IR 05000390/2012008

IR 05000390-12-008, 01/30/2012 - 05/15/2012, Watts Bar Nuclear Plant, Unit 1, Component Design Bases Inspection

ML12165A186
Person / Time
Site:	Watts Bar
Issue date:	06/13/2012
From:	Nease R L NRC/RGN-II/DRS/EB1
To:	Shea J W Tennessee Valley Authority
Linda K. Gruhler 404-997-4633
References
IR-12-008
Download: ML12165A186 (42)

Text

June 13, 2012

Mr. JosephManager, Corp. Nuclear Licensing Programs Tennessee Valley Authority 1101 Market Street, LP 4B-C Chattanooga, TN 37402-2801

SUBJECT: WATTS BAR NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000390/2012008

Dear Mr. Shea:

On, May 15, 2012, U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant, Unit 1. The enclosed inspection report documents the inspection results, which were discussed on May 15, 2012, with Mr. David Gronek and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your licenses. The team reviewed selected procedures and records, observed activities, and interviewed personnel. This report documents three NRC identified findings of very low safety significance (Green), which were determined to involve violations of NRC requirements. The NRC is treating these violations as non-cited violations consistent the NRC Enforcement Policy. If you 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 Watts Bar. Further, if you disagree with the cross-cutting aspect assigned to any finding 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 Watts Bar. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 10 CFR 2.390 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

2 NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document 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 Nease, Chief Engineering Branch 1 Division of Reactor Safety Docket No. 50-390 License No. NPF-90

Enclosure:

Inspection Report 05000390/2012008,

w/Attachment:

Supplemental Information cc w/encl: (See page 3)

2 NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document 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 Nease, Chief Engineering Branch 1 Division of Reactor Safety Docket No. 50-390 License No. NPF-90

Enclosure:

Inspection Report 05000390/2012008,

w/Attachment:

Supplemental Information cc w/encl: (See page 3) Distribution w/encl: J. Hamman, RII, PE, DRP B. Monk, SRI K. Miller, RI C. Evans, RII EICS (Part 72 Only) L. Douglas, RII EICS (Linda Douglas)

OE Mail (email address if applicable) RIDSNRRDIRS PUBLIC RidsNrrPMWattsBar1 Resource RidsNrrPMWattsBar2 Resource X PUBLICLY AVAILABLE G NON-PUBLICLY AVAILABLE G SENSITIVE X NON-SENSITIVE ADAMS: X Yes ACCESSION NUMBER: ML12165A186 X SUNSI REVIEW COMPLETE X FORM 665 ATTACHED OFFICE RII:DRS RII:DRS RII:DRS RII:DCP RII:DRS SIGNATURE PCH1 JAE1 DLM4 WRL via e-mail JLD2 NAME P. HIGGINS J. EARGLE MAS-PENARANDAR. LEWIS J. DYMEK DATE 5/16/2012 5/17/2012 5/16/2012 5/16/2012 5/16/2012 E-MAIL COPY YES NO YES NO YES NO YES NO YES NO OFFICE CONTRACTOR CONTRACTOR RII:DRS RII:DRP RII:DRS SIGNATURE HDC2 via e-mail GBS RLN1 SMS SRS5 NAME H. CAMPBELL G. SKINNER R. NEASE S. SHAEFFER S. SANDAL DATE 5/18/2012 5/17/2012 6/12/2012 6/12/2012 5/16/2012 E-MAIL COPY YES NO YES NO YES NO YES NO YES NO OFFICIAL RECORD COPY DOCUMENT NAME: S:\DRS\ENG BRANCH 1\BRANCH INSPECTION FILES\2011-2012-2013 CYCLE EB1 INSPECTION FOLDERS FOR ALL SITES\WATTS BAR\WATTS BAR CDBI (2012008)\WATTS BAR 2012008 CDBI IR.DOCX TVA 3 cc w/encl: D. E. Grissette Site Vice President Watts Bar Nuclear Plant Tennessee Valley Authority Electronic Mail Distribution David H. Gronek Plant Manager Watts Bar Nuclear Plant, MOB 2R-WBN Tennessee Valley Authority P.O. Box 2000 MOB 2R Spring City, TN 37381 Donna K. Guinn Manager, Licensing Watts Bar Nuclear Plant, ADM 1L-WBN Tennessee Valley Authority P.O. Box 2000 Spring City, TN 37381 Edward J. Vigluicci Assistant General Counsel Tennessee Valley Authority Electronic Mail Distribution Gordon P. Arent New Generation Licensing Manager Watts Bar Nuclear Plant, EQB 1B-WBN Tennessee Valley Authority P.O. Box 2000 Spring City, TN 37381 County Mayor P.O. Box 156 Decatur, TN 37322 County Executive 375 Church Street Suite 215 Dayton, TN 37321 Tennessee Department of Environment & Conservation Division of Radiological Health 401 Church Street Nashville, TN 37243 Senior Resident Inspector U.S. Nuclear Regulatory Commission Watts Bar Nuclear Plant U.S. Nuclear Regulatory Commission 1260 Nuclear Plant Road Spring City, TN 37381-2000 Ann Harris 341 Swing Loop Rockwood, TN 37854 Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION II

