IR 05000335/2008004

IR 05000335-08-004 & 05000389-08-004; on 07/01/2008 - 09/30/2008; St. Lucie Nuclear Plant, Units 1 & 2; Fire Protection and Other Activities

ML083050018
Person / Time
Site:	Saint Lucie
Issue date:	10/30/2008
From:	Marvin Sykes NRC/RGN-II/DRP/RPB3
To:	Stall J Florida Power & Light Co
References
IR-08-004
Download: ML083050018 (28)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ber 30, 2008

SUBJECT:

ST. LUCIE NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000335/2008004, 05000389/2008004

Dear Mr. Stall:

On September 30, 2008, the US Nuclear Regulatory Commission (NRC) completed an inspection at your St. Lucie Plant. The enclosed integrated inspection report documents the inspection findings which were discussed on October 2, 2008, with Mr. Johnston and other members of your staff.

The inspection examined activities conducted under your license as they related 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.

This report documents three NRC identified findings of very low safety significance (Green).

These findings were determined to involve violations of NRC requirements. Additionally, one licensee-identified violation which was determined to be of very low safety significance is listed in Section 4OA7 of this report. However, because of the very low safety significance and because they are entered into your corrective action program, the NRC is treating the findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. 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 facility.

In accordance with 10 CFR 2.390 of the NRCs 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

FP&L 2 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/

Marvin D. Sykes, Chief Rector Projects Branch 3 Division of Reactor Projects Docket Nos. 50-335, 50-389 License Nos. DPR-67, DPR-16

Enclosure:

Inspection Report 05000335/2008004, 05000389/2008004 w/Attachment: Supplemental Information

_________________________ XG SUNSI REVIEW COMPLETE SON OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRS SIGNATURE TLH4 SPS by email RFA by email MDS: rcm SON DAJ2 NAME THoeg SSanchez RAiello MSykes SNinh DJones DATE 10/23/2008 10/23/2008 10/27/2008 10/30/2008 10/23/2008 10/23/2008 10/ /2008 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

FP&L 3

REGION II==

Docket Nos: 50-335, 50-389 License Nos: DPR-67, NPF-16 Report No: 05000335/2008004, 05000389/2008004 Licensee: Florida Power & Light Company (FP&L)

Facility: St. Lucie Nuclear Plant, Units 1 & 2 Location: 6351 South Ocean Drive Jensen Beach, FL 34957 Dates: July 1 to September 30, 2008 Inspectors: T. Hoeg, Senior Resident Inspector S. Sanchez, Resident Inspector R. Aiello, Senior Operations Engineer (Section 4OA7)

D. Jones, Senior Reactor Inspector (Section 4OA5)

Approved by: M. Sykes, Chief Reactor Projects Branch 3 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000335/2008-004, 05000389/2008-004; 07/01/2008 - 09/30/2008; St. Lucie Nuclear

Plant, Units 1 & 2; Fire Protection and Other Activities.

The report covered a three month period of inspection by resident inspectors and two region based reactor inspectors. The significance of most findings is identified 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 NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, and Revision 4, dated December 2006.

A. Inspector Identified & Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Non-Cited Violation (NCV) of 10 CFR 50,

Appendix B, Criterion XVI, Corrective Action, for failure to identify and correct the west side entrance door of the 2B diesel fuel oil storage tank (DFOST)building that was unable to be closed due to a rusted latching mechanism and hinge assembly. The licensee entered the finding in their corrective action program (CAP) for resolution as condition report (CR) 2008-24914.

The finding is greater than minor because it involved the protection against external factors performance attribute of the Mitigating System Cornerstone and affected the objective of ensuring that missile shield equipment is available and capable to prevent damage to mitigating systems. Significance Determination Process (SDP) Phase 1 Screening indicated that the finding is potentially risk significant due to an external event initiator and therefore, a Phase 3 analysis was required. The finding was determined to be of very low safety significance because of the low probability of a strong tornado impacting the region along with the unlikelihood the DFOST would be struck by a generated missile due to its location relative to the subject missile door. In addition, since the only one train of equipment was impacted for less than Technical Specifications (TS) allowed outage time, mitigating systems were available to allow successful core cooling in the event of a tornado. For these reasons, the Phase 3 analysis determined the risk associated with the finding to be

Green.

This finding was related to the identification of issues aspect of the CAP component in the problem identification and resolution crosscutting area (MC 0305 aspect P.1(a)). (Section 1R05)

Green.

The inspectors identified a non-cited violation (NCV) of 10 CFR 50.63 for the licensees non-compliance with the station blackout rule since 1993. The licensee failed to demonstrate that electrical power for Unit 1 could be provided within ten minutes of the onset of a station blackout (SBO) event and subsequently failed to perform a coping analysis when the ten minutes was not demonstrated. The licensee initiated condition report 2007-28746 with an action to perform a coping analysis in lieu of demonstrating that the ten minute commitment could be met. The coping analysis will be reviewed by the NRC (Office of Nuclear Reactor Regulation).

This finding is more than minor because it is associated with the equipment performance attribute and affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to preclude undesirable consequences during a station blackout.

The inspectors did not identify a cross-cutting aspect for this finding. (Section 4OA5.1)

Green.

