IR 05000250/2014007

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IR 05000250-14-007, 05000251-14-007; 01/27/2014 - 02/28/2014; Turkey Point Nuclear Generating Units 3 and 4; Component Design Bases Inspection
ML14098A115
Person / Time
Site: Turkey Point  NextEra Energy icon.png
Issue date: 04/07/2014
From: Nease R
NRC/RGN-II/DRS/EB1
To: Nazar M
Florida Power & Light Co
References
IR-14-007
Download: ML14098A115 (27)
v • d • e


Text

April 7, 2014

SUBJECT:

TURKEY POINT NUCLEAR GENERATING UNITS 3 AND 4 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000250/2014007 AND 05000251/2014007

Dear Mr. Nazar:

On February 28, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Turkey Point Nuclear Generating Units 3 and 4 and discussed the results of this inspection with Mr. Mike Kiley and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented one finding of very low safety significance (Green) in this report.

This finding involved a violation of NRC requirements. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violation or significance of this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at Turkey Point Nuclear Generating Units 3 and 4.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II; and the NRC resident inspector at Turkey Point Nuclear Generating Units 3 and 4. In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Document Access and Management 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

Rebecca L. Nease, Chief

Engineering Branch 1

Division of Reactor Safety

Docket Nos.: 50-250, 50-251 License Nos.: DPR-31, DPR-41

Enclosure:

Inspection Report 05000250/2014007, 05000251/2014007 w/ Attachment: Supplemental Information

REGION II==

Docket Nos.:

50-250, 50-251

License Nos.:

DPR-31, DPR-41

Report Nos.:

05000250/2014007, 05000251/2014007

Licensee:

Florida Power & Light Company (FP&L)

Facility:

Turkey Point Nuclear Generating Units 3 & 4

Location:

9760 S. W. 344th Street Homestead, FL 33035

Dates:

January 27, 2014 - February 28, 2014

Inspectors:

Shakur Walker, Senior Reactor Inspector (Lead)

Tonya Lighty, Reactor Projects Engineer

Delza Mas, Reactor Inspector

Robert Patterson, Reactor Inspector

Jigar Patel, Accompanying Personnel Craig Baron, Contractor (Mechanical)

George Morris, Contractor (Electrical)

Approved by:

Rebecca Nease, Branch Chief Engineering Branch 1 Division of Reactor Safety

SUMMARY

IR 05000250/2014007, 05000251/2014007; 01/27/2014 - 02/28/2014; Turkey Point Nuclear

Generating Units 3 and 4; Component Design Bases Inspection.

This inspection was conducted by a team of five Nuclear Regulatory Commission (NRC)inspectors from Region II, and two NRC contract personnel. One Green non-cited violation (NCV) was identified. The significance of inspection findings is indicated by their color (Green,

White, Yellow, Red) using the NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Components Within the Cross Cutting Areas, dated October 28, 2011. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated January 28, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 4, dated December 200

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Procedures and Programs, for the licensees failure to implement procedure 0-ADM-232,

Time Critical Action Program, to ensure time critical actions (TCAs) important to mitigate design basis events could be performed in the required time. The failure to implement this procedure was a performance deficiency. No documentation existed to demonstrate that the TCA to restore power to the battery chargers during a station blackout could be performed within the required time (30 minutes). The team also identified a TCA to locally isolate the auxiliary feedwater for a faulted steam generator that did not have a job performance measure to demonstrate the successful completion of the action. The licensee entered this issue into the corrective action program as action requests 01944453, 01945532, 01943321, 01943425, and 01943697. For TCAs where no validation documentation could be determined, the licensee completed tabletop exercises, simulator exercises, and field walkdowns to ensure that all of the TCAs to mitigate design basis events could be completed within the required action times.

The performance deficiency was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Specifically, the licensee did not implement 0-ADM-232 adequately to ensure that the TCAs listed in Attachment 1 of the procedure were properly validated; consequently, the licensee could not demonstrate that TCAs could be successfully executed in accordance with the design basis. The team determined the finding to be of very low safety significance (Green)because the finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; and did not represent a loss of system and/or function. The team determined this finding was associated with the cross-cutting aspect of Procedure Adherence in the area of Human Performance because although the procedure was recently revised to include all necessary requirements to maintain the time critical action program, the licensee failed to follow procedure 0-ADM-232, which resulted in several TCAs not being properly validated. [H.8]

(Section 1R21.2)

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 related operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1E-6. The sample included 16 components, one of which was associated with containment large early release frequency (LERF), and six operating experience (OE) items.