Docket Nos: 050000390 License Nos: NPF-90 Report Nos: 05000390/2012008 Licensee: Tennessee Valley Authority (TVA)

Facility: Watts Bar Nuclear Plant, Unit 1 Location: Spring City, TN 37381 Dates: January 30 - May 15, 2012 Inspectors: P. Higgins, Senior Reactor Inspector (Lead) J. Eargle, Reactor Inspector D. Mas-Penaranda, Reactor Inspector R. Lewis, Resident Inspector J. Dymek, Reactor Inspector H. Campbell, Accompanying Personnel G. Skinner, Accompanying Personnel Approved by: Rebecca Nease, Chief Engineering Branch 1 Division of Reactor Safety

SUMMARY OF FINDINGS

IR 05000390/2 2-008; 01/30/2 2 - 05/15/2 2; Watts Bar Nuclear Plant, Unit 1; 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. Three Green non-cited violations (NCV) were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using the NRC Inspection Manual Chapter (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 non-cited violation of CFR Part 50, Appendix B, Criterion XI, "Test Control," for failure to perform capacity (volumetric flow) testing on the safety-related auxiliary control air subsystem (ACAS). The licensee had documented that, for worst case environmental conditions, the air compressor capacity had little margin when compared to required air demand, even for single unit operation. This issue was entered into the licensee's corrective action program as problem evaluation report 501941 for further evaluation of corrective actions. The team determined that the failure to perform capacity testing to ensure the ACAS would meet the required air demand in response to a design basis event was a performance deficiency. This performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to develop a test procedure that would reliably ensure that the ACAS would meet required air demand for its safety-related loads during design basis accidents. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, "Phase 1-Initial Screening and Characterization of Findings." The inspectors determined that the finding had a cross- cutting aspect in the use of conservative assumptions in the decision-making component of the human performance area. Specifically, the licensee did not use conservative assumptions in making the decision to discontinue capacity testing of the ACAS system in 2 2, and stated that if that decision had been made more recently (using available internal guidance and practices regarding the testing of safety-related systems), the resulting decision would have been the same H.1(b). [Section 1R21.2.4]

Green.

The team identified a non-cited violation of CFR Part 50, Appendix B, Criterion XI, "Test Control," for the licensee's failure to establish a test program that demonstrated the adequacy of the auxiliary feedwater (AFW) discharge check valves. Specifically, the licensee failed to develop a test program that would provide assurance that back leakage through the AFW discharge check valves would not prevent the system from providing design flowrates to the steam generators. This issue was entered into the licensee's corrective action program as problem evaluation report 499950. The licensee performed a functional evaluation and determined that the AFW system was operable based on the pumps not currently being degraded to the design limits, and the existence of additional conservatisms in the licensee's design basis hydraulic analysis. The team determined that the licensee's failure to establish a test program to ensure that back leakage through the AFW discharge check valves would not challenge the ability of the AFW system to provide design basis flow to the steam generators was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, it could have the potential to lead to a more significant safety concern. Specifically, AFW check valve back leakage could challenge the system's ability to support removal of decay heat from the reactor, which would not be identified by the licensee's test program. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green), utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, "Phase 1-Initial Screening and Characterization of Findings." Because the licensee performed a self-assessment in December 2011 that included missed opportunities to identify that check valve leakage could negatively impact the AFW system, this finding was assigned a cross-cutting aspect in the self- and independent assessments component of the problem identification and resolution area [P.3 (a)]. [Section 1R21.2.6]

Green.

