Text

IR 05000250-08-006; 05000251-08-006; 09/15/08 - 10/10/08; Turkey Point Nuclear Power Plant, Units 3 & 4; Component Design Basis Inspection
	ML083530016
Person / Time
Site:	Turkey Point
Issue date:	12/18/2008
From:	Binoy Desai; Division of Reactor Safety II, NRC/RGN-II/DRS/EB1
To:	Stall J; Florida Power & Light Co
References
	IR-08-006
	Download: ML083530016 (42)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION cember 18, 2008

SUBJECT:

TURKEY POINT NUCLEAR PLANT - NRC COMPONENT DESIGN BASIS INSPECTION REPORT 05000250/2008006 AND 05000251/2008006

Dear Mr. Stall:

On November 17, 2008, the US Nuclear Regulatory Commission (NRC) completed an inspection at your Turkey Point Units 3 and 4. The enclosed inspection report documents the inspection findings which were discussed on October 10, 2008, with Mr. Jefferson and other members of your staff, and again on November 17, 2008, with Mr. Tomonto 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 team reviewed selected procedures and records, observed activities, and interviewed personnel.

This report documents two NRC identified findings of very low safety significance (Green). Both of these findings were determined to involve violations of NRC requirements. Because these findings were of very low safety significance, and they were entered into your corrective action program, the NRC is treating these findings as non-cited violations consistent with Section VI.A.1 of the NRC Enforcement Policy. If you wish to contest these non-cited violations, 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 20555-001; 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 Turkey Point.

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/

Binoy B. Desai, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-250, 50-251 License Nos.: DPR-31, DPR-41

Enclosure:

Inspection Report 05000250/2008006 and 05000251/2008006 w/Attachment: Supplemental Information

SUNSI REVIEW COMPLETE OFFICE RII:DRS RII:DRS RII:NRO RII:DRS RII:DRS RII:DRS RII:DRS Contractor Contractor SIGNATURE /RA/ /RA/ /RA/ /RA /RA/ /RA/ /RA/ /RA/ /RA/

NAME W Lewis R Baldwin J.Bartleman C.Even J.Eargle M.Coursey R.Williams S.Kobylarz C.Edwards DATE 12/15/2008 12/15/2008 12/15/2008 12/15/2008 12/15/2008 12/15/2008 12/15/2008 12/15/2008 12/15/2008

FP&L 3

REGION II==

Docket Nos.: 50-250, 50-251 License Nos.: DPR-31, DPR-41 Report No: 05000250/2008006, 05000251/2008006 Licensee: Florida Power & Light Company (FP&L)

Facility: Turkey Point Nuclear Plant, Units 3 & 4 Location: 9760 S. W. 344th Street Florida City, FL 33035 Dates: September 15 to November 17, 2008 Team: R. Lewis, Senior Reactor Inspector (Lead)

R. Baldwin, Senior Operations Inspector J. Bartleman, Senior Construction Inspector C. Even, Reactor Inspector S. Kobylarz, Contract Electrical Inspector C. Edwards, Contract Mechanical Inspector Accompanied by: M. Coursey, Reactor Inspector (Training)

R. Williams, Reactor Inspector (Training)

J. Eargle, Reactor Inspector (Training)

Approved by: B. Desai, Branch Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000250/2008-003, 05000251/2008-003; 9/15/2008 - 10/10/2008; Turkey Point

Nuclear Power Plant, Units 3 and 4; Component Design Basis Inspection.

This inspection was conducted by a team of four NRC inspectors from the Region II office, and two NRC contract inspectors. There were two findings of very low safety significance (i.e. Green) identified by the activities covered in this inspection. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, (ROP)

Revision 4, dated December 2006.

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

The team identified a finding of very low safety significance involving two examples of a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III,

Design Control. Specifically, in one example, the licensee failed to establish design control measures to verify or check the adequacy of critical design inputs for the calculation that determined that the Unit 3A and 3B emergency diesel generators (EDGs) would be operable at outside ambient temperatures at or below 75ºF if the 3A and 3B Diesel Building Ventilation Fans are out of service as allowed by procedures. In the second example, the licensee failed to ensure the adequacy of design for the auxiliary feed water steam generator flow control valves.

