IR 05000389/2014008

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IR 05000389-14-008; on 03/10/2014 - 04/28/2014; Saint Lucie Plant Unit 2; Inservice Inspection Activities and Plant Modifications
ML14122A091
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 05/02/2014
From: Omar Lopez-Santiago
NRC/RGN-II/DRS/EB3
To: Nazar M
Florida Power & Light Co
References
IR-14-008
Download: ML14122A091 (17)
v • d • e


Text

UNITED STATES May 2, 2014

SUBJECT:

ST. LUCIE PLANT UNIT 2 - NRC INSPECTION REPORT 05000389/2014008

Dear Mr. Nazar:

On May 1, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection of inservice inspection activities, including steam generator tube examinations, at your Saint Lucie Plant Unit 2. On March 21, 2014, the NRC inspectors discussed the preliminary results of this inspection with Mr. Joseph Jensen, Site Vice-President, and other members of your staff in an exit meeting conducted on-site. On May 1, 2014, the inspectors held a conference call with Mr.

Eric Katzman, Licensing Manager, and other members of your staff to discuss the final results of this inspection as documented in the enclosed inspection report.

No NRC-identified or self-revealing findings were identified during this inspection. However, the inspectors documented a licensee-identified violation which was determined to be of very low safety significance. 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 the Saint Lucie Plant.

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 and its enclosure will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents 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 Omar López-Santiago, Chief (Acting)

Division of Reactor Safety Engineering Branch 3 Docket No.: 50-389 License No.: NPF-16

Enclosure:

Inspection Report 05000389/2014008 w/Attachment: Supplementary Information

REGION II==

Docket No: 05000389 License No: NPF-16 Report No: 05000389/2014008 Licensee: Florida Power & Light Company (FPL)

Facility: Saint Lucie Plant, Unit 2 Location: 6501 South Ocean Drive Jensen Beach, FL 34957 Dates: March 10 - 21, 2014 (on-site inspection)

March 24 - April 28, 2014 (in-office review)

Inspection Team: J. Rivera-Ortiz, Senior Reactor Inspector (Sections 1R08 & 1R18)

A. Butcavage, Reactor Inspector (Section 1R08)

Approved by: Omar López-Santiago, Chief (Acting)

Division of Reactor Safety Engineering Branch 3 Enclosure

SUMMARY

IR 05000389/2014008; 03/10/2014 - 04/28/2014; Saint Lucie Plant Unit 2; Inservice Inspection

Activities and Plant Modifications The report covers an inspection conducted by two Region II inspectors in accordance with Nuclear Regulatory Commission (NRC) Inspection Procedure 71111.08, Inservice Inspection, dated January 1, 2012, and Inspection Procedure 71111.18, Plant Modifications, dated January 1, 2011. There were no NRC-identified or self-revealing findings. The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White,

Yellow, Red) and determined using IMC 0609, Significance Determination Process, dated June 2, 2011. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4.

A violation of very low safety significance that was identified by the licensee has been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and corrective action tracking number is listed in Section 4OA7 of this report.

REPORT DETAILS

1R08 Inservice Inspection Activities

Main article: IP 71111.08

a. Inspection Scope

Non-Destructive Examination and Welding Activities: The inspectors conducted an on-site review of the implementation of the licensees Inservice Inspection (ISI) Program for monitoring degradation of the reactor coolant system, emergency feedwater systems, risk-significant piping and components, and containment systems during the Unit 2 End-of-Cycle 20 refueling outage (March 2014). The inspectors activities included a review of the non-destructive examinations (NDEs) listed below to evaluate compliance with the applicable edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC),Section XI and to verify that indications and defects (if present) were appropriately evaluated and dispositioned in accordance with the requirements of the ASME Code acceptance standards or an NRC-approved alternative requirement. The inspectors activities included record reviews and direct observation of in-progress NDEs as specified below. The Code of Record for Saint Lucie Plant, Unit 2, was the 2007 Edition with 2008 Addenda of the ASME BPVC,Section XI.