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65(a)(1) for the licensees failure to monitor the SBO cross-tie cable for Units 1 and 2 against license established goals. The cable has not been tested or energized since 1993. The licensee initiated condition report 2007-36986 for the development of a monitoring program for the cross-tie cable.

This finding is more than minor because it is associated with the design control attribute and affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to preclude undesirable consequences during a station blackout. The inspectors did not identify a cross-cutting aspect for this finding. (Section 4OA5.2)

Licensee Identified Violations

One violation of very low safety significance was identified by the licensee and has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into their corrective action program. This violation and corrective actions are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 and Unit 2 began the period at full Rated Thermal Power (RTP) and operated at full power for the entire period, except for a forced shutdown of Unit 1 due to high chloride levels in the steam generators on August 20, 2008. Unit 1 returned to full power operation on August 26,

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

.1 External Flooding

a. Inspection Scope

The inspectors reviewed lessons learned from previous tropical storm and hurricane events at St. Lucie and performed walkdown inspections of Unit 1 and Unit 2 Auxiliary Feedwater (AFW) pump areas. The inspectors reviewed the applicable Updated Final Safety Analysis Report (UFSAR) section for flooding including specific plant design features to accommodate the maximum flood level. The inspectors reviewed UFSAR Section 13.8.2.3.1 requirements for beach dune inspections and verified the inspections were completed as scheduled. The inspectors also reviewed ADM-04.01, Hurricane Season Preparation, with regard to protective actions to prevent excessive flooding in the AFW Pump area; and reviewed AP-0005753, Severe Weather Preparations, with regard to potential external flooding.

b. Findings

No findings of significance were identified.

.2 Impending Adverse Weather Conditions

a. Inspection Scope

On August 17-18, the inspectors reviewed and verified licensee actions taken in accordance with their procedural requirements prior to the onset of Tropical Storm Fay. The inspectors observed plant conditions and evaluated those conditions using criteria documented in licensee Administrative Procedure 005753, Revision 48, Severe Weather Preparations. The inspectors performed site walkdowns and plant tours to verify the licensee had made the required preparations. The inspectors performed reviews of plant exterior areas vulnerable to high winds and tropical storm conditions including the following areas:

• Turbine Building

• Intake Cooling Water (ICW) Basin

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

Partial Equipment Walkdowns

a. Inspection Scope

The inspectors conducted four partial alignment verifications of the safety-related systems listed below. These inspections included reviews using plant lineup procedures, operating procedures, and piping and instrumentation drawings, which were compared with observed equipment configurations to verify that the critical portions of the systems were correctly aligned to support operability. The inspectors also verified that the licensee had identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers by entering them into the CAP.

• 1B Emergency Diesel Generator (EDG)

• 1B Component Cooling Water (CCW) System

• 1B and 2B Startup Transformers

• 1A and 1B ICW Systems

b. Findings

No findings of significance were identified.

1R05 Fire Protection

.1 Fire Area Walkdowns

a. Inspection Scope

The inspectors toured the following seven plant areas during this inspection period to evaluate conditions related to control of transient combustibles and ignition sources, the material condition and operational status of fire protection systems including fire barriers used to prevent fire damage or fire propagation. The inspectors reviewed these activities against provisions in the licensees procedure ADM-1800022, Fire Protection Plan, and 10 CFR Part 50, Appendix R. The licensees fire impairment lists, updated on an as-needed basis, were routinely reviewed. In addition, the inspectors reviewed the CR database to verify that fire protection problems were being identified and appropriately resolved. The following areas were inspected:

• Unit 1 Turbine Switchgear Room

• Unit 1 CCW Heat Exchanger Platform

• Unit 1 Diesel Fuel Oil Storage Tanks (DFOST) Area

• Unit 2 Control Element Drive Mechanism Control System Room

• Unit 1 43 Elevation Heating and Ventilation Fan Room

• Unit 1 Emergency Core Cooling System (ECCS) Pump Room

• Unit 2 2B Diesel Fuel Oil Storage Tank Area b, Findings

Introduction:

The inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for St. Lucie failing to identify and correct a condition adverse to quality when missile shield door 294 was left open and found unable to be closed due to a rusted latching mechanism and hinge assembly. Specifically, on August 4, 2008, the inspectors identified an open missile shield door on the safety related 2B emergency diesel fuel oil storage tank enclosure building that was required to be closed to prevent damage to the storage tank from flying debris during high wind conditions.

Description:

On the morning of August 4, 2008, the inspectors performed a fire protection walkdown on the 2B EDG fuel oil storage tank building and found the west side entrance door open approximately 3-4 inches. The west side door is a missile shield door required to be closed and latched during normal operations. The inspectors attempted to close the door and determined the hinges and door handle latching mechanism to be corroded and unable to be operated. The inspectors notified the Unit 2 control room supervisor of the as-found condition. The shift manager and Unit 2 auxiliary operator responded to the door and were unable to close and latch the door. The control room supervisor declared the 2B EDG inoperable and entered a 14 day Technical Specification (TS) Limiting Condition of Operation (LCO) until the condition of the door could be evaluated and repaired by the engineering and maintenance departments. Later that day on August 4, the St.

Lucie maintenance department was able to lubricate and exercise the door mechanism to allow it to be closed and latched in its required position and exited the TS LCO.