The finding was more than minor because it was associated with the Design Control attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

The team assessed the finding using the SDP and determined that the finding was of very low safety significance (Green) since it was a design deficiency determined not to have resulted in the loss of safety function. Specifically, in both cases, the licensee had not operated in a condition for which the design deficiencies in question were relied upon for operation. Both examples had been entered in the licensees corrective action system. The finding was evaluated for cross-cutting aspects and none were identified. (Sections 1R21.2.3 and 1R21.2.4)

Green.

The team identified a NCV of 10 CFR Part 50, Appendix B, Criterion III,

Design Control, having very low safety significance involving the failure to verify and ensure that the 125 VDC safety-related batteries would remain operable if all the inter-cell and terminal connections were at the resistance value (< 150 micro-ohms) allowed by technical specification (TS) surveillance requirement (SR)4.8.2.1.b(2)/c(3) and maintenance procedure 0-SME-003.3/.4/.15.

The finding was more than minor because if left uncorrected, the finding would become a more significant safety concern. Specifically, the 125 VDC safety-related batteries would become incapable of meeting their design basis function if the inter-cell and connection resistance were allowed to increase to the TS allowed value. The finding was of very low safety significance since it was a design deficiency determined not to have resulted in the loss of safety function.

No cross cutting aspect was identified for this finding. The licensee entered this deficiency into their corrective action program. (Section 1R21.2.16)

Licensee-Identified Violations

None

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R21 Component Design Bases Inspection

Main article: IP 71111.21

.1 Inspection Sample Selection Process

The team selected risk significant components and operator actions for review using information contained in the licensees Probabilistic Risk Assessment (PRA). In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum value greater than 1 X10-6. The components selected were located within the high head safety injection (SI), auxiliary feed water (AFW), emergency diesel generator (EDG), accumulator, reactor protection, AC and DC electrical distribution, intake cooling water (ICW) and residual heat removal (RHR) systems. The sample selection included 21 components, 5 operator actions, and 4 operating experience items. Additionally, the team reviewed 5 modifications by performing activities identified in IP 71111.17, Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases had been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These included items such as failed performance test results, significant corrective action, repeated maintenance, maintenance rule (a)1 status, RIS 05-020 (formerly GL 91-18) conditions, NRC resident inspector input of problem equipment, System Health Reports, industry operating experience and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.

.2 Results of Detailed Reviews

.2.1 Reactor Water Storage Tank (RWST) Valve to SI & RHR Pumps, MOV-3/4-864A/B

a. Inspection Scope

The team reviewed the Technical Specifications (TS), Final Safety Analysis Report (as updated, UFSAR), and design basis (DB) documentation to identify the component design basis function and related accident analysis assumptions. The team conducted a visual inspection of these Motor-Operated Valves (MOVs) to verify that degraded material conditions were being addressed appropriately. In addition, the team verified that the power demand requirements for the valves were captured in electrical load and degraded voltage calculations. The team also verified that the worst case/highest differential pressure (dP) was used to determine the maximum valve opening and/or closing requirements to ensure the valves would perform their intended safety-related design basis function. A review was conducted of the licensees testing procedures and results from actual diagnostic valve testing that was performed to verify the MOVs were tested in a manner that would detect a malfunctioning valve and verify proper operation of the valve. A review of operating and maintenance procedures was conducted to verify that a malfunctioning MOV could be identified by operators, and that the valves were being properly maintained and corrected. The team reviewed historical maintenance work orders and modification packages associated with these valves to verify they were being maintained and changed accordingly to ensure reliability and margin for safe operation.

b. Findings

No findings of significance were identified.

.2.2 AFW Turbine Trip and Throttle Valve, MOV-6459A/B/C

a. Inspection Scope

The team reviewed the TS, UFSAR, and DB documentation to identify the component design basis functions and related accident analysis assumptions. The team conducted a visual inspection of these MOVs to verify that degraded material conditions were being addressed appropriately. In addition, the team verified that the power demand requirements for the valves were captured in electrical load and degraded voltage calculations. The team also verified that the worst case/highest dP was used to determine the maximum valve opening and/or closing requirements to ensure the valves would perform their intended safety-related design basis function. A review was conducted of the licensees testing procedures and results from actual diagnostic valve testing that was performed to verify the MOVs were tested in a manner that would detect a malfunctioning valve and verify proper operation of the valve. A review of operating and maintenance procedures was conducted to verify that a malfunctioning MOV could be identified by operators, and that the valves were being properly maintained and corrected. The team reviewed historical maintenance work orders and modification packages associated with these valves to verify they were being maintained and changed accordingly to ensure reliability and margin for safe operation.

b. Findings

No findings of significance were identified.