  • Visual examination (VT-3) of spring support SI-2407-3000, ASME Code Class 2 (direct observation)
  • Visual examination (VT-1) of pressurizer manway stud and nut, ASME Code Class 1 (direct observation)

The inspectors reviewed the welding activity referenced below and reviewed associated documents in order to verify compliance with licensee procedures and the ASME Code.

The inspectors reviewed the work order, weld data sheets, welding procedures, procedure qualification records, welder performance qualification records, and NDE reports.

  • Work Order 40196531, Valve 3811-Replace Valve, ASME Code Class 1 During non-destructive surface and volumetric examinations performed since the previous refueling outage, the licensee did not identify any relevant indications that were analytically evaluated and accepted for continued service. Therefore, no NRC review was completed for this inspection procedure attribute.

Pressurized Water Reactor Vessel Upper Head Penetration Inspection Activities: The inspectors reviewed the licensees schedule of NDEs for the Unit 2 reactor vessel upper head penetrations (VUHPs) to verify compliance with the requirements in ASME Code Case N-729-1 as modified by 10 CFR 50.55a. For the Unit 2 End-of-Cycle 20 refueling outage, the inspectors did not identify a requirement to conduct bare metal visual (BMV)or volumetric/surface examinations of the VUHPs. Therefore, no NRC review was completed for these inspection procedure attributes.

However, the inspectors reviewed the inspection results for a follow-up examination of in-core instrumentation (ICI) penetrations 95, 97, 98, 100, and 101 associated with bolted connection leakage identified in a previous outage. The licensee conducted this examination as a corrective action for Action Request (AR) 1801708 to monitor and correct any leakage from components above the VUHPs. The inspectors reviewed the NDE summary report (Report Number 010698) which documented the examination of the ICI nozzle areas and attached photographs to determine if the examination activities and disposition of results were conducted in accordance with the standards similar to ASME Code Case N-729-1 and 10 CFR 50.55a(g)(6)(ii)(D).

Boric Acid Corrosion Control Inspection Activities: The inspectors reviewed the licensees Boric Acid Corrosion Control (BACC) program activities to ensure implementation with commitments made in response to NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable industry guidance documents. Specifically, the inspectors performed an on-site record review of procedures and the results of the licensees containment walk-down inspections performed during the current End-of-Cycle 20 refueling outage and one specific AR from the last outage (AR 01872729) related to leakage from valve 3811 in the vicinity of a reactor coolant pump. The inspectors also interviewed the BACC program owner, conducted an independent walk-down of containment to evaluate compliance with licensees BACC program requirements, and verified that degraded or non-conforming conditions, such as boric acid leaks, were properly identified and corrected in accordance with the licensees BACC and corrective action programs.

The inspectors reviewed the following ARs and engineering evaluations completed for evidence of boric acid leakage to determine if degraded components were documented in the corrective action program. The inspectors also evaluated corrective actions for any degraded components to determine if they met the ASME Section XI Code and/or and NRC-approved alternative.

  • AR 01872729, Leakage at Valve 3811, Pipe Cap, Engineering Disposition
  • AR 01878687, Leak at Valve 3598 Packing, Engineering Disposition Steam Generator Tube Inspection Activities: The inspectors reviewed the eddy current examination activities performed in Unit 2 steam generators SG-2A and SG-2B during the End-of-Cycle 20 refueling outage to verify compliance with the licensees Technical Specifications, ASME BPVC Section XI, and Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines. The inspectors interviewed licensee personnel and vendor staff responsible for the steam generator inspection project, and reviewed documentation associated with the steam generator inspections and integrity assessments as described in this report section.

The inspectors reviewed the scope of the eddy current examinations to verify that known and potential areas of tube degradation were inspected. The inspectors also verified that inspection scope expansion criteria were implemented based on inspection results as directed by the Electric Power Research Institute (EPRI) Pressurized Water Reactor Steam Generator Examination Guidelines, Revision 7.