The team performed a margin assessment and a detailed review of the selected risk-significant components and operator actions to verify that the design bases had been correctly implemented and maintained. Where possible, this margin was determined by the review of the design basis and Updated Final Safety Analysis Report (UFSAR)response times associated with operator actions. This margin assessment also considered original design issues, margin reductions due to modifications, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for a detailed review. These reliability issues included items related to failed performance test results, significant corrective action, repeated maintenance, maintenance rule status, Regulatory Issue Summary 05-020 (formerly Generic Letter 91-18) conditions, NRC resident inspector input regarding problem equipment, system health reports, industry OE, and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, OE, 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 Component Reviews

a. Inspection Scope

Components

  • Security diesel generator, the security battery, and the security battery charger
  • 125 VDC safety-related batteries 3A, 3B, 4A, and 4B
  • 125 VDC safety-related battery chargers
  • 120 VAC vital instrument power inverters and panels
  • 480VAC load centers 3A, 3B, 3B, 3C, 3D, 3H, 4A, 4B, 4C and 480 V motor control centers - 3B, 3C, 4B, 4C
  • AFW injection valves to the steam generators - CV-2816/17/18 or CV-2831/32/33 - condensate storage tank (CST) level instruments
  • Refueling water storage tank (RWST) and associated level instruments
  • DC bus 3D23
  • Pressurizer auxiliary spray valve - CV-3-311
  • Power transformers - main generator transformers MT 3X01 and 4X01; startup transformers SUT 3X01 and 4X0; unit auxiliary transformers UAT 3X01 and 4X01

Components with LERF Implications

  • Pressurizer power-operated relief valves - PCV-455C, PCV-456

For the 16 components listed above, the team reviewed the plant technical specifications (TS), UFSAR, design bases documents (DBDs), and drawings to establish an overall understanding of the design bases of the components. Design calculations and procedures were reviewed to verify that the design and licensing bases had been appropriately translated into these documents. Test procedures and recent test results were reviewed against DBDs to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents, and that individual tests and analyses served to validate component operation under accident conditions.

Maintenance procedures were reviewed to ensure components were appropriately included in the licensees preventive maintenance program. System modifications, vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action program documents were reviewed (as applicable) in order to verify that the performance capability of the component was not negatively impacted, and that potential degradation was monitored or prevented. Maintenance Rule information was reviewed to verify that the component was properly scoped, and that appropriate preventive maintenance was being performed to justify current Maintenance Rule status. Component walkdowns and interviews were conducted to verify that the installed configurations would support their design and licensing bases functions under accident conditions and had been maintained to be consistent with design assumptions.

Additionally, the team performed the following component-specific reviews:

  • The team reviewed the calibration of loop elements and the bases for the associated comparator setpoints in the operating and alarm procedures, for the pressure and level instruments associates with the RWST, CST, and the reactor coolant pressure control system.
  • The team observed the control room indicators and alarms for the AC power systems.
  • The team reviewed major modifications for the power transformers related to the EPU: EC 242520, 242521, 242435, and 242440.
  • The team reviewed the battery room hydrogen dilution calculation to verify that the hydrogen concentration would stay below flammable limits during normal and postulated accident conditions.
  • The team reviewed the capability of the security inverter and associate instrument panel to provide instrumentation and control power during all conditions, but particularly during accident conditions.
  • The calculation for loading and voltage drop was reviewed for the security inverter and associate instrument panel to ensure that sufficient capacity exists for all normal and accident loading, and that sufficient voltage was available for all loads.
  • The security battery loading study was reviewed to verify that the battery was capable of providing the appropriate voltage for the inverter during one hour.
  • The team reviewed security diesel generator protective relaying schemes and calculations to determine that the security equipment was adequately protected, and to determine whether protective devices featured proper selective tripping coordination.
  • The team reviewed the degraded voltage protection scheme to determine whether the voltage setpoints were selected based on the voltage requirements for security loads.
  • The team reviewed the design and testing of the material and coating of the RWST.
  • The team reviewed the design basis analyses associated with the capacity of the nitrogen bottles to operate the AFW flow control valves for the required time and the capability of the operators to replace the exhausted bottles when required.
  • The team reviewed the design and testing of the control circuit interlocks between the AFW steam admission valves and the AFW flow control valves to verify that these circuits would perform the required functions assuming a single component failure and that the required circuits were fully tested on a periodic basis.
  • The team reviewed the capability of the operators to locally isolate AFW flow to a faulted steam generator in the event of the single failure of an AFW flow control valve to close from the control room. This component failure would require operators to identify the condition and close a manual valve within 10 minutes.
  • The team reviewed the TS requirements associated with taking a single AFW steam admission valve out of service. Specifically, the team investigated whether the AFW system would be capable of performing its required function with a faulted steam generator, the single failure of a steam admission valve, and one valve out of service.
  • The team reviewed the basis of the reactor coolant system pressure interlocks associated with the motor operated RHR supply valves.
  • The team reviewed the design and testing of the control circuit interlocks associated with the RHR supply valves to verify that these circuits would perform the required functions assuming a single component failure and that the required circuits were fully tested on a periodic basis.
  • The team reviewed the TS requirements associated with operability of RHR and SI system components during Mode 4 operation. Specifically, the team investigated whether the RHR system would be capable of mitigating a postulated LOCA with only the minimum TS equipment available to the operators.
  • The team reviewed the design pressures and overpressure protection of piping associated with interfaces between high and low pressure portions of the RHR system. Specifically, the team reviewed the design of the outboard piping to verify that it would not be over-pressurized due to valve leakage or the inadvertent opening of a single valve.
  • The team reviewed the design basis analyses and supporting calculations for containment heat up, to verify that, during a design basis accident with a dual unit loss of offsite power and a LOCA on one unit, the non-accident unit containment temperatures would not affect safety-related instrumentation operability.
  • The team reviewed the digital feedwater regulator valve and bypass valve modifications to ensure no additional initiating events were introduced as a result of the digital upgrades.
  • The team observed simulator scenarios involving station blackout (SBO) time critical operator actions: 1) manually align safety related battery chargers to the SBO diesel supplied bus to verify the actions could be accomplished in 30 minutes from the start of the event; and 2) an action to manually align the SBO diesel to supply safety loads to both units from the control room within 10 minutes.
  • The team reviewed Extended Power Uprate (EPU) calculations for flow-induced vibrations to ensure they correctly modeled plant configuration and verified that any errors introduced by components that were not correctly modeled, were bounded by the available margin.
  • The team reviewed different testing methodologies for evaluating time critical operator scenarios, specifically use of video when evaluating operator crews, to ensure the consistency of time validation of operator actions during the scenarios.
  • The team reviewed main feedwater line supports, deflector shields, and whip restraint modifications to ensure compliance with high-energy line breaks after the EPU.
  • The team reviewed the emergency operating procedures with respect to high head safety injection pumps during main steam line break inside containment, steam generator tube rupture, small break LOCA, or an inadvertent SI actuation to verify if guidance exists to shut the pumps off within the required time to protect the pump.
  • The team the reviewed the loss of all feedwater scenario to ensure adequate guidance for feed and bleed operation via SI and the pressure operated valves were adequate and were incorporated into the licensed operator continuous training program.
  • The team reviewed completed surveillances for steam generator level and flow transmitters to verify they were updated to incorporate the new setpoints from the EPU.
  • The team reviewed the flooding analysis to ensure that RWST level instrumentation would remain functional during design basis flooding event.
  • The team reviewed the sites set point methodology for pressurizer pressure instrumentation to ensure compliance with the sites current licensing basis.

b. Findings

Introduction:

The team identified a Green non-cited violation (NCV) of TS 6.8.1, Procedures and Programs, for the licensees failure to implement procedure 0-ADM-232, Time Critical Action Program, adequately to ensure time critical actions (TCAs) important to mitigate design basis events could be performed in the required time.

Description:

During an NRC observation of a requested simulator exercise for a SBO crosstie event on Unit 3 and Unit 4, and verifying the safety related battery chargers were manually loaded to a powered EDG within 30 minutes from the start of the event, the licensee did not meet the 30-minute time requirement for the battery chargers. The team requested documentation from previous performances of the TCA to ensure the 30-minute requirement could be met. The licensee could not provide any documentation that the TCA had been previously performed within the required time of 30 minutes.

In addition, the team determined that a job performance measure (JPM) for locally isolating AFW flow to the faulted steam generator during a main steam line break event inside containment within 10 minutes of event initiation, had not been created or validated. Isolation of AFW would usually be performed from the control room using an air-operated flow control valve; however, the emergency operation procedures 3/4-EOP-E-0 and 3/4-EOP-E-2 directed operators to manually isolate AFW flow if isolation from the control room is unsuccessful. This manual isolation would be performed using a manual isolation valve in series with the failed flow control valve.