.2.3 Auxiliary Feedwater (AFW) Pump, P2A/B/C

a. Inspection Scope

The team reviewed the design basis documentation, pump vendor manual and related vendor correspondence, drawings, and the UFSAR to identify design, maintenance, and operational requirements related to pump flow and developed head, achieved system flow, net positive suction head (NPSH), vortex formation and prevention, minimum flow requirements, and run out protection. Calculations supporting NPSH requirements and vortex prevention were reviewed to verify that design bases and design assumptions have been appropriately translated into design calculations. Operating procedures for aligning AFW pumps during a small break LOCA or other event scenario that causes a loss of main feed water were reviewed to verify that operation of this component is consistent with design basis requirements and analyzed conditions. System Health Reports, corrective maintenance and condition report (CR) historical records were reviewed to assess adverse trends and to verify that potential degradation was monitored or prevented. Test procedures and acceptance values were reviewed to verify that acceptance criteria were supported by calculations or other engineering documents, and to ensure that design and licensing bases were met and that tests and/or analyses validate component operation under accident/event conditions. A component and system walk down was performed to verify that the component installed configuration would support its design basis function under accident/event conditions, that component configurations had been maintained consistent with design assumptions, and that the equipment was adequately protected and not impeded by recent security modifications.

b. Findings

Introduction The team identified a finding of very low safety significance (Green)involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control. Specifically, in this example, the licensee failed to establish design control measures to verify or check the adequacy of design of the flow control valves (FCVs)across the full range of anticipated operating conditions. Specifically, the design of the valve was not verified for the flow condition when the valve was throttled at the near-closed position during shutdown cooling. An additional example of this design control finding is discussed in Section 1R21.2.4.

Description The licensee implemented an AFW constant-speed governed flow management in 1986 (PC/M 83-049). Continued flow management improvements were sought in 1986 through PC/M 85-131, by changing to a reduced port trim in the FCVs downstream of the pump. Both of these changes resulted in non-conservative contributions to the tendency towards cavitation and choked flow in the FCVs; however, no analysis or review of these effects was performed at that time. Consequently, the licensees procurement documents for the new trim package called for soft seat material and standard, non-hardened stainless steel seat rings/plugs, neither of which are generally considered acceptable under extended cavitation conditions. Additionally, no consideration was given to offset these effects through other changes, such as implementing a multi-ported valve package.

The teams review of the 2006 rebuild of the AFW pump (PC/M 05-029) identified further non-conservative contribution to the preceding flow relationship by increasing the pumps total dynamic head (TDH) by approximately 250 feet across the speed spectrum.

Once again, no analysis or review of the susceptibility to cavitation and choked flow was performed. These phenomenon challenge component and system level availability, reliability, and capability to provide cooling water to the steam generators when the steam generators are used for cooling the reactor coolant system (RCS).

Using generic flow nozzle cavitation methodology presented in Crane Technical Paper 410, the team and licensee initially determined that for certain expected operating flow/pressure conditions, both cavitation and choked flow could occur, even for the original weaker pumps from 1986 to 2006. However, the licensee then engaged the valve vendor (Flowserve) and, utilizing control valve sizing software, the analysis was refined to determine that although cavitation will be present at lower steam generator pressures, choked flow conditions would not be achieved (with the possible exception of a steam generator failure). As the AFW system was not normally used for RCS cooldown purposes (for which it is credited in events such as small break LOCA and loss of offsite power), the team found no occurrence of prolonged operation in the susceptible flow regime that could jeopardize the valves ability to perform their safety-related function.

Analysis The licensees failure to establish measures to verify or check the adequacy of design inputs supporting unbounded AFW flow to the steam generators across the full range of anticipated operating conditions was determined to be a performance deficiency. The finding is more than minor because it is associated with the Design Control attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of the AFW system to perform its intended safety function. The team reviewed the finding using the Phase 1 significance determination process (SDP) worksheet for mitigating systems and determined the finding was of very low safety significance (Green) since it was a design deficiency determined not to have resulted in the loss of safety function. This finding was reviewed for cross-cutting aspects and none were identified since the performance deficiency was not indicative of current licensee performance. The finding was entered into the licensees corrective actions program as CR 2008-31041.