The team identified five examples of a non-cited violation of CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures. The five examples were entered into the licensee's corrective action program. The inspectors determined that the failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures, was a performance deficiency. The performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding is similar to Inspection Manual Chapter 0612, Appendix E (example 4.a), because the failure to ensure correct translation of acceptance criteria into procedures was not isolated. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4,

"Phase 1-Initial Screening and Characterization of Findings." This finding had a cross-cutting aspect in the resources component of the human performance area, because the licensee had not ensured that complete, accurate, and up-to-date procedures were consistent with vendor and design specifications; and therefore, the procedures were not available and adequate to assure nuclear safety H.2(c). [Section 1R21.2.9] Licensee-Identified Violations None

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R21 Component Design Bases Inspection

.1 Inspection Sample Selection Process

The team selected risk significant components and related operator actions for review using information contained in the licensee's 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 1 X10-6. The sample included 15 components, including one associated with containment large early release frequency, and four operating experience 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 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.

.2.1 Essential Raw Cooling Water (ERCW) Discharge Check Valves (CKV

503F, -503H)

a. Inspection Scope

The team reviewed applicable portions of the plant's UFSAR, Technical Specifications (TS), system description document (SDD), and piping and instrumentation diagrams (P&IDs) to identify the design bases requirements of the ERCW check valves. The team examined system health reports, records of surveillance testing and maintenance activities, and applicable corrective actions to verify that potential degradation issues were being monitored, prevented and/or corrected. The team performed a walkdown of the ERCW system to review the installed configuration of the pumps, associated piping and check valves. Discussions with system and design engineers were conducted regarding pump and check valve design and performance requirements. A review of ERCW pump curves and inservice testing procedures was used to evaluate potential back-leakage through the subject check valves. For the following operator actions, the team performed a walkthrough of associated Abnormal Operating Instructions (AOIs), Annunciator Response Instructions (ARIs), Standard Operating Instructions (SOIs), and other applicable operations procedures (e.g. temporary instructions, etc.) with plant operators, maintenance personnel 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 to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following events/evolutions were reviewed:

• Crosstie opposite unit and train ERCW headers (1B-B strainer plugged) - [HAERCW1A]

• Operators fail to clear ERCW screens before plant trip - [DHAERCWS]

b. Findings

No findings were identified.

.2.2 Residual Heat Removal (RHR) Pump (1A-A, 1PMP

0010-A)

a. Inspection Scope

The team reviewed the UFSAR, system design criteria, current system health report, selected drawings, operating procedures, and past corrective action reports for the RHR pump. The team interviewed the RHR system engineer to discuss the overall health and condition of the RHR system and pumps. Further, the team interviewed inservice test engineers responsible for maintaining and updating the required TS surveillance procedures. Also, the team interviewed instrumentation and control engineers responsible for performing the associated instrumentation uncertainty calculations. Calculations addressing required RHR pump performance requirements during design bases accidents, calculations addressing the uncertainties of the instruments used to verify pump performance during required inservice test surveillances, quarterly and comprehensive surveillance procedures, and test results used to verify required RHR pump performance were reviewed. A sample of emergency operating procedures, which incorporated RHR pump, flow instrumentation and start logic were also reviewed. The team verified by review of schematic drawings, that operation of the pump motors was consistent with the design basis and operational requirements. The team reviewed the protection coordination calculation for the RHR pump motor and verified that the circuit breaker ratings and protective device's trip settings and alarm functions were consistent with the licensing basis and operational requirements.

The team verified that the brake horsepower required by the pump was within the motor rating. The team performed a walkdown of the RHR system to assess observable material condition of the pump motors. The team verified that the ambient conditions were consistent with vendor recommendation for the motors.

b. Findings

No findings were identified.

.2.3 Standby Diesel Generator (SDG) Heat Exchangers (2B1 and 2B2)

a. Inspection Scope

The team reviewed the plant TS, UFSAR, SDDs and associated P&IDs to establish an overall understanding of the cooling water requirements for the SDG systems.

Design calculations were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. The team also reviewed 10 CFR Appendix A to Part 50; General Design Criteria (GDC) 44 "Cooling Water", GDC 45, "Inspection of Cooling Water" and GDC 46 "Testing of Cooling Water" as well as Generic Letter 89-13 "Service Water System Problems Affecting Safety-Related Equipment."

Component walkdowns were performed to verify the equipment was installed in accordance with design drawings, system descriptions and that the system was being maintained in accordance with the design basis. The overall condition of areas containing this equipment was examined, and that cooling water flow control and throttling valves to the heat exchangers were in their correct positions and properly secured. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Heat exchanger vendor documentation was reviewed to verify assumed flow rates, temperature differential(s)and fouling factor limits were accounted for in plant technical instructions and surveillances. System health reports, preventative and corrective maintenance, inservice testing and surveillance testing was reviewed to confirm that testing is being done in accordance with manufacturers requirements. Docketed plant correspondence was reviewed to determine plant commitments to Generic Letter 89-13 for initial baseline testing and periodic re-testing of open cycle service water systems against minimum frequency intervals as specified in Generic Letter 89-13.

b. Findings

No findings were identified.