Following the discovery of the open missile shield door the inspectors reviewed the licensees CR database to determine if this condition was previously identified and if any corrective actions were taken in the past. The inspectors determined that no previous condition existed in the licensees CAP database that identified missile shield door 294 as being open vice closed. The inspectors did find CR 2007-5836 dated 3/21/2007, which identified the door as hard to open due to its hinge pins being rusted and in need of replacement. The licensee wrote work order (WO) 35023080 to replace the hinges at a later date that was not yet scheduled at the completion of this inspection period. There was no reference to the condition of the latching mechanism or the inability to close the door.

Analysis:

The inspectors determined that failure of the licensee to ensure the missile shield door was kept closed during normal operations was a performance deficiency.

The finding is greater than minor because it involved the protection against external factors performance attribute of the Mitigating System Cornerstone and affected the objective of ensuring that missile shield equipment is available and capable to prevent damage to mitigating systems. Significance Determination Process (SDP)

Phase 1 Screening indicated that the finding is potentially risk significant due to an external event initiator and therefore, a Phase 3 analysis was required. The finding was determined to be of very low safety significance because of the low probability of a strong tornado impacting the region along with the unlikelihood the DFOST would be struck by a generated missile due to its location relative to the subject missile door. In addition, since the only one train of equipment was impacted for less than TS allowed outage time, mitigating systems were available to allow successful core cooling in the event of a tornado. For these reasons, the Phase 3 analysis determined the risk associated with the finding to be Green. The inspectors also determined that the cause of this finding was related to the complete, accurate and timely identification of issues potentially impacting the nuclear safety aspect of the corrective action program component in the problem identification and resolution crosscutting area (MC 0305 aspect P.1(a)).

Enforcement:

Criterion XVI of 10 CFR Part 50, Appendix B, states 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 non-conformances, are promptly identified and corrected. Contrary to this requirement, the licensee failed to identify and correct an open degraded missile shield door required to be closed during normal operations. Because the licensee entered the issue into their CAP as CR 2008-24915 and the finding is of very low safety significance (Green), this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000389/2008004-01: Failure to Identify and Correct an Open and Degraded Missile Shield Door.

.2 Fire Protection - Drill Observation

a. Inspection Scope

The inspectors observed a fire drill conducted in the Unit 1 Turbine Building 19.5' Elevation Switchgear Room on July 14, 2008. The drill was observed to evaluate the readiness of the plant fire brigade to fight fires. The inspectors verified that the licensee staff identified deficiencies, openly discussed them in a self-critical manner at the debrief, and took appropriate corrective actions as required. Specific attributes evaluated were:

(1) proper wearing of turnout gear and self-contained breathing apparatus;

(2) proper use and layout of fire hoses;

(3) employment of appropriate fire fighting techniques;

(4) sufficient fire fighting equipment brought to the scene; (5)effectiveness of command and control;

(6) search for victims and propagation of the fire into other plant areas;

(7) smoke removal operations;

(8) utilization of pre-planned strategies;

(9) adherence to the pre-planned drill scenario; and

(10) drill objectives.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures

Internal Flooding

a. Inspection Scope

The inspectors reviewed UFSAR Section 3.4, Water Level (Flood) Design and UFSAR Table 3.2-1, Design Classification of Structures, System and Components for the Unit 1 ECCS Pipe Tunnel. Equipment potentially affected by a flood in this area included several safety-related motor operated valves associated with the high pressure safety injection (HPSI), Low Pressure Safety injection (LPSI), and Containment Spray (CS) systems. The inspectors also reviewed procedure 1-ONP-24.01, Reactor Auxiliary Building Flooding, and verified certain actions required to be taken could be accomplished as written. The inspectors also verified the CAP was being used to identify equipment issues that could be impacted by potential internal flooding.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Training Program

.1 Resident Inspector Quarterly Review

a. Inspection Scope

On September 30, 2008, the inspectors observed and assessed licensed operator actions during a simulated unisolable steam line break in the reactor containment building with complications, to verify that operator performance was adequate and that evaluators were identifying and documenting crew performance problems. The exercise was performed in accordance with St. Lucie Plant Simulator Exercise Guide 0815003, Revision 18. The inspectors also reviewed simulator physical fidelity and specifically evaluated the following attributes related to the operating crews performance:

• Clarity and formality of communication

• Prioritization, interpretation, and verification of alarms

• Control board operation and manipulation, including high-risk operator actions

• Oversight and direction provided by operations supervision, including the ability to identify and implement appropriate TS actions, regulatory reporting requirements, and emergency plan actions and notifications

• Effectiveness of the post-evaluation critique.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed system performance data and associated CRs for the two systems listed below to verify that the licensees maintenance efforts met the requirements of 10 CFR 50.65 (Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants) and licensee Administrative Procedure ADM-17-08, Implementation of 10CFR50.65, Maintenance Rule. The inspectors efforts focused on maintenance rule scoping, characterization of maintenance problems and failed components, risk significance, determination of a(1) and a(2) classification, corrective actions, and the appropriateness of established performance goals and monitoring criteria. The inspectors also interviewed responsible engineers and observed some of the corrective maintenance activities. The inspectors also attended applicable expert panel meetings and reviewed associated system healthreports. The inspectors verified that equipment problems were being identified and entered into the CAP