Enforcement 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design. Contrary to the above, the licensee failed to verify the adequacy of the AFW flow control valve design. Specifically, the licensee failed to verify that the cavitation index of the flow control valves, in combination with the system operating characteristics, would preclude choked flow conditions. However, because this violation was of very low safety significance, and since it was entered into the licensees corrective action program as CR 2008-31041, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000250(251)/2008-06-01, Inadequate Design Verification Affecting AFW Flow Control Valves and Unit 3 EDG Building Room Ventilation. This is example 1 of NCV 05000250(251)/2008-06-01.

.2.4 Emergency Diesel Generator (EDG) 3A/3B Room Ventilation

a. Inspection Scope

The team reviewed the TS, UFSAR, and DB documentation to identify the component design basis functions and related accident analysis assumptions. Calculations supporting the installed system capability were reviewed to verify that design bases and assumptions were appropriately translated and that conclusions supported overall system capability. Elementary electrical diagrams and selected preventative maintenance history were reviewed to verify that energy sources, including those used for control functions, would be available and unimpeded during accident/event conditions.

A component and system walkdown was performed in order to verify that the components installed configuration supported its design function under accident/event conditions. Selected CRs and work orders were reviewed by the team in order to verify that potential degradation was monitored or prevented and that component replacement was consistent with in-service/equipment qualification life. Operating procedures were reviewed to verify that operator actions were consistent for accident/event conditions.

b. Findings

Introduction The team identified an additional example of this design control finding, previously discussed in Section 1R21.2.3. The finding was of very low safety significance (Green) involving a NCV of 10 CFR 50, Appendix B, Criterion III, Design Control. Specifically, in this example, the licensee failed to establish design control measures to verify or check the adequacy of critical design inputs for the calculation that determined that the Unit 3A and 3B Emergency Diesel Generators (EDG) would be operable at outside ambient temperatures at or below 75ºF if the 3A and 3B Diesel Building Ventilation Fans were out of service.

Description The teams review of the 3A and 3B EDG Building Ventilation Systems determined that the 3A and 3B EDG Building Ventilation Fans were allowed to be taken out of service at temperatures at or below 75ºF without declaring the respective EDGs inoperable. Procedure 3-OP-023, Emergency Diesel Generator, dated 04/21/08, and Procedure 3-OSP-023.1, Diesel Generator Operability Test, dated 04/21/08 stated that the EDG was to be considered operable if the EDG Room Exhaust Fan failed to start and the outdoors ambient temperature was at or below 75ºF. Otherwise, the diesel was to be declared inoperable if the fan failed to start.

Calculation PTN-BFSM-00-006, Unit 3 EDG Maximum Outdoor Temperature, dated 09/27/00, calculated the required air flow with the outside temperature at 75ºF, and the exhaust air fan out of service to be 12,600 cubic feet per minute (cfm). PTN-BFSM-00-006 also stated that the natural circulation flow rate available was calculated to be only 12,000 cfm, but that this was acceptable based on Condition Report CR 98-612, Thermal Overload Trip, dated 04/10/98. CR 98-612 was generated as a result of the 3A EDG being declared inoperable due to tripped thermal overloads on the 3A EDG Building Ventilation Fan. CR 98-612 stated that in addition to the 12,000 cfm of natural circulation, the diesel engines would intake 10,725 cfm during operation and this would increase the overall flowrate into the room. Therefore, according to CR 98-612 the bulk room temperature would be expected to remain less than 115 ºF, since the calculated natural circulation flowrate and the EDG induced flow far exceed the required airflow rate in the room.

The team noted that the EDG intake airflow could sufficiently disrupt establishment and maintenance of the natural circulation flow circuit. The team concluded that the calculation assumption allowing an additive relationship between the natural circulation and forced circulation flow circuits was a critical design assumption. Further, this assumption was principal in ensuring EDG operability at the established 75ºF threshold and was not checked or verified for adequacy by analysis or test. The licensee indicated that a more comprehensive analysis would be required to ensure a maximum bulk room temperature of 115 ºF was maintained under these conditions. During the inspection, the applicable procedures were modified to remove the statement which indicated the EDGs were operable without ventilation fans at an ambient temperatures at or below 75ºF.