.2.4 Auxiliary Control Air Subsystem (ACAS) (A-A and B-B)

a. Inspection Scope

The team reviewed applicable portions of the plant's TS, UFSAR, SDD, and P&IDs to establish an overall understanding of the design bases of the safety-related ACAS.

Design calculations were reviewed to verify that design basis air capacity flow rate requirements had been appropriately translated into these documents. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented. Also, current surveillance testing including start-logic, and previous testing, which included compressor air capacity testing, was reviewed. Further, several discussions with both the current and former system engineers took place, with focus on maintenance practices and testing requirements and results of the ACAS. For the following operator actions, the team performed a walkthrough of associated AOIs, ARIs, SOIs, and other operations procedures (e.g. temporary instructions, etc.) with plant operators, maintenance personnel 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 to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following event/evolution were reviewed:

• Restore Auxiliary Feedwater (AFW) control following initiator and loss of air [HAFR1]

b. Findings

Introduction:

The team identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for the failure to perform capacity (volumetric flow) testing on the safety-related auxiliary control air subsystem (ACAS). The licensee had documented that for worst case environmental conditions, the compressor capacity had little margin when compared to required air demand, even for single unit operation.

Description:

The ACAS is a safety-related system required to supply air to safety-related equipment during design basis events (e.g., air is to be supplied at adequate pressure and capacity to auxiliary feedwater system level control valves and steam generator pressure control valves during postulated design basis accidents). Although the licensee's ACAS test procedures demonstrated start logic functionality and the capability to develop adequate system pressure, the test procedures did not evaluate or measure compressor air capacity. Capacity testing was previously performed on this system by the licensee, but was removed from the test program in 2002. Instead of capacity testing, the licensee relied upon the preventive maintenance program, during which various component parts were replaced, to ensure the system remained capable of operating at proper capacity. The licensee did not perform as-found capacity tests prior to such maintenance and did not perform post-maintenance capacity testing. The licensee had concluded that under design basis conditions, the ACAS capacity margin was less than 5 standard cubic feet per minute, and that this represents little margin (even for single unit operation). In response to the team's questions, this issue was entered into the licensee's corrective action program as problem evaluation report 501941.

Analysis:

The team determined that the failure to perform capacity testing to ensure the ACAS would meet the required air demand in response to a design basis event was a performance deficiency.

This performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of equipment performance to ensure the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to develop a test procedure that would reliably ensure that the ACAS would meet required air demand for its safety-related loads during design basis accidents. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of Inspection Manual Chapter 0609, Attachment 4, "Phase 1-Initial Screening and Characterization of Findings." The finding did not represent a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the use of conservative assumptions in the decision-making component of the human performance area. Specifically, the licensee did not use conservative assumptions in making the decision to discontinue capacity testing of the ACAS system in 2002, and stated that if that decision had been made more recently (using available internal guidance and practices regarding the testing of safety-related systems), the resulting decision would have been the same H.1(b).

Enforcement:

10 CFR Part 50, Appendix B, Criterion XI, "Test Control," states in part, that a program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate acceptance limits contained in applicable documents. Contrary to this requirement, since 2002, the licensee had failed to establish a test control program to assure that the ACAS would perform satisfactorily in response to a design basis event. Specifically, the licensee did not have a test procedure to verify that ACAS capacity met design basis accident performance requirements. Because this violation was of very low safety significance and was entered into the licensee's corrective action program, this violation is being treated as an NCV, consistent with the Enforcement Policy: NCV 05000390/2012008-01, "Failure to Establish Test Procedures to Assure Satisfactory ACAS Performance during Design Basis Accidents."

.2.5 RHR Pump Room Coolers (A-A and B-B)

a. Inspection Scope

The team reviewed applicable portions of the plant's TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the coolers. Design calculations were reviewed to verify that design basis heat removal requirements, capability, and flow rates had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents, and that individual tests and/or analyses served to validate component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Environmental qualification and seismic documents/calculations were reviewed to verify that the cooler was appropriately qualified. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.

b. Findings

No findings were identified.