The inspectors reviewed documentation for a sample of eddy current data analysts, eddy current probes, and eddy current testers to verify that personnel and equipment were qualified to detect the existing and potential degradation mechanisms applicable to Saint Lucie Unit 2 steam generator tubes in accordance with the EPRI Examination Guidelines. This review included a sample of site-specific Examination Technique Specification Sheets (ETSSs) that were selected based on plant-specific and industry operating experience to ensure that their qualification and site-specific implementation were consistent with Appendix H or I of the EPRI Examination Guidelines. The selected ETSSs for review consisted of bobbin and array probe techniques that were used to detect wear at the tube interface with support structures (i.e. tube support plates, anti-vibration bar (AVB), and V-shaped support pads), wear associated with potential loose parts, and tube-to-tube wear.

The inspectors also reviewed a sample of eddy current data with a qualified data analyst to confirm that data analysis was performed in accordance with the applicable ETSSs and site-specific analysis guidelines. The inspectors verified that the equipment configuration was consistent with the essential parameters of the applicable technique.

The inspectors also verified that recordable indications were detected and sized in accordance with vendor procedures. As part of the eddy current data review, the inspectors verified that the eddy current indications on each selected tube were consistent with historical data relative to the number of indications, location, and size.

The sample of eddy current data selected for review is listed below:

Steam Generator Tube Eddy Current Indication Type Row/Column Probe 2A R106/C89 Bobbin AVB Wear 2A R78/C105 Bobbin AVB Wear 2A R133/C98 Array (X-Probe) Potential Loose Parts 2B R108/C85 Bobbin AVB Wear The inspectors selected a sample of degradation mechanisms (i.e. wear at tube support plates, anti-vibration bar, and V-shaped support pads) from the Unit 2 Steam Generator Degradation Assessment and verified that the in-situ pressure testing criteria were determined in accordance with the EPRI Tube Integrity Guidelines. Additionally, the inspectors reviewed eddy current indication reports to determine whether tubes with relevant indications were appropriately screened for in-situ pressure testing.

The inspectors compared the recent eddy current examination results with the last Condition Monitoring and Operational Assessment report for Unit 2 steam generators to assess the licensees prediction capability for maximum tube degradation and number of tubes with indications. The inspectors verified that the licensees evaluation was conservative and that current examination results were bound by the Operational Assessment projections.

The inspectors also compared past examination results discussed in the latest Degradation Assessment with the recent eddy current examination results to verify that new degradation mechanisms, if any, were identified and evaluated before plant startup.

The review of eddy current examination results included the disposition of potential loose part indications on the steam generator secondary side to verify that corrective actions for evaluating and retrieving loose parts were consistent with the EPRI Guidelines. The inspectors also reviewed a sample of primary-to-secondary leakage data for Unit 2 to confirm that operational leakage in both steam generators remained below the detection or action level threshold during the previous operating cycle.

The inspectors review included the implementation of tube repair criteria and repair methods to verify they were consistent with plant Technical Specifications and industry guidelines. The inspectors compared the final list of eddy current examination indications with the final list of tubes plugged to verify that the licensee had selected the appropriate tubes for plugging based on the Technical Specification plugging criteria.

The inspectors also reviewed the tube plugging procedure and plugging results for 12 of the 63 tubes plugged in SG-2A and for the 6 tubes plugged in SG-2B to determine if the licensee installed the tube plugs in accordance with the applicable procedures.

Based on the review of the final eddy current examination results for both steam generators and interviews with the licensee, the inspectors confirmed that:

  • no new degradation mechanisms were identified
  • no eddy current scope expansion was required
  • none of the SG tubes examined met the criteria for in-situ pressure testing
  • none of the indications left in-service required repair Furthermore, the inspectors interviewed licensee staff and reviewed a sample of inspection results for the inspection conducted in the Unit 2 steam generators secondary side internals, to verify that potential areas of degradation based on site-specific operating experience were inspected, and appropriate corrective actions were taken to address degradation indications. This review included the results of Foreign Object Search and Retrieval (FOSAR) activities in both steam generators and an evaluation for a potential loose part in the SG-2B secondary side.