Analysis The licensees failure to establish measures to verify or check the adequacy of design inputs for the calculation that determined that the Unit 3A and 3B EDGs would be operable at outside ambient temperatures at or below 75ºF if the 3A and 3B Diesel Building Ventilation Fans were out of service was a performance deficiency. The finding was more than minor because it was associated with the Design Control attribute of the Mitigating System Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of the EDGs to perform their intended safety function. The team reviewed the finding using the Phase 1 SDP worksheet for mitigating systems and determined the finding was of very low safety significance (Green) since it was a design deficiency determined not to have resulted in the loss of safety function.

Specifically, the EDGs had not been operated without the building ventilation fans since the procedural allowance was added. This finding was reviewed for cross-cutting aspects and none were identified since the performance deficiency was not indicative of current licensee performance. The finding was entered into the licensees corrective actions program as CR 2008-29507.

Enforcement 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design. Contrary to the above, design control measures were not provided to verify or check the adequacy of the design inputs for the calculation that determined that the Unit 3A and 3B EDGs would be operable at outside ambient temperatures at or below 75ºF if the 3A and 3B Diesel Building Ventilation Fans were out of service as allowed by procedures. Because the finding was of very low safety significance and was entered into the licensee's corrective action program (CR 2008-29507), this violation is being treated as a non-cited violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy: NCV05000250(251)/2008-06-01, Inadequate Design Verification Affecting AFW Flow Control Valves and Unit 3 EDG Building Room Ventilation). This is example 2 of NCV 05000250(251)/2008-06-01.

.2.5 RHR Heat Exchanger, 3/4E206A/B

.2.10 Hand Control Valve, HCV-3/4-758

a. Inspection Scope

The team reviewed the TS, UFSAR, and DB documentation to identify the component design basis functions and related accident analysis assumptions. Drawings, procedures, and a system walkdown with operators was used to verify that component operation and alignments were consistent with design and licensing basis assumptions.

Calculations supporting the installed system capability were reviewed to verify that design bases and design assumptions were appropriately translated into design calculations and procedures. Operating procedures for controlling valve position in response to cooling requirements were reviewed to verify that operation of this component flow path was consistent with design basis requirements and analyzed conditions. Additionally, operator walkthroughs were conducted in order for the team to verify that instrumentation and alarms were available to operators for making necessary decisions. System Health Reports, corrective maintenance and CR historical records were reviewed, and engineering personnel interviews were conducted to verify that potential degradation was monitored or prevented and that component replacement was consistent with in-service/equipment qualification life. A component walk down was performed to verify that the component installed configuration would support its design basis function under accident/event conditions, that component configurations had been maintained consistent with design assumptions, and that the equipment was adequately protected and not impeded by recent security modifications.

b. Findings

No findings of significance were identified.

.2.11 AFW Min Flow Path Including Restricting Orifice for AFW Pump Recirc to Condensate

Description In reviewing the safety related station battery sizing and voltage drop calculation, the team found that the licensee had not considered the TS allowed limit on battery intercell connection resistance in determining the voltage drop to critical safety-related components, such as vital inverters. For example, the licensee found that for inverter 3Y07 powered from the 4A battery, the calculated voltage available at the inverter was 103 volts during the 1st minute of the design basis two hour profile, which considered the inverter at fully loaded conditions for conservatism. The licensee stated that 103 volts was also the vital inverter design basis minimum voltage. However, the team found that the calculated voltage was determined by the licensee without any consideration for either the TS 150 micro-ohms acceptance criteria or the Maintenance Level Acceptance Criteria (MLAC) resistance for intercell connections, which were approximately 6 micro-ohms. When the MLAC resistance limit was considered for the battery intercell connections, the licensee determined that the voltage at the inverter would by 0.23 volts less (102.77 volts) during the 1st minute load. When the TS allowable 150 micro-ohm limit was considered for the intercell connections, the team found that the additional 1st minute voltage drop would be approximately 6.83 volts, which resulted in 96.17 volts at the inverter.

In response to the teams concern on the adequacy of the available voltage at the inverter, the licensee performed a preliminary evaluation for the MLAC intercell resistance, which also considered the actual calculated accident load on the inverter versus a fully loaded inverter, and determined that a slight positive margin of 0.11 volts (or 103.11 volts) would result. As part of the preliminary analysis, the licensee used an assumed inverter efficiency for the accident load condition. The team was advised by the licensee that any such design assumptions would be verified as part of an updated battery sizing and voltage analysis to be performed by the licensee to address the MLAC limits for battery intercell connections. The above items were entered in the licensees corrective action program as CR 2008-29004.