.2.6 Auxiliary Feedwater Check Valve

(3-805)

a. Inspection Scope

The team reviewed applicable portions of the plant's TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the check valve. Design calculations (e.g., differential pressure) were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. Operating procedures were reviewed to verify that operator actions involving the high-pressure fire protection system alignment of auxiliary feedwater (AFW) were feasible. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Seismic documents/calculations were reviewed to verify that the valve was appropriately qualified. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.

b. Findings

Introduction:

The team identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for the licensee's failure to establish a test program that demonstrated the adequacy of the AFW discharge check valves. Specifically, the licensee failed to develop a test program that would provide assurance that back leakage through the AFW discharge check valves would not prevent the system from providing design flowrates to the steam generators.

Description:

The AFW system supplies feedwater to the steam generators in the event of a loss of main feedwater to remove reactor decay heat and avoid reactor coolant system over pressurization. The AFW system consists of two motor-driven pumps and one turbine-driven pump. Check valves are installed on the discharge of the pumps with design functions of (1) opening on an AFW initiation, and (2) closing on a pump failure to maintain the running pumps' flow path integrity. If a check valve or combination of check valves on a non-running pump were to leak, that flow would be re-circulated back to the condensate storage tank rather than being delivered to the steam generators.

The team determined that calculation HCGTBG091981, "Design Parameter for Motor and Turbine Driven AFW Pumps," Rev. 8, did not account for any check valve leakage losses. The team noted that at worst case allowable degradation of the pumps, little to no margin existed in meeting the system flow requirements for various system operating conditions. In this regard, under the worst case scenario of a loss of the turbine-driven AFW pump and a 5% degradation of both motor-driven AFW pumps (as accounted for in the licensee's hydraulic analysis), the allowable loss due to check valve back leakage would be zero. The team determined that licensee has no testing provisions in place to ensure that the zero leakage condition was met for the configuration described above, or other possible system alignments. This issue was entered into the licensee's corrective action program as problem evaluation report 499950. The licensee performed a functional evaluation and determined that the AFW system was operable based on the pumps' performance not currently being degraded to the design limit, and the existence of additional conservatisms in the licensee's hydraulic analysis. Additionally, the team noted that the licensee had performed a self-assessment in December 2011, to determine if check valve back leakage could challenge design basis analyses used to demonstrate the adequacy of system design. Although the licensee evaluated the essential raw cooling water system check valves for this potential vulnerability, the licensee had not applied the evaluation to other safety-related systems including AFW.

Analysis:

The team determined that the licensee's failure to establish a test program to ensure that back leakage through the AFW discharge check valves would not challenge the ability of the AFW system to provide design basis flow to the steam generators, was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, the performance deficiency could have the potential to lead to a more significant safety concern. Specifically, AFW check valve back leakage could challenge the system's ability to support removal of decay heat from the reactor and this condition would not be identified by the licensee's test program. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, "Phase 1-Initial Screening and Characterization of Findings." The finding did not represent a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Because the licensee performed a self-assessment in December 2011, that included missed opportunities to identify that check valve leakage could negatively impact the AFW system, this finding was assigned a cross-cutting aspect in the self- and independent assessments component of the problem identification and resolution area P.3(a).

Enforcement:

Appendix B to 10 CFR Part 50, Criterion XI states in part, that a test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to this requirement, since initial plant operation on February 7, 1996, the licensee failed to establish a test program that demonstrated that the AFW discharge check valves would perform satisfactorily in service. Specifically, the licensee failed to develop a test program that would provide assurance that back leakage through the AFW discharge check valves would not prevent the system from providing design flowrates to the steam generators. Because this violation was of very low safety significance and it was entered into the licensee's corrective action program, this violation is being treated as an NCV, consistent with the Enforcement Policy: NCV 05000390/2012008-02, "Failure to Adequately Test the AFW Discharge Check Valves."

.2.7 Component Cooling Water Heat Exchangers

a. Inspection Scope

The team reviewed applicable portions of the plant's TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the heat exchangers. Design calculations were reviewed to verify that design basis heat removal requirements, capability, and flow rates had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents, and that individual tests and/or analyses served to validate component operation under accident/event conditions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored, or prevented, and the component replacement was consistent with inservice/equipment qualification life. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.

For the following operator actions, the team performed a walkthrough of associated AOIs, ARIs, SOIs, and other operations procedures (e.g., temporary instructions, etc.) with plant operators, maintenance personnel, 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 to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following event/evolution were reviewed:

• Align and initiate alternate cooling to centrifugal charging pump A [HACCSR2]

b. Findings

No findings were identified.

.2.8 Auxiliary Feedwater Pressure Control Valve (3-132)

a. Inspection Scope

The team reviewed applicable portions of the plant's TS, UFSAR, SDDs, and P&IDs to establish an overall understanding of the design bases of the valve. Design calculations (i.e., differential pressure and required torque/thrust) were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. The team also reviewed instrument loop diagrams, loop setpoints and scaling documents, and loop accuracy calculations to verify that the valve controls were consistent with design bases assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.

b. Findings

No findings were identified.