Procedure 0-ADM-232 requires TCAs listed in Attachment 1, that are also credited in the licensees design basis, be validated to ensure they can be performed within their required action times. The licensee could not provide documentation that the TCA for the battery chargers could be completed in the required timeframe. In addition, no JPM had been created for closing a manual isolation valve to the faulted steam generator within 10 minutes from the start of a main steam line break inside containment.

The licensee entered each issue into the corrective action program as action requests 01944453, 01945532, 01943321, 01943425, and 01943697 and completed tabletop exercises, simulator exercises, and field walkdowns to ensure that all of the TCAs listed in Attachment 1 could be completed within the required action times. The team verified appropriate corrective actions were taken if any deficiencies were identified during the review of Attachment 1 TCAs.

Analysis:

The team determined that failure to implement procedure 0-ADM-232 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not implement 0-ADM-232 adequately to ensure that the TCAs listed in Attachment 1 of the procedure were properly validated and thus could not demonstrate that TCAs could be successfully executed in accordance with the design basis. The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; and did not represent a loss of system and/or function.

The team determined this finding was associated with the cross-cutting aspect of Procedure Adherence in the area of Human Performance because although the procedure was recently revised to include all necessary requirements to maintain the time critical action program, the licensee failed to follow procedure 0-ADM-232, which resulted in several TCAs not being properly validated. [H.8]

Enforcement:

Technical Specification 6.8.1 requires that procedures required by the Florida Power and Light Quality Assurance Topical Report be implemented and maintained. The Quality Assurance Topical Report includes procedures listed in Appendix A of NRC Regulatory Guide 1.33, Revision 2, dated February 1978, which lists procedures for authorities and responsibilities for safe operation and shutdown.

Procedure 0-ADM-232 is a quality related procedure that would ensure safe operation of the plant by ensuring time critical operator actions could be performed in the required action times to mitigate design basis events. Contrary to the above, procedure 0-ADM-232, Time Critical Operator Action Program, was not being properly implemented.

Specifically, multiple procedural requirements were not being met, such as record retention, TCA validation, maintaining adequate emergency operating procedures, and development of JPMs and scenarios to address all TCAs. This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. This violation was entered into the licensees corrective action program as action requests 01944453, 01945532, 01943321, 01943425, and 01943697. (NCV 05000250,251/2014007-01, Failure to Properly Implement Time Critical Operator Action Program Procedure)

.3 Operating Experience

a. Inspection Scope

The team reviewed six operating experience issues for applicability at Turkey Point Nuclear Generating Units 3 and 4. The team performed an independent review of these issues and, where applicable, assessed the licensees evaluation and dispositioning of each item. The issues that received a detailed review by the team included:

  • NRC IN 2013-05, Battery Expected Life and Its Potential Impact on Surveillance Requirements, dated March 19, 2013
  • Regulatory Issues Summary 2005-29, Anticipated Transient That Could Develop Into More Serious Events, dated December 14, 2005
  • Instrument calibrations and surveillance testing associated with Turkey Point Extended Power Uprate

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

On February 28, 2014, the team presented the inspection results to Mr. Mike Kiley and other members of the licensees staff. The inspectors verified that no proprietary information was documented in this report.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

A. Chomat, Unit Supervisor
T. Conboy, Plant General Manager
P. Czaya, Licensing
C. Domingos, Engineering Director
J. Garcia, Engineering Director
D. Jenkins, Licensing Engineer
A. Katz, Maintenance Manager
M. Kiley, Site Vice-President
R. Smith, Engineering
B. Stamp, Training Manager
R. Tomonto, Licensing Manager

NRC personnel

D. Rich, Chief, Projects Branch 3, Division of Reactor Projects
T. Hoeg, Senior Resident Inspector, Division of Reactor Projects, Turkey Point Resident Office
M. Endress, Resident Inspector, Division of Reactor Projects, Turkey Point Resident Office
J. Hanna, Senior Reactor Analyst, Division of Reactor Projects

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened and Closed

05000250, 251/2014007-01 NCV Failure to Properly Implement Time Critical Operator Action Program Procedure [Section 1R21.2]

LIST OF DOCUMENTS REVIEWED

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