECT:
TURKEY POINT NUCLEAR PLANT - INTEGRATED INSPECTION REPORT
05000250/2003004 AND 05000251/2003004
Dear Mr. Stall:
On September 27, 2003, the US Nuclear Regulatory Commission (NRC) completed an inspection
at your Turkey Point Units 3 and 4. The enclosed integrated inspection report documents the
inspection findings which were discussed on October 3, 2003, with Mr. T. Jones and other
members of your staff.
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 inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, there was one self-revealing finding of very low safety
significance (Green). The finding was determined to involve a violation of NRC requirements.
However, because of the very low safety significance and because the violation was entered into
your corrective action program, the NRC is treating this violation as non-cited violation (NCV)
consistent with Section VI.A of the NRC Enforcement Policy. Additionally, a licensee-identified
violation which was determined to be of very low safety significance is listed in Section 4OA7 of
this report. If you contest the non-cited violation in this report, 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 Senior
Resident Inspector at the Turkey Point facility.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document Room
or from the Publicly Available Records (PARS) component of the NRCs document system
2
(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/
Joel T. Munday, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Docket Nos. 50-250, 50-251
Enclosure: Inspection Report 05000250/2003004 and 05000251/2003004
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
3
cc w/encl:
T. O. Jones
Site Vice President
Turkey Point Nuclear Plant
Florida Power and Light Company
Electronic Mail Distribution
Walter Parker
Licensing Manager
Turkey Point Nuclear Plant
Florida Power and Light Company
Electronic Mail Distribution
Michael O. Pearce
Plant General Manager
Turkey Point Nuclear Plant
Florida Power and Light Company
Electronic Mail Distribution
Don Mothena, Manager
Nuclear Plant Support Services
Florida Power & Light Company
Electronic Mail Distribution
Rajiv S. Kundalkar
Vice President - Nuclear Engineering
Florida Power & Light Company
Electronic Mail Distribution
M. S. Ross, Attorney
Florida Power & Light Company
Electronic Mail Distribution
Linda Tudor
Document Control Supervisor
Florida Power & Light Company
Electronic Mail Distribution
Attorney General
Department of Legal Affairs
The Capitol
Tallahassee, FL 32304
William A. Passetti
Bureau of Radiation Control
Department of Health
Electronic Mail Distribution
County Manager
Metropolitan Dade County
Electronic Mail Distribution
Craig Fugate, Director
Division of Emergency Preparedness
Department of Community Affairs
Electronic Mail Distribution
Curtis Ivy
City Manager of Homestead
Electronic Mail Distribution
Distribution w/encl: (See page 4)
4
Distribution w/encl:
E. Brown, NRR
RIDSNRRDIPMLIPB
PUBLIC
OFFICE
DRP/RII
DRP/RII
DRP/RII
DRP/RII
SIGNATURE
sn
kc (for)
kg
jh
NAME
SNinh:vyg
CPatterson
KGreenBates
JHanna
DATE
10/22/2003
10/24/2003
1024/2003
10/24/2003
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
PUBLIC DOCUMENT
YES
NO
OFFICIAL RECORD COPY DOCUMENT NAME: C:\\ORPCheckout\\FileNET\\ML033020375.wpd
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-250, 50-251
License Nos:
Report Nos:
05000250/2003004 and 05000251/2003004
Licensee:
Florida Power & Light Company
Facility:
Turkey Point Nuclear Plant, Units 3 & 4
Location:
9760 S. W. 344th Street
Florida City, FL 33035
Dates:
June 29, 2003 - September 27, 2003
Inspectors:
C. Patterson, Senior Resident Inspector
J. Hanna, Acting Senior Resident Inspector
K. Green-Bates, Resident Inspector
S. Ninh, Senior Project Engineer
Approved by:
Joel T. Munday, Chief
Reactor Projects Branch 3
Division of Reactor Projects
SUMMARY OF FINDINGS
IR 05000250/2003-004, 05000251/2003-004; 06/29/2003 - 09/27/20003; Turkey Point Nuclear
Power Plant, Unit 3 & 4; Event Followup.
The report covered a three month period of inspection by resident inspectors. One Green non-
cited violation was identified. The significance of the finding is indicated by its 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 NRCs program for overseeing the safe operation of commercial
nuclear power reactors is described in NUREG-1649, Reactor Overnight Process, Revision 3,
dated July 2000.