The inspectors review of secondary side activities included the results of visual inspections for the SG-2A and SG-2B feedwater distribution rings (also known as feedrings). The inspectors reviewed corrective actions taken to address damage of the feedrings and their supports as a result of a thermal-hydraulic transient attributed to leakage through the feedring bolted inspection covers. The extent of the damage included distortion of support components, failure of a single welded part, and localized deformation of the feedring piping and inspection covers bolting. The inspectors reviewed the licensees operability evaluation of the as-found condition to verify that the affected safety-related components remained capable of performing their design function based on the calculated stresses for design basis conditions and the loads imposed by the transient. The inspectors verified that the licensee had reasonable assurance of the most probable cause for the transient before returning to full power operation and interim corrective actions were adequate to support continuous operation. The inspectors also verified that the licensee initiated a root cause evaluation to confirm the preliminary cause and adequacy of corrective actions. The inspectors also confirmed that no damage to the tubes occurred as a result of the transient.

Additionally, the inspectors reviewed the licensees evaluation for damage to the SG-2B hot leg channel head surfaces due to a loose part in the primary side of the steam generator. The inspectors reviewed the licensees operability evaluation to verify it provided adequate technical justification to demonstrate that the steam generator remained capable of performing its design functions.

Identification and Resolution of Problems: The inspectors performed a review of a sample of ISI-related problems which were identified by the licensee and entered into the corrective action program as ARs. The inspectors reviewed the ARs to confirm the licensee had appropriately described the scope of the problem and had initiated adequate corrective actions. The scope of this AR review included corrective actions for AR 01926063 associated with the dislodgement of a thermal sleeve on the 2A2 cold leg emergency core cooling system nozzle. The ARs selected for review also included the licensees assessment of applicable operating experience information. The inspectors performed these reviews to ensure compliance with 10CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and applicable ASME Code requirements. The corrective action documents (i.e. ARs) reviewed by the inspectors are listed in the report attachment.

b. Findings

1. (Opened) Unresolved ltem (URl)05000389/2014008-001, Loose Part Damage in the Hot

Leg Channel Head of Steam Generator 2B

Introduction:

The inspectors identified an unresolved item associated with damage to the hot leg channel head of SG-2B as a result of foreign material in the reactor coolant system.

Description:

On April 8th, 2014, the licensee received indications from two channels of the loose part monitoring system during plant restart operations (Mode 3). Further evaluation of the loose part monitor signals indicated that a potential loose part was contained inside the SG-2B hot leg channel head. The licensee then commenced a cooldown and depressurization of the Unit 2 reactor to inspect the SG-2B hot leg channel head for damage and retrieve any loose object. The inspection found a single metallic (stainless steel) loose part and various indications of impact damage on the internal surfaces of the channel head. The extent of the damage included indentation marks on the channel head stainless steel cladding, tubesheet cladding, divider plate surface and welds, tube end seal welds, and existing tube plugs.

The licensee entered the issue in the corrective action program as Condition Report 01955927 and initiated a root cause evaluation. The licensee determined that the loose part was foreign material, but had not yet determined how it was introduced into the reactor coolant system. The licensee evaluated the operability of the affected components for the next operating cycle (Areva Report 51-9222481-000) and initiated corrective actions to further review potential long term effects. The operability evaluation addressed the impact damage to all the channel head surfaces with respect to the structural integrity of the materials and corrosion of carbon steel components due to potential cladding damage. The licensee also evaluated the damage to the tube seal welds and the existing tube plugs with respect to the ability of the tubes to meet the tube integrity criteria during the next operating cycle. The tube integrity evaluation included conservative postulated damage and pressure testing of a tubesheet mock-up containing tube plugs that were exposed to similar impact damage. Even though the pressure test demonstrated that the damaged tube plugs maintained their design function, the licensee conservatively decided to replace the most degraded tube plug in SG-2B. The inspectors did not identify issues of concern with the operability evaluation of SG-2B for the next operating cycle.