• Unit 2 EDGs

• Unit 1 CCW System

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors completed in-office reviews, plant walkdowns, and control room inspections of the licensees risk assessment of six emergent or planned maintenance activities. The inspectors verified the licensees risk assessment and risk management activities using the requirements of 10 CFR 50.65(a)(4); the recommendations of Nuclear Management and Resource Council 93-01, Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 3; and procedure ADM-17.16, Implementation of the Configuration Risk Management Program. The inspectors also reviewed the effectiveness of the licensees contingency actions to mitigate increased risk resulting from the degraded equipment. The inspectors interviewed responsible Senior Reactor Operators on-shift, verified actual system configurations, and specifically evaluated results from the online risk monitor (OLRM) for the combinations of out of service (OOS) risk significant systems, structures, and components (SSCs) listed below:

• 1A EDG, 1B CCW Pump

• 1A CCW, 1A ICW Pump, 1A Containment Fan Cooler

• 1B EDG, 1B Charging Pump

• 1A Start Up Transformer, 1A ECCS Pump Room Ventilation

• 1C ICW Pump, 1A LPSI Pump, 1A EDG

• 1C AFW Pump, 1C ICW Pump

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following seven CR interim dispositions and operability determinations to ensure that operability was properly supported and the affected SSCs remained available to perform its safety function with no increase in risk. The inspectors reviewed the applicable UFSAR, and associated supporting documents and procedures, and interviewed plant personnel to assess the adequacy of the interim disposition.

CR 2008-23601, 2B EDG Starting Air System Leakage CR 2008-23656, 1A ICW Pump Check Valve Leakage CR 2008-24625, Valve MV-09-9 Seat Leakage CR 2008-24915, 2B EDG Fuel Oil Storage Tank Missile Shield Door CR 2008-24001, 1A CCW Heat Exchanger Foundation CR 2008-26443, Valve 9280 Seat Leakage CR 2008-29223, 1A and 1B ICW Pump Motor Amperage

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

a. Inspection Scope

The inspectors reviewed the documentation for the following Temporary System Alteration (TSA) associated with Unit 2:

• TSA 02-08-12, Unit 2 K-23 Alarm Logic Power Supply The inspectors reviewed the 10 CFR 50.59 screening and evaluation, fire protection review, environmental review, As Low As Reasonably Achievable (ALARA)screening, and license renewal review, to verify that the modification had not affected system operability/availability. The inspectors reviewed all associated plant drawings and UFSAR documents impacted by this TSA and discussed the changes with plant staff to verify that the installation was consistent with the modification documents.

Additionally, the inspectors verified that problems associated with modifications were being identified and entered into the CAP.

b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

For the six 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 ADM-78.01, Post Maintenance Testing, were incorporated into test requirements.

The inspectors reviewed the following WOs and/or work requests (WR):

• WR 38008533, Fan 2-HVE 9A Relay Replacement

• WO 38014273, Valve 2-V5200 Repair

• WO 36022260, 1C CCW Pump Breaker Maintenance

• WR 38009290, Unit 1 Pressurizer Back Up Heater Bank B1 Power Supply Repair

• WO 38009657, 1C ICW Pump Realignment

• WO 37022734, Valve HCV-08-2B Remote Shutdown Panel Instrument Replacement b Findings No findings of significance were identified.

1R20 Refueling and Other Outage Activities

Unit 1 Forced Outage On August 20, 2008, Unit 1 operators performed a rapid manual downpower and reactor plant shutdown when steam generator chloride levels exceeded normal operating limits. The inspectors observed control room activities during the reactor plant downpower and reactor shutdown, and the reactor startup including synchronizing the turbine generator to the grid.

.1 Monitoring and Shutdown Activities

a. Inspection Scope

The inspectors observed portions of the plant shutdown to hot standby to verify that operating restrictions and similar procedural requirements were followed. The inspectors observed control room operator communications, place keeping, and reviewed chronological log entries.

b. Findings

No findings of significance were identified.

.2 Monitoring of Heat up and Startup Activities

a. Inspection Scope

On August 24, 2008, the inspectors observed activities during the reactor restart to verify that reactor parameters were within safety limits and that the startup evolutions were performed in accordance with licensee procedure 2-GOP-302, Reactor Startup Mode 3 to Mode 2.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors either reviewed or witnessed the following seven surveillance tests to verify that the tests met the TS, the UFSAR, the licensees procedural requirements and demonstrated the systems were capable of performing their intended safety functions and their operational readiness. In addition, the inspectors evaluated the effect of the testing activities on the plant to ensure that conditions were adequately addressed by the licensee staff and that after completion of the testing activities, equipment was returned to the positions/status required for the system to perform its safety function. The tests reviewed included two in-service tests (IST) and one reactor coolant system leakage detection surveillance. The inspectors verified that surveillance issues were documented in the CAP.