.2.9 Shutdown Board (1B-B)

a. Inspection Scope

The team reviewed bus loading calculations to determine whether the 6.9 kilovolt (kV) system had sufficient capacity to support its required loads under worst case accident loading and grid voltage conditions. The team reviewed the design of the degraded voltage protection scheme to determine whether it afforded adequate voltage to safety related devices at all voltage distribution levels. This included review of degraded voltage relay setpoint calculations, review of the degraded voltage logic scheme, and the licensee's response to NRC Information Notice 95-05, "Undervoltage Protection Relay Settings Out of Tolerance Due to Test Equipment Harmonics." The team reviewed the overcurrent protection scheme for the 6.9 kV buses including drawings and calculations to determine whether loads were adequately protected and immune from spurious tripping. The team reviewed 125 volt direct current (VDC) system voltage drop calculations to determine whether 6.9 kV bus circuit breakers had adequate control voltage. The team reviewed the load shedding and load sequencing schemes to determine whether they were consistent with the design bases and design calculations.

The team reviewed maintenance schedules and procedures for the 6.9 kV bus and its associated circuit breakers to determine whether the equipment was being properly maintained. This included reviewing acceptance criteria in procedures for consistency with vendor recommendations and design calculations. The team reviewed the fast bus transfer scheme, including drawings, calculations and procedures, to determine whether the transfer capability described in the UFSAR could be achieved without adverse effects on equipment and systems. The team reviewed calculations for switchgear temperature rise to determine to assess the effect of a loss of ventilation. The team reviewed corrective action documents and maintenance records to determine whether there were any adverse operating trends. In addition, the team performed a visual inspection of the 6.9 kV safety buses to assess material condition and the presence of hazards. For the following operator actions, the team performed a walkthrough of associated AOIs, ARIs, SOIs, and other operations procedures (e.g., temporary instructions, etc.) with plant operators, maintenance personnel 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 to further understand and assess the procedural rationale and approach to meeting the design basis and UFSAR response and performance requirements. Selected operator actions associated with the following event/evolution were reviewed:

• Operator action to crosstie shutdown boards [HASBXT] b.

.1 Findings and Observations

Introduction:

The team identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," with five examples, for the licensee's failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures.

Description:

The team identified the following instances where the design as documented in calculations and vendor instructions was not properly translated into maintenance procedures: 6.9 kV and 480 volt (V) Breaker Operating Voltage Maintenance procedure MI-57.002 provided instructions for checking the operating voltage of 480 V DS type circuit breaker close coils at reduced voltage. A note associated with the applicable step stated that if the minimum voltage exceeds 100 VDC then a problem evaluation report (PER) may be required. The minimum close coil voltage specified by the manufacturer and used as acceptance criteria in calculation WBN EEB-MS-TI11-0004 was 90 VDC. Consequently, an operating voltage of 100 VDC may not be adequate to ensure proper breaker operation under design basis conditions at the start of an accident. A similar discrepancy existed in maintenance procedure MI-57.001 for 6.9 kV Magne-blast circuit breakers, which also cited an acceptance criteria of 100 VDC versus the 90 VDC used in the design basis. In response to this concern, the licensee initiated PER 507645, and performed a partial review (approximately 85%) of maintenance records to confirm that no test results for installed breakers had failed to meet the manufacturer's rating of 90 VDC.

4160 V to 480 V Transformer Insulation Resistance Maintenance procedure MI-57.200 provided acceptance criteria for the 4160 V to 480 V transformer insulation resistance (IR) as 1.5 mega-ohm secondary and 30 mega-ohm primary. Vendor manual VTD W120 2356 specified the minimum IR as 32 mega-ohm for 1.2 kV class windings and 230 mega-ohm for 8.7 kV insulation class windings. In response to this concern, the licensee issued PER 511469 and confirmed that the last performance of maintenance procedure MI-57.200 for each 480 V shutdown board transformer showed actual IR readings above the manufacturer's minimum recommendations.