This is an unresolved item pending review of the licenses causal analysis of foreign material identified in the reactor coolant system. This issue will be tracked as unresolved item URI 05000389/2014008-001, Loose Part Damage in the Hot Leg Channel Head of Steam Generator 2B.

1R18 Plant Modifications

Main article: IP 71111.18

a. Inspection Scope

The inspectors reviewed a permanent plant modification of the SG-2A and SG-2B feedring supports and inspection cover (i.e. feedring end-caps) to verify that it did not affect the safety functions of important safety systems. The inspectors confirmed the modification did not degrade the design bases, licensing bases, and performance capability of risk significant structures, systems and components. Additionally, the inspectors evaluated whether the structural integrity of the modified supports served the functional requirements under design conditions, and the Code and safety classification was consistent with the design bases.

The inspectors also verified that the necessary Technical Specification changes had been identified and NRC approvals, if required, were obtained prior to modification implementation. Specifically, the inspectors verified the licensees conclusions in the screening evaluation as part of the implementation process for 10 CFR 50.59. In addition, the inspectors reviewed a sample of related corrective action documents to verify the licensee was identifying and correcting any deficiencies associated with the modification. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA6 Meetings

a. Exit Meeting On March 21, 2014, the NRC presented the preliminary results of this inspection to Mr. Joseph Jensen, Site Vice-President, and other members of the licensees staff in an exit meeting conducted on-site. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

On May 1, 2014, the inspectors held a conference call with Mr. Eric Katzman, Licensing Manager, and other members of the licensees staff to discuss the final disposition of inspection issues as presented in this report.

4OA7 Licensee-Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of an NRC requirement which met the criteria of the NRC Enforcement Policy for being dispositioned as a Non-Cited Violation.

  • A licensee-identified violation of 10 CFR 50, Appendix B, Criterion III, was discovered in March 2014, during a planned visual inspection of the upper internals in SG-2A and SG-2B. The licensee identified damage to the feedring in both steam generators as a result of a thermal-hydraulic transient. The transient caused an increase in feedring pressure, damage to the feedring supports, and increased stresses on the feedring piping and feedwater nozzle. The transient was attributed to leakage from the feedring bolted inspection covers, which were modified in 2011 through the engineering change process. 10 CFR 50, Appendix B, Criterion III, Design Control, states that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, on January 18, 2011, the licensee failed to establish design control measures that provided for verifying or checking the adequacy of the modified feedring inspection cover design to maintain the safety-related function of the steam generators. The modification replaced the feedring inspection covers, the gaskets, and the bolts lock washers in SG-2A and SG-2B.

However, in March 2014, the modification was determined to be inadequate, resulting in leakage through the bolted inspection covers and consequently in a thermal-hydraulic transient that affected the safety-related portion of the steam generators.

This violation was determined to be of very low safety significance (Green) because SG-2A and SG-2B maintained their operability and functionality, and therefore resulted in no loss of safety function. The licensee entered this violation in the corrective action program as AR 01951667. (Section 1R08)

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Fox, Site - Section XI, Repair Replacement Program Owner
R. Gil, Steam Generator Program Manager
D. Griffin, Site - Boric Acid Program Owner
J. Jensen, Site Vice-President
E. Katzman, Licensing Manager
E. Korkowski, Steam Generator Program Owner
D. Nowakowski, ISI Program Owner
R. Sciscente, Principal Engineer, Licensing

LIST OF REPORT ITEMS

Opened

05000389/2014008-001 URI Loose Part Damage in the Hot Leg Channel Head of Steam Generator 2B (Section 1R08)

LIST OF DOCUMENTS REVIEWED

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