A.
Inspector Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. A self-revealing finding was identified concerning a failure to comply with
10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings.
Licensee drawings and instructions used to research Clearance Zone 28-01 relay
tagouts were not sufficient to assure that the design basis Engineering Safety
Feature Actuation Signal (ESFAS) function of these components was protected.
As a result, a plant configuration was established which rendered the automatic
start of all AFW pumps on a Low-Low Steam Generator Level signal unavailable
while U3 was in Operational Mode 3.
This finding is greater than minor since it affected the Mitigating System
Cornerstone objective for Equipment Availability and had an actual safety impact
of rendering the automatic start of all AFW pumps on a Low-Low Steam
Generator Level signal unavailable while in Operational Mode 3. This finding was
reviewed using the Significance Determination Process and was determined to be
of very low safety significance because for the two applicable design basis
accidents requiring this signal, alternative methods would have started the AFW
pumps and the system would have been able to perform its safety function.
(Section 4OA3.1)
B.
Licensee Identified Violations
A violation of very low safety significance, which was identified by the licensee, has been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the corrective action program. The violation and corrective action
tracking number are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status:
Unit 3 operated at full power during most of the inspection period. On August 15, 2003, Unit 3
reduced power to 20% due to temperature control problems associated with the cooling of the
main turbine generator exciter. The two turbine plant cooling water heat exchangers were
cleaned and the unit returned to full power on August 17, 2003.
Unit 4 operated at full power during most of the inspection period. On August 1, 2003, Unit 4
reduced power to 30% due to temperature control problems associated with cooling of the main
turbine generator exciter. The two turbine plant cooling water heat exchangers were cleaned
and the unit returned to full power on August 4, 2003. The unit started power coastdown on
September 21, 2003, in preparation for a refueling outage and was at 94% at the close of the
inspection period.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity (Reactor-R),
1R01
Adverse Weather Protection
a.
Inspection Scope
In September, due to the proximity of Hurricane Isabel to the site, the inspectors
performed a walkdown of the site and reviewed the licensees preparations for hurricane
high winds/rain and implementation of 0-OSP-102.1 Flood Protection Stoplog Inspection;
EPIP-20106 Natural Emergencies; and 0-ONOP-103.3 Severe Weather Preparations, to
verify that those preparations limited the risk of weather related initiating events, ensured
accessibility to accident mitigation system equipment, and adequately protected accident
mitigation systems from adverse weather effects. The inspectors also reviewed the
condition of selected flood mitigation structures and components and verified that
corrective actions were taken at the appropriate thresholds within a time schedule which
met the local onset of hurricane season. Where licensee identified deficiencies were
observed, the inspectors verified that the deficiencies were properly entered into the
corrective action program and timely resolution was being pursued.
b.
Findings
No findings of significance were identified.
1R04
Equipment Alignment
a.
Inspection Scope
Partial Equipment Walkdowns
The inspectors conducted four partial alignment verifications of the safety related
systems listed below during the inspection period to review the operability of required
2
redundant trains or backup systems while the other trains were inoperable or out of
service. These inspections included reviews of plant lineup procedures, operating
procedures, and piping and instrumentation drawings which were compared with
observed equipment configurations to identify any discrepancies that could affect
operability of the redundant train or backup system. The inspectors reviewed the
following systems:
3B Intake Cooling Water (ICW) header while the 3A ICW header was out of
service for cleaning the 3A ICW/(Component Cooling Water) CCW basket
strainer
Unit 4 (Auxiliary Feedwater) AFW Train 1 while AFW Train 2 was out of service
for operability testing
Unit 4A, 3A and 3B (Emergency Diesel Generators) EDGs while the Unit 4B EDG
was declared inoperable due to testing
Unit 4 AFW Pump A Train 1 while Train 2 AFW Pumps B and C were out of
service
b.
Findings
No findings of significance were identified.
1R05
Fire Protection
a.