• 1-OSP-25.06A, Train A Shield Building Ventilation Monthly Test

• 1-OSP-59.01B, 1B Emergency Diesel Generator Monthly Test

• OP-3200051, Unit 1 Moderator Temperature Coefficient Test

• 1-OSP-21.01C, 1C Intake Cooling Water Pump Test

• 1-OP-0010125A, FCV 25-7 In-service Test

• 2-OSP-01.03, RCS Inventory Balance

• 2-OSP-02.07, Boration Flowpath and Sources Verification

b. Findings

No findings of significance were identified.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation Emergency Preparedness Drill

a. Inspection Scope

On August 6, 2008, the inspectors observed an operating crew in the simulator, operations support center staff, and technical support center staff during a drill of the site emergency response organization. The drill included a loss of off-site power and a Unit 2 station blackout followed by a loss of coolant accident. During the drill the inspectors assessed operator actions to verify that emergency classification and notifications were made in accordance with licensee emergency plan implementing procedures and 10 CFR 50.72 requirements. The inspectors specifically reviewed the Site Area Emergency and General Emergency classifications and notifications were in accordance with licensee procedures EPIP-01, Classification of Emergencies and EPIP-02, Duties and Responsibilities of the Emergency Coordinator. The inspectors also observed whether the initial activation of the emergency response centers was timely and as specified in the licensees emergency plan. The required TS actions for the drill scenario were reviewed to assess correct implementation.

Licensee identified critique items were discussed with the licensee and reviewed to verify that drill weaknesses were identified and captured.

b. Findings

No findings of significance were identified.

2. Radiation Safety Cornerstone

Adverse Weather: Hurricane Ike Although radiation protection inspection activities started during this inspection period, the NRC chose to delay the onsite portion of the inspection until the Hurricane Ike threat had passed. the results of this activity will be documented in a subsequent integrated resident inspection report.

4OA1 Performance Indicator Verification

Initiating Events and Mitigating Systems Cornerstones

a. Inspection Scope

The inspectors checked licensee submittals for the performance indicators (PIs) listed below for the period July 1, 2007, through June 30, 2008, to verify the accuracy of the PI data reported during that period. Performance indicator definitions and guidance contained in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, and licensee procedures ADM-25.02, NRC Performance Indicators, and NAP-206, NRC Performance Indicators, were used to check the reporting for each data element. The inspectors checked operator logs, plant status reports, CRs, system health reports, and PI data sheets to verify that the licensee had identified the required data, as applicable. The inspectors interviewed licensee personnel associated with performance indicator data collection, evaluation, and distribution.

• Unit 1 Safety System Functional Failures

• Unit 2 Safety System Functional Failures

b. Findings

No findings of significance were identified.

4OA2 Problem Identification and Resolution

.1 Daily Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, and to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a screening of items entered daily into the licensees CAP. This review was accomplished by reviewing daily printed summaries of CRs and by reviewing the licensees electronic CR database. Additionally, reactor coolant system unidentified leakage was checked on a daily basis to verify no substantive or unexplained changes.

b. Findings

No findings of significance were identified.

.2 Annual Sample: Review of 2C CCW Pump Suction Header Cracked Socket Weld

a. Inspection Scope

The inspectors selected CR 2008-12847, 2C CCW Pump Suction Valve (V 14140)

Socket Weld Leak, for a more in-depth review of the circumstances that led up to the 2C CCW pump suction header being placed in service with a leaking cracked weld and the corrective actions that followed. On April 15, 2008, the inspectors identified a CCW leak on the running 2C CCW pump suction header from a drain line socket weld connection upstream of valve V14140.

The inspectors reviewed the licensees evaluation of the event and the associated corrective actions. The inspectors reviewed the apparent cause evaluation and interviewed Operations personnel. The inspectors evaluated the licensees administration of this selected condition report in accordance with their CAP as specified in licensee procedure NAP-204, Condition Reporting.

b. Findings and Observations

No findings of significance were identified. On April 15, 2008, the inspectors noticed the unisolable socket weld inboard of the isolation valve was leaking about 3-4 drops per minute. Valve 14140 is a 1 inch drain valve providing pressure boundary isolation for the 2C CCW suction header. The inspectors notified the Unit 2 control room supervisor of the condition who dispatched an operator to evaluate the leak. The St.

Lucie operators verified the cracked socket weld leak and began taking actions to return the 2B CCW train back to service. On April 16, 2008, the 2B CCW train was returned to service and the 2C CCW pump suction header was isolated for repair.

The cracked weld was removed for evaluation and the piping repaired before returning it to service on April 17, 2008, to allow completion of the planned maintenance on the 2B CCW train. The licensee later determined the cracked weld was caused by low stress high cycle fatigue. The associated socket weld was designed to ASME Section III, Class 3 and Seismic Class I, safety related requirements. The licensee visually inspected all other similar CCW header drain line configurations and found no other cracking conditions. In addition, an evaluation of vibration on the system and effects on small bore piping was initiated and was being further reviewed at the close of this inspection period. The inspectors determined the apparent cause analysis of the cracked socket weld was thorough and provided additional details of contributing causes. The corrective actions taken by the licensee or planned were in accordance with their procedure NAP-204, Condition Reporting.

.3 Semi-Annual Trend Review

a. Inspection Scope

As required by Inspection Procedure 71152, Identification and Resolution of Problems, the inspectors reviewed the licensees CAP and associated documents to identify trends that could indicate the existence of a more significant safety issue.