Breaker Closing Time Maintenance procedure MI-57.001, dated July 5, 2011, provided for as-left testing for the closing time of 6.9kV circuit breakers after maintenance and provided acceptance criterion of less than or equal to 84 milliseconds (5 cycles). Calculation WBN EEB-MS-TI06-0027 evaluated the effects of the fast transfer capability of the 6.9 kV systems and determined that a dead bus time of approximately 4.1 cycles could result in exceeding the industry standard evaluation criteria of 1.33 Volt/Hertz. The acceptance criteria in procedure MI-57.001 was not consistent with the maximum desirable dead bus time determined in the calculation. In addition, maintenance procedure MI-57.001, Step 6.2 [23] did not provide any acceptance criteria for as-found closing time testing. The team noted that the last test for the alternate source breaker for 6.9 kV bus 1B-B, was 72 milliseconds, which was consistent with the dead bus time evaluated in calculation WBN EEB-MS-TI06-0027. In response to this concern, the licensee issued PER 517095.

480 V Breaker Service Life Limitations The vendor manual for Westinghouse type 'DS' circuit breakers specified breaker and subcomponent service life limits. Maintenance procedure MI-57.002 identified the service life limit for the breaker (either 4000 or 1500 cycles, depending on model), but did not reference the more limiting service life limits for breaker subcomponents such as the direct trip actuator (400 cycles) and overcurrent trip switch (400 cycles). In response to this concern the licensee initiated PER 511010 and provided data to justify that no subcomponents have exceeded their service life limits. Specification for Harmonic Distortion Limits for Relay Test Sets Vendor manual WBN-VTD-AS04-0080 for Asea Brown Boveri (ABB) type 27N degraded voltage relays specified the use of a calibration voltage source with a maximum harmonic distortion of 0.3%. Surveillance procedure 1-SI 5-B, Step 4.2 [2] specified the use of a Doble relay test set with a maximum harmonic distortion of 2%, which was non-conservative with respect to the vendor specification. In response to this concern, the licensee issued PER 513893 and confirmed that the Doble test sets actually available and used for calibrating the degraded voltage relays conformed to the vendor's specifications.

Analysis:

The inspectors determined that the failure to correctly translate vendor specifications and design calculations into maintenance and surveillance procedures was a performance deficiency. The performance deficiency was more than minor because it affected the mitigating systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, the finding is similar to Inspection Manual Chapter 0612, Appendix E (example 4.a), because the failure to ensure correct translation of acceptance criteria into procedures was not isolated. The team performed a phase one significance determination process screening and determined the finding to have very low safety significance (green) utilizing the mitigating systems cornerstone column of IMC 0609, Attachment 4, "Phase 1-Initial Screening and Characterization of Findings." The finding did not represent a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the resources component of the human performance area, because the licensee had not ensured that complete, accurate, and up-to-date procedures were consistent with vendor and design specifications; and therefore, the procedures were not available and adequate to assure nuclear safety H.2(c).

Enforcement:

Appendix B to 10 CFR 50, Criterion III, "Design Control," requires in part, that measures be established to assure that the design basis is correctly translated into specifications, drawings, procedures, and instructions. Contrary to this requirement, since the revision of maintenance procedure MI-57.001 on July 5, 2011, the licensee's design control measures had failed to assure that vendor specifications and design calculations were correctly translated into specifications relied on in maintenance and surveillance procedures. Specifically, the licensee failed to implement measures to ensure that specifications for circuit breaker control voltage, transformer insulation resistance, circuit breaker closing times, circuit breaker service life limitations, and relay test set harmonic distortion were correctly reflected in maintenance and surveillance procedures. Because this violation was of very low safety significance and it was entered into the licensee's corrective action program, this violation is being treated as an NCV, consistent with the Enforcement Policy: NCV 05000390/2012008-03, "Inadequate Acceptance Criteria in Maintenance and Surveillance Procedures (5 Examples)." b.

.2 Findings and Observations

Introduction:

The team identified an issue of concern and an unresolved item related to the effect of electrical system harmonics on safety-related degraded voltage relays. Specifically, in 1993, the licensee identified that harmonic distortions adversely affected the 6.9 kilovolt (kV) bus overvoltage relays by causing them to alarm unnecessarily. The licensee entered this issue into their corrective action program and modified the overvoltage relays to minimize the effects. However, the licensee did not identify (or otherwise evaluate) the adverse effect that harmonics could have on the ability of the degraded voltage relays to perform their safety function, as required by limiting condition for operation 3.3.5 of the plant's technical specifications.