Inspection Scope
The inspectors toured the following eight plant areas to evaluate conditions related to
control of transient combustibles and ignition sources, the material condition and
operational status of fire protection systems, and selected fire barriers used to prevent
fire damage or fire propagation. The inspectors reviewed these activities against
provisions in the licensees Off Normal Operating Procedure, 0-ONOP-016.8, Response
to a Fire/Smoke Detection System Alarm, 0-SME-091.1, Fire and Smoke Detection
System Annual Test, 0-ADM-016, Fire Protection Plan, and 10 CFR Part 50, Appendix R.
The following areas were inspected:
Unit 4B 4160 Switchgear Room (Fire Zone 67)
Unit 4A 4160 Switchgear Room (Fire Zone 68)
Unit 4 Auxiliary Transformer Area (Fire Zone 82)
Unit 3 Auxiliary Transformer Area (Fire Zone 87)
Unit 3 and Unit 4 Auxiliary Feedwater Pump Area (Fire Zone 84)
Unit 3 and Unit 4 Pipe and Valve Room Fire Zone 35)
Unit 3 and Unit 4 Auxiliary Building Hallway (Fire Zone 58)
Unit 3 and Unit 4 AFW Pump Area (Fire Zone 84)
b.
Findings
No findings of significance were identified.
3
1R06
Flood Protection Measures
a.
Inspection Scope
The inspectors reviewed Turkey Point Final Safety Analysis Report Sections 1.6 and 1.3,
as well as the procedures and other flood mitigation documents listed below, which
depicted design flood levels and protection for areas containing risk and safety-related
equipment to determine consistency with design requirements and identify areas that
may be affected by internal flooding.
Drawing No. JPN-PTN-SECJ-90-057; Drains Subject to BackFlow Inside Flood
Protection Barrier Network
Bechtel Power Corporation No. SFB-3274; Turkey Point Units 3 & 4 Engineering
Guideline for Internal Flood Protection.
Turkey Point Units 3 & 4 Design No. 5610-000-DB-001 Sect VIII; Internal
Flooding Criteria
Turkey Point Units 3 & 4 Design No. 5610-000-DB-001 Sect IX; External Flooding
Criteria
Procedure No. EPIP-20106, Natural Emergencies
Procedure No. 0-OSP-102.1, Flood Protection Stoplog Inspection
A general site walkdown was conducted, with a specific walkdown of the risk significant
Unit 3 and Unit 4 Load Center rooms to ensure that flood protection measures were in
accordance with design specifications. Specific attributes that were checked included
structural integrity, flood platform heights for safety equipment, the sealing of the
switchgear room penetrations, and unobstructed floor drains. Equipment used for flood
mitigation, such as switchgear room sump pumps, sump system level alarms and
external drains, were reviewed for operability and/or structural integrity. Potential
flooding sources were examined to verify proper maintenance.
A review of outstanding maintenance work orders and related condition reports was
performed to verify that deficiencies did not significantly affect the Unit 3 and Unit 4 load
center room flood mitigating functions. The inspectors discussed with engineering and
maintenance management equipment issues to verify that identified problems were being
appropriately resolved in a timely fashion.
b.
Findings
No findings of significance were identified.
1R07
Heat Sink Performance
a.
Inspection Scope
The CCW system at Turkey Point is a safety-related high risk significant system. The
inspectors reviewed the Unit 3 and Unit 4 CCW heat exchanger (HX) thermal
performance testing results that were conducted in the month of July 2003, to verify that
the CCW HXs were capable of removing the basis accident heat load as required. The
inspectors also reviewed Technical Specification (TS) 3/4.7.2, Final Safety Analysis
4
Report (FSAR) Section 9, PTN-BFJI-95-003, Effect of Instrumentation Uncertainty on
Allowable ICW Temperature Calculation, Revision 1, and Calculation No.
PTN-BFJM-96-004, Revised CCW HX Operability Curves for Thermal Uprate, to ensure
that test acceptance criteria, number of plugged tubes, instrument errors, and frequency
of surveillance or testing were appropriately accounted for and included in the licensees
procedures 3/4 OSP-030,4, Component Cooling Water Heat Exchanger Performance
Test and 3/4 OSP-019.4, Component Cooling Water Heat Exchanger Performance
Monitoring.
b.
Findings
No findings of significance were identified.
1R11
Licensed Operator Requalification
a.