The inspectors review was focused on repetitive equipment issues, but also considered the results of daily inspector corrective actions item screening discussed in Section 4OA2.1 above, plant status reviews, plant tours, document reviews, and licensee trending efforts. The inspectors review nominally considered the six month period of January through June 2008. Corrective actions associated with a sample of the issues identified in the licensees CAP were reviewed for adequacy.

b. Assessment and Observations No findings of significance were identified. During the inspectors review of control room logs and CAP documents, it was apparent that an above average number of unplanned LCOs were entered during the above referenced time period. The licensee documented this observation in CR 2008-21987 to evaluate if indeed this were the case. The licensee determined that after reviewing two years of data, the number of unplanned LCO entries had been consistent between three and six most months, with a high of nine and a low of two. The last few months have followed this trend and therefore, according to the licensee, there is no upward trend of unplanned LCO entries. The inspectors determined the licensee actions taken to be appropriate and timely.

.4 Annual Sample: Review of Operator Workarounds (OWAs)

a. Inspection Scope

The inspectors performed an evaluation of the potential cumulative affects of all outstanding Unit 2 OWAs to determine whether or not they could affect the reliability, availability, and potential for misoperation of a mitigating system; affect multiple mitigating systems; or affect the ability of operators to respond in a correct and timely manner to plant transients and accidents. The inspectors discussed these potential effects with control room supervision and operators. The inspectors also assessed whether OWAs were being identified and entered into the licensees CAP at an appropriate threshold.

b. Findings and Observations

No findings of significance were identified.

4OA3 Event Follow-up

.1 (Closed) LER 05000389/2007-003-00, Manual Reactor Trip Following Misalignment

of Five Control Element Assemblies (CEAs) During Plant Startup On December 29, 2007, Unit 2 was in Mode 3 returning from a scheduled refueling outage when a manual reactor trip was initiated following misalignment of five CEAs (Subgroup 15) after being held on the hold bus for approximately five hours.

Subgroup 15 slipped 20 inches into the core. After extensive troubleshooting and the inability to reproduce the event, the licensee determined that the most probable cause was aging of the hold bus power supplies, with a contributing factor being a voltage isolation card fault. Corrective actions included revising the operating procedures to add a precaution on the use of the hold bus for an extended time, replacing the voltage isolation card and measuring the input and output voltages to ensure it was properly functioning, replacement of the hold bus power supplies during the next refuel outage, and develop a plan to improve the current maintenance practices of the control element drive mechanism system. The LER was reviewed by the inspectors and no findings of significance were identified. This LER is closed.

4OA5 Other Activities

.1 (Closed) URI 05000335/2007006-04: Unit 1 Compliance with the Requirements of

ten CFR 50.63, SBO Rule During the component design bases inspection conducted August 27, 2007, to September 28, 2007, an unresolved item (URI) was identified concerning the capability to restore alternate AC power to Unit 1 within ten minutes of a Station Blackout (SBO). This URI was initially discussed in NRC Report No. 05000335, 389/2006 dated November 5, 2007.

Introduction:

The inspectors identified a non-cited violation (NCV) of 10 CFR 50.63 for the licensees non-compliance with the station blackout rule (SBO) since 1993.

The licensee failed to demonstrate that electrical power for Unit 1 could be provided within ten minutes of the onset of a SBO event and subsequently failed to perform a coping analysis when the ten minutes was not demonstrated.

Description:

10 CFR 50.63 requires that nuclear plants be able to withstand and recover from a station blackout (SBO) of a specified duration. St. Lucie committed to provide alternate AC power to the Unit 1 shutdown buses within ten minutes of the onset of SBO from Unit 2s emergency diesel generators via a cross-tie cable. The licensees commitment to power Unit 1 within ten minutes exempted St. Lucie from performing a coping analyses as required by 10 CFR 50.63(c)(2).

The Safety Evaluation by the Office of Nuclear Reactor Regulation, dated September 12, 1991 documented the NRCs review of St. Lucies commitment for conformance with the SBO rule. The Safety Evaluation stated, in part, that Unit 1s SBO rule conformance was contingent upon FPLs implementation of the following:

• A modification to install a 4160V cable connection between safety buses 1AB of Unit 1 and 2AB of Unit 2B (Cross-tie)

• Ability to power the Unit 1 loads from Unit 2 within ten minutes of the confirmation of a SBO event

• A test to demonstrate that the ten minute commitment could be met In April 1993, the licensee completed the modification that installed the cross-tie. The inspectors determined that after the modification was installed, the licensee failed to perform a test to demonstrate that alternate AC power would be available to power the Unit 1 shutdown buses in ten minutes. Additionally, the licensees current operator lesson plan (0702830) allows operators twenty-five minutes to energize Unit 1 buses following an SBO event. The licensee initiated CR 2007-28746 with an action to perform a coping analysis in lieu of demonstrating that the ten minute commitment could be met. The coping analysis, when completed will be reviewed by the NRC (Office of Nuclear Reactor Regulation).

Analysis:

The performance deficiency is St. Lucies failure to demonstrate that alternate AC power could be provided to the Unit 1 shutdown buses within ten minutes of the onset of SBO. This finding is more than minor because it is associated with the design control attribute and affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to preclude undesirable consequences during a station blackout.