Analysis The team determined that the licensees failure to perform the required calculation and verify that the resistance value (< 150 micro-ohms) specified in TS SR 4.8.2.1.b(2)/c(3) and maintenance procedures 0-SME-003.3/.4/.15 was sufficient to ensure safety-related battery operability was a performance deficiency and a finding.

The team performed a margin assessment and detailed review of five risk significant and time critical operator actions. Where possible, margins were determined by the review of the assumed design basis and UFSAR response times. For the selected operator actions, the team performed a walkthrough of associated Emergency Operating Procedures (EOPs), Off-Normal Operating Procedures (ONPs), Annunciator Response Procedures (ARPs), and other operations procedures with appropriate plant operators and engineers to assess operator knowledge level, adequacy of procedures, availability of special equipment when required, and the conditions under which the procedures would be performed. Detailed reviews were also conducted with operations and training department leadership, and through observation and utilization of a simulator training period to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements.

Operator actions were observed on the plant simulator and during plant walk downs.

The following operator actions were observed on the licensees operator training simulator:

Loss of Instrument Air, actions to recover Instrument Air and how to operate AFW and Main Feed Water without Instrument Air: (0-ONOP-013, Loss of Instrument Air, 3-EOP-E-0, Reactor Trip or Safety Injection, 3-EOP-ES-0.1)

Mode 4 Small Break LOCA, actions to use procedures associated with shutdown conditions: (3-ONOP-41.3, Excessive Reactor Coolant System Leakage and 3-ONOP-41.7, Shutdown LOCA [Mode 3 (Less than 1000 PSIG) or Mode 4])

Reactor Coolant Pump Locked Rotor /ATWS / Seal Package Failure, actions to recover the unit using manual actions to shutdown the reactor manually while mitigating a loss of coolant event: (3-EOP-E-0, 3-EOP-FR-S.1, Response to Nuclear Power Generation/ATWS, 3-EOP-E-1, Loss of Reactor or Secondary Coolant)

Additionally, the team walked down, table-topped and investigated a number of aspects associated with the above, actions supporting the component list associated with this inspection, as well as the following specific operational scenarios:

Manual Valve Controls

Local/Manual actions to re-establish RHR Cooling. Procedures 0-ONOP-105, Control Room Evacuation. Attachments 7 (Non-fire brigade Number 1 Senior Nuclear Plant Operator (Outside SNPO)) and 8 (Non-fire brigade Number 1 Senior Nuclear Plant Operator (Inside SNPO)).

b. Findings

No findings of significance were identified.

.4 Review of Industry Operating Experience

a. Inspection Scope

The team reviewed selected operating experience issues that had occurred at domestic and foreign nuclear facilities for applicability at the Turkey Point Nuclear Plant. The team performed an independent applicability review for issues that were identified as applicable to the Turkey Point Nuclear Plant and were selected for a detailed review.

The issues that received a detailed review by the team included:

IN 90-45, Turbine Driven Auxiliary Feedwater Pump Overspeed

IN 2006-026, Failure of Magnesium Rotors in Motor-operated Valve Actuators

IN 2002-12, Submerged Safety-Related Electrical Cables

IN 2006-03, Motor Starter Failures Due To Mechanical-Interlock Binding

b. Findings

No findings of significance were identified.

.5 Review of Permanent Plant Modifications

a. Inspection Scope

The team reviewed five modifications related to the selected risk significant components in detail to verify that the design bases, licensing bases, and performance capability of the components have not been degraded through modifications. The adequacy of design and post modification testing of these modifications was reviewed by performing activities identified in IP 71111.17, Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications. The following modifications were reviewed:

PC/M No.98-049, MOV Enhancement - Limitorque Technical Update 98-01, for MOV-4-863A/B, MOV-4-864A/B, MOV-4-872, MOV-4-843A/B, MOV-4-1417 and MOV-4-1418, Rev. 1

PC/M No.99-019, MOV Enhancement - Limitorque Technical Update 98-01, for MOV-3-863A/B, MOV-3-864A/B, MOV-3-872, MOV-3-843A/B, MOV-3-880A/B, MOV-878A and MOV-3-744A/B , Rev. 2

PC/M No.99-056, MOV Enhancement - Limitorque Technical Update 98-01, for MOV-4-880A/B, MOV-4-6386, MOV-4-1420, MOV-4-1421, MOV-878B and MOV-4-744A/B, Rev. 2

PC/M No.05-097, Revise the Control Logic for MOV-3-1420, 1421, 1425, 1426 and 1427 to Limit Close, Rev. 0

PCM 08-031, Unit 3 EDG A & B Radiators Replacement, Rev. 0

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

4OA6 Meetings, Including Exit

Exit Meeting Summary

On October 10, 2008, the team presented the inspection results to Mr. Jefferson and other members of the licensee staff. On November 17, 2008, the team re-exited with Mr.