Inspection Scope
On July 23, 2003, the inspectors observed and assessed licensed operator actions on a
simulator scenario for a main steam isolation valve failing closed and a steam leak inside
containment that also involved the failure of critical safety equipment. The licensee used
Simulator Practice Scenario 008 Team Training, Attachments 5, 11 and 58. The
inspectors specifically evaluated the following attributes related to operating crew
performance:
Clarity and formality of communication
Ability to take timely action to safely control the unit
Prioritization, interpretation, and verification of alarms
Correct use and implementation of Emergency Operating Procedures and
Emergency Plan Implementing Procedures
Control board operation and manipulation, including high-risk operator actions
Oversight and direction provided by Operations supervision, including ability to
identify and implement appropriate TS actions, regulatory reporting requirements,
and emergency plan actions and notifications
Effectiveness of the post training critique
b.
Findings
No findings of significance were identified.
1R12
Maintenance Effectiveness
a.
Inspection Scope
The inspectors reviewed the following two equipment problems and associated Condition
Reports (CRs) to verify the licensees maintenance efforts met the requirements of 10 CFR 50.65 (Requirements for Monitoring the Effectiveness of Maintenance at Nuclear
Power Plants) and Administrative Procedure ADM-728. The inspectors efforts focused
on maintenance rule scoping, characterization of the failed components, risk significance,
determination of a(1) classification, corrective actions, and the appropriateness of
5
established performance goals and monitoring criteria. The inspectors also attended
applicable expert panel meetings, interviewed responsible engineers, and observed
some of the corrective maintenance activities. Furthermore, the inspectors verified
whether equipment problems were being identified at the appropriate level and entered
into the corrective action program.
CR 03-1525
Unit 3 Steam Jet Air Ejector Found Flooded With Water
CR 03-1560
Inverter & Battery Room Air Conditioning Unit Grounded
b.
Findings
No findings of significance were identified.
1R13
Maintenance Risk Assessments and Emergent Work Control
a.
Inspection Scope
The inspectors reviewed the following seven emergent items, as described in the
referenced CRs or work orders (WOs). The inspectors verified that the emergent work
activities were adequately planned and controlled, as described in 0-ADM-068, Work
Week Management and O-ADM-225, On Line Risk Assessment and Management. The
inspectors verified that, as appropriate, contingencies were in place to reduce risk,
minimize time spent in increased risk configurations, and avoid initiating events. The
following items were reviewed:
CR 03-1528
Gauge Calibration
CR 03-1624
Boric Acid on High Head Safety Injection
CR 03-1647
3A Load Center Undervoltage Relay Failure
CR 03-2249
Demineralized Water Storage Tank - Water on Top of
Bladder
CR 03-1563
3C CCW Inservice Test (IST) - CCW Not Receiving a Valid
Auto Start Signal
CR 03-1331
Unit-3 Backup Pressurizer Heaters Continuously Energized
CR 03-2174
AFW Lube Oil Pump Footvalve Failure
b.
Findings
No findings of significance were identified.
1R14
Personnel Performance During Non-routine Plant Evolutions and Events
a.
Inspection Scope
For the non-routine events described below, the inspectors reviewed operator logs, plant
computer data, and strip charts to determine what occurred and how the operators
responded, and to verify that the response was in accordance with plant procedures:
6
On July 14, 2003, Unit 3 3A load center under voltage relay 327H/3A2 did not
reset during the performance of surveillance test 3-OSP-006.2, U3 480 Volt
Switchgear Undervoltage Test. Unit 3 made a brief planned entry into TS 3.0.3
for the troubleshooting of the load center under voltage relay 327H/3A2. The
licensee determined that the 3A2 switch did not work properly and needed to be
replaced. The switch remained jumpered while waiting for the replacement. The
switch was repaired on September 17, 2003. (CR 03-1647 & CR 03-1648)
On July 14, 2003, during the performance of a Unit 3 3B Charging pump
Equipment Clearance Order 3-03-07-027 of the system train, steps were
performed out of sequence. (CR 03-1705, 03-1708, 03-1710, 03-2021)
b.
Findings
No findings of significance were identified.
1R15
Operability Evaluations
a.
Inspection Scope
The inspectors reviewed the following six operability determinations to ensure that TS
operability was properly supported and the system, structure or component remained
available to perform its safety function with no unrecognized increase in risk. The
inspectors reviewed the UFSAR, applicable supporting documents and procedures, and
interviewed plant personnel to assess the adequacy of the interim CR disposition.