Specifically, the licensee failed to adequately design and implement the station blackout modification to ensure that the shutdown buses would be energized within ten minutes of the onset of a SBO and the licensee subsequently failed to perform a coping analysis to verify Unit 1s capability to withstand the loss of all ac power for 25 minutes. The inspectors determined that the finding is of very low safety significance (Green) because it is a design deficiency confirmed not to result in loss of operability per Part 9900, Technical Guidance, Operability Determination Process for Operability and Functional Assessment. The team concluded there was no immediate safety issue associated with the non-compliance because FSAR Section 15.2.13, Station Blackout Analysis, provides the results of a RETRAN-03 SBO analysis for reactor coolant inventory that states at the end of a four-hour SBO event that the core cooling is maintained and sufficient liquid inventory remains in the vessel to ensure that the core does not uncover and therefore no fuel failures occur.

Additionally, inspectors observed an SBO scenario on the plant simulator and noted that reactor core cooling and containment integrity were not challenged in the twenty-five minutes required to restore AC power to the shutdown buses.

The inspectors did not identify a cross-cutting aspect because the performance deficiency which occurred in 1993 is not reflective of current licensee performance.

Enforcement:

10 CFR 50.63(c)(2) states, in part, that the alternate AC power source will constitute acceptable capability to withstand station blackout provided an analysis is performed which demonstrates that the plant has this capability from onset of the station blackout until the alternate AC source and required shutdown equipment are started and lined up to operate.

Further, 10 CFR 50.63(c)(2) states, in part, that if the alternate AC source can be demonstrated by test to be available to power the shutdown buses within ten minutes of the onset of station blackout, then no coping analysis is required.

Contrary to the above, the licensee failed to demonstrate by test the alternate ac source would be available to power the shutdown buses within ten minutes of the onset of SBO and failed to perform a coping analysis which demonstrates that the plant has the capability to withstand a SBO until the alternate AC source and required shutdown equipment are started and lined up to operate. Specifically, after implementing the SBO modification in 1993, St. Lucie failed to demonstrate by test that alternate ac power would be available within ten minutes of the onset of SBO and then failed to perform the analysis that would demonstrate that the plant has the capability for coping with a station blackout for the twenty-five minutes it takes to energize the shutdown buses. St. Lucie has not been in non-compliance with 10 CFR 50.63(c)(2) since 1993. Because the failure to comply with 10 CFR 50.63, Loss of All Alternating Current Power, is of very low safety significance and has been entered into the licensee's corrective actions program as CR 2007-28746, this violation is being identified as an NCV, consistent with Section VI.A. of the NRC Enforcement Policy. This item is identified as NCV 05000335/2008004-02, Failure to Demonstrate Ten Minute Station Blackout Requirement.

.2 (Closed) URI 05000335/2007006-06: Lack of periodic testing of SBO AAC Recovery

Equipment During the component design bases inspection conducted August 27, 2007, to September 28, 2007, an unresolved item (URI) was identified concerning the periodic testing of the station blackout cross-tie cable. This URI was initially discussed in NRC Report No. 05000335, 389/2006 dated November 5, 2007.

Introduction:

The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65(a)(1) for the licensees failure to monitor the station blackout (SBO) cross-tie cable against license established goals. The cable has not been tested or energized since 1993.

Description:

The Safety Evaluation by the Office of Nuclear Reactor Regulation, dated September 12, 1991 documented the NRCs review of St. Lucies plan for conformance with the SBO rule for Unit 1. The SER stated, in part, that Unit 1s SBO rule conformance was contingent upon FPLs implementation of the following:

• A modification to install a 4160V cable connection between safety buses 1AB of Unit 1 and 2AB of Unit 2B (Cross-tie)

• Establishment of plant procedures to reflect the appropriate testing and surveillance requirements to ensure the operability of the necessary SBO equipment The team determined that the licensee failed to meet their commitment of establishing an appropriate testing and surveillance program for the SBO equipment because the cross-tie cable has not been tested (meggar, high potential testing, etc.

or energized since 1993. The licensees Maintenance Rule Program states that a function of the 4.16KV cables (System 52) which includes the SBO cross-tie, is to supply power to the other unit during SBO conditions via the AB bus. However, the licensees monitoring program failed to test or energize the cross-tie to provide reasonable assurance that it was capable of fulfilling its intended function. The licensee initiated CR 2007-36986.

Analysis:

The performance deficiency is St. Lucies failure to periodically test or energize the cross-tie cable. This finding is more than minor because it is associated with the equipment performance attribute and affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to preclude undesirable consequences (i.e. core damage). Specifically, the lack of periodic testing or energization of the cable resulted in the failure to provide reasonable assurance that the component is capable of fulfilling its safety function during a SBO event. The inspectors determined that the finding is of very low safety significance because it was not was not a design or qualification deficiency, did not result an actual loss of safety function for a system or train, and was not risk significant due to an seismic, fire, flooding, or severe weather initiating event.

The inspectors did not identify a cross-cutting aspect because the cause of the performance deficiency is not reflective of current licensee performance - the licensees failure to establish a monitoring program for the cross-tie cable occurred in 1996.

Enforcement:

10 CFR 50.65(a)(1) states, in part, that licensees shall monitor the performance or condition of structures, systems, and components against license established goals, in a manner sufficient to provide reasonable assurance that such components are capable of fulfilling its intended function.