Tomonto and other members of the licensee staff to communicate changes in the results of the inspection. No proprietary information was reviewed as part of this inspection.

4OA7 Licensee-Identified Violations

None.

ATTACHMENT: SUPPPLEMENTAL INFORMATION

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

Sergio Chaviano, Design Engineering Manager

Al Dunstan, Staff Engineer

Ron Everett, Licensing Engineer

Mike Danford, Inspection Support Manager

Phil Barnes, Mechanical Design Engineering Supervisor

NRC personnel

Scott Stewart, Senior Resident Inspector

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened and Closed

05000250&251/2008-06-01 NCV Inadequate Design Verification Affecting AFW Flow Control Valves and Unit 3 EDG Building Room Ventilation

05000250&251/2008-06-02 NCV Failure to Verify Limiting Value Adequacy for Safety-Related Battery Intercell Resistance

LIST OF DOCUMENTS REVIEWED

IR 05000250/2008006

Text

SUBJECT:

Dear Mr. Stall:

Enclosure:

SUMMARY OF FINDINGS

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Licensee-Identified Violations

REPORT DETAILS

REACTOR SAFETY

1R21 Component Design Bases Inspection

.1 Inspection Sample Selection Process

.2 Results of Detailed Reviews

.2.1 Reactor Water Storage Tank (RWST) Valve to SI & RHR Pumps, MOV-3/4-864A/B

a. Inspection Scope

b. Findings

.2.2 AFW Turbine Trip and Throttle Valve, MOV-6459A/B/C

a. Inspection Scope

b. Findings

.2.3 Auxiliary Feedwater (AFW) Pump, P2A/B/C

a. Inspection Scope

b. Findings

.2.4 Emergency Diesel Generator (EDG) 3A/3B Room Ventilation

a. Inspection Scope

b. Findings

.2.5 RHR Heat Exchanger, 3/4E206A/B

a. Inspection Scope

b. Findings

.2.6 Low Head Safety Injection (LHSI) Discharge Isolation, MOV-3/4-744A/B

a. Inspection Scope

b. Findings

.2.7 Basket Strainer and Isolation Valves, BS-3/4-1402

a. Inspection Scope

b. Findings

.2.8 POV-3/4-4882/4883

a. Inspection Scope

b. Findings

.2.9 RHR HX CCW Outlet MOV, MOV-3/4-749A/B

a. Inspection Scope

b. Findings

.2.10 Hand Control Valve, HCV-3/4-758

a. Inspection Scope

b. Findings

.2.11 AFW Min Flow Path Including Restricting Orifice for AFW Pump Recirc to Condensate

a. Inspection Scope

b. Findings

.2.12 AMSAC/ATWS Circuitry

a. Inspection Scope

b. Findings

.2.13 RHR Pump, 3/4P210A/B, Motor

a. Inspection Scope

b. Findings

.2.14 Bypass Flow Control Valve, FCV-3/4-605

a. Inspection Scope

b. Findings

.2.15 Condensate Storage Tank, 3/4T8

a. Inspection Scope

b. Findings

.2.16 125VDC Battery

a. Inspection Scope

b. Findings

.2.17 Accumulator Inclusive of Flow Path, 3/4T229A/B/C

a. Inspection Scope

b. Findings

.2.18 Emergency Diesel Generator 3A/3B Engine Cooling System

a. Inspection Scope

b. Findings

.2.19 Start-up Transformers, X03/4

a. Inspection Scope

b. Findings

.2.20 RHR Pump Recirculation to RWST MOV, MOV-3/4-863A/B

a. Inspection Scope

b. Findings

.2.21 RHR Sump Suction to Sump MOV, MOV-3/4-860A/B and 861A/B

a. Inspection Scope

b. Findings

.3 Review of Low Margin Operator Actions

a. Inspection Scope

b. Findings