CR 03-1441
Turbine Plant Cooling Water Isolation Valve
CR 03-1597
4A Emergency Containment Cooler
PTN-ENG-SENS-03-009
Control Room Habiltability
CR 03-1847
Failure of Control Room Damper, D-2
CR 03-0895, Sup. 1
4A EDG Fuse Clips
CR 03-2306
3A EDG Wrong Oil Added
b.
Findings
No findings of significance were identified.
1R19
Post Maintenance Testing
a.
Inspection Scope
For the following four post maintenance tests listed below, the inspectors reviewed the
test procedures and either witnessed the testing and/or reviewed test records to
determine whether the scope of testing adequately verified that the work performed was
correctly completed and demonstrated that the affected equipment was functional and
operable. The inspectors verified that the requirements of procedure 0-ADM-737, Post
Maintenance Testing, were incorporated into test requirements. The inspectors reviewed
the following list of tests:
7
Fire Pump Casing Vent
4-OSP-023.1
Diesel Generator Operability Test
WO 33015160-8
WO 33014567-5
b.
Findings
No findings of significance were identified.
1R20
Refueling and Outage Activities
a.
Inspection Scope
The inspectors reviewed the outage plans and contingency plans for the Unit 4 refueling
outage, scheduled for October 6 - 26, 2003, to confirm that the licensee had
appropriately considered risk, industry experience, and previous site-specific problems in
developing and implementing a plan that assured maintenance of defense-in-depth. The
inspectors also observed portions of the planned outage activities listed below.
Outage Risk
Prior to the start of the refueling outage the inspectors reviewed the outage risk
assessment with the licensee. The outage risk status or color and plant evolutions
during the outage were reviewed. The risk assessment was planned according to plant
procedure, O-ADM-051, Outage Risk Management. The inspectors reviewed that the
outage unit risk as described in the plan was consistent with the outage work orders on
file.
Refueling Activities
The inspectors observed new fuel pool load activities in the control room and spent fuel
pool areas. Core load activities were observed and activities verified in accordance with
procedure 4-OP-038.5, Refueling Pre-Shuffle in Spent Fuel Pit.
In addition, the nuclear fuel supplier informed the licensee of an issue involving a
nonconformance associated with the top and bottom nozzles of some fuel assemblies.
The inspectors observed the licensee inspect the nozzles on the new fuel assemblies
and verified that CRs were generated as appropriate.
b.
Findings
No findings of significance were identified.
1R22
Surveillance Testing
a.
Inspection Scope
The inspectors reviewed the following five surveillance tests to verify that the tests met
the TS, the Updated Final Safety Analysis Report (UFSAR), and licensee procedure
8
requirements and demonstrated the systems were capable of performing their intended
safety functions and their operational readiness.
3-OSP-030.4, Component Cooling Water Heat Exchanger Performance Test
4-OSP-030.4, Component Cooling Water Heat Exchanger Performance Test
3-OSP- 023.2, 3A EDG 24 Hour Full Load Test and Load Rejection
3-OSP-030.1, 3C Component Cooling Water IST
4-OSP-075.2, B AFW Operability Test
b.
Findings
No findings of significance were identified.
1R23
Temporary Plant Modifications
a.
Inspection Scope
The inspectors reviewed the following two active temporary modifications to verify that
risk significant items did not adversely affect the operation of a system that was altered.
The inspectors reviewed plant procedure 0-ADM-503, Control and Use of Temporary
System Alterations (TSA), to verify that the modifications were controlled as required by
procedure. In addition, the inspectors toured plant areas and specifically looked for any
temporary modifications that might not be identified to ensure that all issues were
recognized. The following active temporary modifications were reviewed:
U4 AFW Nitrogen Backup Cage - Trip Sensitive Equipment Protection
Modification
TSA 3-03-006-022 Troubleshooting Modifications for 3A 480 Volt Undervoltage
Circuit
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness (EP)
1EP6
Drill Evaluation
a.
Inspection Scope
On August 19, 2003, the inspectors observed an operating crew in the simulator during
the 3rd quarter EP drill of the site emergency response organization. During this drill the
inspectors assessed operator actions in the control room simulator to verify whether
emergency classification, notification, and protective action recommendations were made
in accordance with implementing procedures. Additionally, the inspectors evaluated the
adequacy of the post drill critiques conducted in the simulator.