Contrary to the above, the licensee failed to monitor the performance or condition of a component in a manner sufficient to provide reasonable assurance that such component is capable of fulfilling its intended function in that, the licensee failed to perform appropriate predictive or performance monitoring of the SBO cross-tie to provide reasonable assurance that the cross-tie was capable of fulfilling its intended function during an SBO event. Specifically, since July 1996, when the cross-tie was classified as being within the scope of the monitoring program the licensee has not energized or tested the cable. Because the finding was of very low safety significance, and has been entered into the Corrective Action Program as CR 2007-36986, it is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000335, 389/2008004-03, Failure to Monitor the Station Blackout Cross-tie Cable.

.3 (Open) NRC TI 3515/176, Emergency Diesel Generator Technical Specification

Requirements Regarding Endurance and Margin Testing

a. Inspection Scope

The objective of this TI was to gather information to assess the adequacy of nuclear power plant emergency diesel generator (EDG) endurance and margin testing as prescribed by plant-specific technical specifications (TS). The inspector interfaced with the appropriate station staff to obtain the information specified in Attachment 1 of the TI, Worksheet. The TI applies to all operating nuclear power reactor licensees that use EDGs as the onsite standby power supply. The inspector verified the accuracy of the information by review of TS, EDG Design Basis Event (DBE) loading calculations, EDG endurance run test procedures, test data from the last three endurance tests performed on each EDG, EDG ratings, and EDG operating history.

The information gathered will be forwarded to Nuclear Reactor Regulation/Division of Engineering/Electrical Engineering Branch (NRR/DE/EEEB) for further review to assess the adequacy and consistency of EDG testing at nuclear stations.

b. Findings and Observations

The TI is presently scheduled to be open until August 31, 2009, pending completion of the NRR/DE/EEEB review.

4OA6 Exit

Exit Meeting Summary

The resident inspectors presented the inspection results to Mr. Johnston and other members of licensee management on October 2, 2008. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary information. The licensee did not identify any proprietary information.

A regional based reactor inspector also presented the in-office review inspection results of two Unresolved Items (URIs) with Mr. E. Katzman and other members of the staff on September 9, 2008.

4OA7 Licensee Identified Violations

The following violation of very low safety significance (Severity Level IV) was identified by the licensee and is a violation of NRC requirements which met the criteria of Section VI of the NRC Enforcement policy, NUREG-1600, for NCV.

• 10 CFR 55.25 states If, during the term of the license, the licensee develops a permanent physical or mental condition that causes the licensee to fail to meet the requirements of § 55.21 of this part, the facility licensee shall notify the Commission, within 30 days of learning of the diagnosis, in accordance with § 50.74(c). For conditions for which a conditional license (as described in §55.33(b)of this part) is requested, the facility licensee shall provide medical certification on Form NRC 396 to the Commission (as described in § 55.23 of this part).

Contrary to the above, on June 13, 2008, the licensee discovered that they had failed to notify the Commission within 30 days after one licensed operator was diagnosed with a permanent physical medical condition as required by 10 CFR 55.25. This finding was evaluated using the traditional enforcement process because it impacted the Commissions ability to perform its regulatory licensing function. This finding was of very low safety significance because the medication was prescribed and/or the condition was under control with no impact on the individuals ability to perform licensed duties. The licensee has entered this deficiency into their CAP as CR 2008-19885.

ATTACHMENT: SUPPPLEMENTAL INFORMATION

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

D. Calabrese, Emergency Preparedness Supervisor

D. Cecchett, Licensing Engineer

T. Cosgrove, Site Engineering Manager

C. Costanzo, Plant General Manager

M. Delowery, Maintenance Manager

K. Frehafer, Licensing Engineer

B. Jacques, Security Manager

G. Johnston, Site Vice President

E. Katzman, Licensing Manager

R. McDaniel, Fire Protection Supervisor

M. Moore, Radiation Protection Manager

M. Page, Acting Operations Manger

W. Parks, Work Control Manager

T. Patterson, Performance Improvement Department Manager

M. Snyder, Site Quality Manager

G. Swider, Systems Engineering Manager

NRC personnel

M. Sykes, Chief, Branch 3, Division of Reactor Projects

S. Ninh, Senior Projects Engineer, Branch 3, Division of Reactor Projects

D. Jones, Sr. Reactor Inspector

B. Desai, Branch Chief, Engineering Branch 1

LIST OF ITEMS

Opened

NONE

Opened and Closed

05000389/2008004-01 NCV Failure to Identify and Correct an Open and Degraded Missile Shield Door (1R05.1)

05000335/2008004-02 NCV Failure to Demonstrate Ten Minute Station Blackout Requirement (4OA5)

05000335, 389/2008004-03 NCV Failure to Monitor the Station Blackout Cross-tie Cable (4OA5)

Closed

05000335/2007006-04 URI Unit 1 Compliance with the Requirements of 10 CFR 50.63, SBO Rule (4OA5)

05000335/2007006-05 URI Lack of Periodic Testing of SBO AAC Recovery Equipment (4OA5)

05000389/200 7-003-00 LER Manual Reactor Trip Following Misalignment of Five Control Element Assemblies During Plant Startup (4OA3.1)

Discussed

05000335, 389/2515/176 TI Emergency Diesel Generator Technical Specification Requirements Regarding Endurance and Margin Testing