9
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
a.
Inspection Scope
The inspectors sampled licensee submittals for the two performance indicators (PIs)
listed below for the period from third quarter 2002 through second quarter 2003. To
verify the accuracy of the PI data reported during that period, PI definitions and guidance
contained in NEI 99-02, Regulatory Assessment Indicator Guideline, Rev. 2, were used
to verify the basis in reporting for each data element.
Reactor Safety Cornerstone
Safety System Unavailability - High Pressure Injection System
Safety System Unavailability - Emergency AC Power System
The inspector reviewed a selection of licensee event reports (LERs), portions of Unit 3
and Unit 4 operator log entries, daily morning reports (including the daily CR
descriptions), the monthly operating reports, and PI data sheets to verify that the licensee
had adequately identified the number of unavailable hours that occurred during the
previous four quarters. These unavailable hours were compared to the number reported
for the PI during the current quarter. In addition, the inspectors also interviewed licensee
personnel associated with the PI data collection, evaluation, and distribution.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
a. Inspection Scope
The inspectors selected the following CRs for detailed review and discussion with the
licensee. These CRs were examined to verify whether problem identification was timely,
complete and accurate; safety concerns were properly classified and prioritized for
resolution; technical issues were evaluated and dispositioned to address operability and
reportability; root cause or apparent cause determinations were sufficiently thorough;
extent of condition, generic implications, common causes, and previous history were
adequately considered; and appropriate corrective actions (short and long-term) were
implemented or planned in a manner consistent with safety and TS compliance. The
inspectors evaluated the CRs against the requirements of the licensees corrective action
program as delineated in Administrative Procedures ADM-518, Condition Reports, ADM-
059, Root Cause Analysis, and 10 CFR 50, Appendix B.
10
CR 03-1331
U3 Pressurizer Back-up Heaters and Millstone OE
CR 03-1847
Control Room Emergency Intake Duct Flow Balancing
b.
Findings
No findings of significance were identified.
4OA3 Event Follow-up
.1
(Closed) Licensee Event Report (LER) 05000250/2003-005-00, Disabling Both Auxiliary
Feedwater Trains Inadvertently During Mode 3
Introduction: A self revealing Green NCV was identified for failure to comply with 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings. Licensee
drawings and documents used to develop Clearance Zone 28-01 relay tagouts were not
sufficient to assure that the design basis ESFAS function of these components was
protected. As a result, a plant configuration was established which rendered the
automatic start of all AFW pumps on a Low-Low Steam Generator Level signal
unavailable while Unit 3 was in Operational Mode 3.
Description: On March 1, 2003, Unit 3 was in Operational Mode 4 (Hot Standby) during
a planned shutdown for Refueling Outage (RFO) Cycle 20. During the shutdown, when
steam generators were drained to below the 10% AFW initiation setpoint, the AFW
Steam Supply MOVs failed to open as required. The licensees analyses concluded that
the cause of the failure was the reactor protection and engineered safety features
actuation system (ESFAS) relays for both trains of AFW had been inadvertently disabled.
Equipment tagout of Clearance Zone 28-01 performed earlier in the day to isolate
equipment, had caused two breakers (3D23-08 and 3D01-40) to be opened.
Subsequent investigation into plant design basis, identified that opening these two
breakers disabled both channels of AFW automatic actuation logic and relays and
therefore impacted operation of the AFW system on a Unit 3 Low-Low Steam Generator
(SG) Level signal. The licensee did not realize that plant documents used to develop the
AFW clearance were inadequate until the valves failed to open.
Analysis: This finding is greater than minor since it affected the Mitigating System
Cornerstone objective for Equipment Availability and had an actual safety impact of
rendering the automatic start of all AFW pumps on a Low-Low Steam Generator Level
signal unavailable while in Operational Mode 3. The finding was assessed using the
Significance Determination Process for Reactor Inspection Findings, Phase 2 worksheets
for the applicable initiating event likelihood; the exposure time for this condition was less
than 3 days; and the following plant conditions and assumptions were made.
1.
For a Loss of Normal Feedwater design basis accident, annunciator response
procedure 3-ARP-097.CR and emergency operating procedure 3-EOP-E-0
instruct the operator to verify AFW pumps are started and if not, manually start
them. The inspectors found that there was ample time for the operator to perform
this action, given the fact that the unit had already been shutdown for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
11
2.
For a Loss of all Non-Emergency AC Power design basis accident, automatic
actuation of the AFW pumps would still have occurred on a bus stripping signal.
This finding was determined to be of very low safety significance (Green).
Enforcement: 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and
Drawings, states in part that measures shall be established to assure that applicable
regulatory requirements and the design basis are correctly translated into specifications,
drawings, procedures, and instructions. Contrary to this, plant drawing 5613-M-430-
146, Sheet 12 A, Rev.4, Reactor Protection System Control Circuits Train A and
document 5610-E-855, Rev. 490, AC/DC Breaker List, did not adequately reflect the
ESFAS design function. As a result, on March 1, 2003, while in Mode 3, when these
documents were used to develop a clearance of the AFW system, the automatic start
feature of AFW on a Unit 3 Low-Low Steam Generators Level signal was defeated and
rendered all AFW pumps inoperable. Because this self revealing failure for a mitigating
system was determined to be of very low safety significance and has been entered into
the corrective action program (CR 03-0406, 03-2802) this violation is being treated as a
non-cited violation (NCV) consistent with Section VI.A of the NRC Enforcement Policy,
NUREG-1600: NCV 05000250/2003004-01, Failure to Maintain Design Documentation
to Prevent Inadvertent Loss of Both Trains of AFW Automatic Actuation Logic and
Relays. This LER is closed.
.2
(Closed) Licensee Event Report (LER) 05000251/2003-001-00: Channel Failure of the
Qualified Safety Parameter Display System
On April 10, 2003, the 4A Core Exit Thermocouple Subcooling Margin Monitor of the
Qualified Safety Parameter Display System (QSPDS) was not responding during the
power reduction. The licensee identified that some inputs to the 4A Channel of the
QSPDS had stopped responding to actual plant conditions since March 22, 2003. These
inputs were inoperable for more than 18 days which exceeded the 7 days TS Action
Statements 31 and 37 of Accident Monitoring TS 3.3.3.3 for inoperability of the 4A
In-Core Thermocouples, and the Reactor Vessel Level Monitoring System. The licensee
determined the apparent cause of the failure was a sticking input relay on one of the
thermocouple input board on Chassis number 2 of the 4A QSPDS. Corrective actions
included increased monitoring and trending of QSPDS. This finding is greater than minor
because it affected the Reactor Safety Mitigating System cornerstone objective in that
operators may rely on equipment availability and reliability to respond to initiating events
to prevent undesirable consequences during the accidents. The issue was considered
to have very low safety significance because the remaining independent and redundant
channel was operable to provide indication to operators in the control room. This event
did not involve a performance issue and was not assessed through the SDP but was
reviewed by NRC management. This licensee-identified issue involved a violation of TS 3.3.3.3 Action Statements 31 and 37. The enforcement aspects of the violation are
discussed in Section 4OA7. This LER is closed.
12
.3
(Closed) Licensee Event Report (LER) 05000250/2003-006-00: Technical Specification
Required Shutdown Due to Inoperable Containment Isolation Valve.
On April 28, 2003, with Unit 3 in Mode 1 at 100 percent power, the licensee entered TS 3.6.4, Containment Isolation Valves, Action Statement D, which required that the unit be
in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30
hours. This TS was entered in order to effect repairs to an inoperable containment
isolation valve, CV-3-200B, which had excessive leakage. The licensees apparent
cause of the excessive leakage was attributed to wear on the valve stems, cages and
plugs due to a lack of a defined preventative maintenance program of the letdown
isolation valves. Corrective actions included the repair of CV-3-200A, B, C letdown
isolation valves and revision of the preventative maintenance program to improve the
reliability of these components. The inspectors noted that the licensees failure to
perform an as found leak Type C test prior to adjusting the frame cap screws on
CV-3-200B was in violation of plant procedure 0-ADM-531, Containment Leakage Rate
Testing Program and Administrative TS 6.8.4.h, the containment leakage rate testing
requirements. Also, step 5.8.7.2 of 0-ADM-531 procedure states that if as found testing
is not performed, then the component shall be tested at a frequency of at least once per
30 months. A review of the performance of the isolation valves since startup from the
previous refueling outage did not provide any evidence that containment leakage rate
limits required by TS 3.6.1.2 were exceeded prior to the unit shutdown on April 28, 2003.
Therefore, this finding constitutes a violation of minor significance that is not subject to
enforcement action in accordance with Section IV of the NRCs Enforcement Policy. The
licensee documented this violation CR 03-1014. This LER is closed.
.4
(Closed) LER 05000250/2003-008-00: Manual Reactor Trip to Repair Shutdown Bank B
Rod Control System Logic Failure
On May 20, 2003, a controlled shutdown of Unit 3 was conducted due to a failure in the
rod control system. Operator performance was previously discussed in NRC Inspection
Report 250,251/2003-03, Section 1R14. The required Technical Specification (TS) for an
inoperable shutdown control rod bank was followed and there was no violation of TS.
The failure was determined to be due to a degraded connector in a control circuit card.
Several corrective actions were taken including correcting the specific card failure and
scheduling an inspection of all the card connectors with the vendor during the next
refueling outage for both units. This LER is closed.
4OA6 Meetings, including Exit
Exit Meeting Summary
On October 3, 2003, the resident inspectors presented the inspection results to Mr. T.
Jones and other members of his staff, who acknowledged the findings. The inspectors
confirmed that proprietary information was not provided or examined during the
inspection.
13
4OA7 Licensee Identified Violations
The following violation of very low significance (Green) was identified by the licensee and
is a violation of NRC requirements which meet the criteria of Section VI of the NRC
Enforcement Policy, NUREG-1600, for disposition as a non-cited violation (NCV).
Action Statement 31 of TS 3.3.3.3, applicable to In-core Thermocouples, states that if the
number of OPERABLE channels is less than the Total Number of Channels, either
restore the inoperable channel to OPERABLE status within 7 days or be in at least HOT
STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Action Statement 37 of TS 3.3.3.3, applicable to the
Reactor Vessel Level Monitoring System, states that if the number of OPERABLE
channels is one less than the total than the Total Number of Channels, either restore the
inoperable channel to OPERABLE status within 7 days or, if repairs are not feasible
without shutting down, prepare and submit a Special Report to the NRC within 30 days
following the event. On April 10, 2003, the 4A Channel of In-Core Thermocouples and
Reactor Vessel Level Monitoring System had been inoperable for more than18 days and
exceeded the 7 days TS Action Statements 31 and 37 of TS 3.3.3.3. This violation was
identified in licensees corrective action program as CR-03-0909. This finding is only of
very low safety significance because the remaining independent and redundant channel
was operable to provide indication to operators in the control room.
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel:
M. Cornel, Training Manager
M. Chambers, AFW System Engineer
T. Jones, Site Vice-President
M. Lacal, Operations Manager
T. Miller, Maintenance Manager
M. Moore, Performance Improvement Manager
S. Wisler, Health Physics Supervisor
W. Parker, Licensing Manager
M. Pearce, Plant General Manager
W. Prevatt, Work Control Manager
G. Warriner, Quality Assurance Manager
A. Zielonka, Site Engineering Manager
NRC personnel:
K. Green-Bates, Resident Inspector
J. Hanna, Acting Senior Resident Inspector
J. Munday, Branch Chief
S. Ninh, Project Engineer
C. Patterson, Senior Resident Inspector
LIST OF ITEMS OPENED AND CLOSED
Opened and Closed
Failure to Maintain Design
Documentation to Prevent
Inadvertent Loss of Both Trains of
AFW Automatic Actuation Logic and
Relays(Section 4OA3.1)
Closed
05000250/2003-005-00
LER
Disabling Both Auxiliary Feedwater
Trains Inadvertently During Mode 3
(Section 4OA3.1)
2
05000251/2003-001-00
LER
Channel Failure of the Qualified
Safety Parameter Display System
(Section 4OA3.2)
05000250/2003-006-00
LER
Technical Specifications Required
Shutdown Due to Inoperable
Containment Isolation Valve (Section
4OA3.3)
05000250/2003-008-00
LER
Manual Reactor Trip to Repair
Shutdown Bank B Rod Control
System Logic Failure (Section
4OA3.4)