Description:

The Watts Bar degraded voltage protection scheme features three ABB type 27N relays for each 6.9 kV safety bus, arranged in a two out of three tripping scheme. The ABB instruction bulletin 7.4.1.7-7 contained in vendor manual WBN-VTD-AS04-0080 states that (1) the relay employs a peak voltage detector, and (2) harmonic distortion on the AC waveform can have a noticeable effect on the relay operating point and the measuring instruments used to calibrate the relay. The bulletin also notes that the relay is available with an internal harmonic filter for applications where waveform distortion is a factor. The team noted that calculation WBPE2119202001, "6.9kV Shutdown & Logic Boards Undervoltage Relay Requirements/Demonstrated Accuracy Calculation," identified the relay as a model not equipped with a harmonic filter, but did not address the basis for excluding harmonic distortion as a factor which affected relay accuracy. In response to the team's inquiries, the licensee provided PER 930397 that addressed spurious actuations of the ABB type 59H overvoltage relays which are similar in design to the ABB type 27N degraded voltage relays. Troubleshooting tests performed to identify the cause of the 59H spurious actuations revealed that high levels of 6.9 kV system harmonics from sources both external and internal to the station accompanied the spurious operations. The 'causal factor' section of PER 930397 stated that the relays sometimes trip on harmonic distortion although the root mean square voltages are at acceptable levels. Corrective actions consisted of replacing the type 59H overvoltage relays with a model equipped with harmonic filters.

The team further noted that the 'extent of condition' section of PER 930397 did not identify or address whether the degraded voltage relay's operating point could also be affected by the same harmonics implicated in the maloperation of the overvoltage relays. The team was concerned that harmonics on the 6.9 kV system could cause the degraded voltage relays to fail to actuate at the setpoint specified by technical specifications. Persistent harmonics can be produced by factors external to the nuclear site or by internal phenomena. A typical internal source of harmonics at nuclear power plants is motor defects. The team was also concerned that transient harmonics could cause the relays to spuriously reset during an actual degraded voltage event, thereby delaying the protective function beyond the 10 seconds stipulated in technical specification limiting condition for operation 3.3.5. Specifically, the degraded voltage relay's design features an 'instantaneous' reset characteristic that could allow reset of the degraded voltage relay in less than two cycles in the presence of harmonics, thereby reinitiating the external 10 seconds timer. The reset function of the existing degraded voltage relays is identical to the tripping function of the overvoltage relays that actuated due to transient harmonics in 1993. In 1993, transient harmonics were measured at levels of greater than 10% total harmonic distortion during the troubleshooting for PER 930397 versus the 0.3% distortion deemed acceptable by the relay vendor. The transient harmonics documented in PER 930397 were attributed to events that included the trip of the nearby Sequoyah generating station, and to breaker operations at the Watts Bar station. The team noted that similar conditions could exist during an accident scenario when proper performance of the degraded voltage scheme time delay would be critical with respect to satisfying the response time assumptions in the accident analysis. In response to the team's concerns, the licensee provided information regarding condition monitoring of large motors that consisted of periodic measurement and analysis of motor bearing vibration from which various defects that may produce harmonics could be identified. The team noted, however, that there was no written guidance or acceptance criteria for these tests that would prompt engineering to investigate whether suspected motor defects could produce harmonics that would adversely affect the accuracy of degraded voltage relays. Specifically, there was no recognition in design or maintenance documents regarding the susceptibility of the degraded voltage relays to harmonic distortion, or the need to investigate suspected motor defects with respect to this susceptibility. The team further noted that during normal bus voltage conditions when voltage is above the degraded voltage relay reset setpoint, harmonics would shift system peak voltage away from the degraded voltage relay operating setpoint rather than closer to it, and so the presence of harmful harmonics would not 'self-reveal' by spurious actuations. The overvoltage relays are now equipped with harmonic filters so they will also not reveal the presence of either transient or persistent harmonics. Based on the team's observations, the licensee has entered these concerns into their corrective action program as PER 515413 and PER 546072. Summary: The team determined that additional review of information recently received from the licensee regarding Watts Bar's design and licensing bases was necessary to determine if the licensee's performance constituted a violation of NRC regulatory requirements. Additionally, the team determined that additional consultation with the Office of Nuclear Reactor Regulation was warranted before reaching a final disposition of the unresolved item. This unresolved item is open pending (1) the review of additional information from the licensee regarding the design and licensing basis of the degraded voltage relays and (2) consultation with the Office of Nuclear Reactor Regulation: URI 05000390/2012008-04, "Effect of System Harmonics on Degraded Voltage Relay Function."

.2.10 Shutdown Transformer (1A1-A)

a. Inspection Scope

The team reviewed load flow calculations to determine whether the transformer was applied within its specified ratings. The team reviewed maintenance schedules, vendor recommendations, and procedures to determine whether the transformers were being properly maintained. This included reviewing acceptance criteria in procedures for consistency with vendor recommendations and design calculations. The team reviewed protective relaying schemes and calculations to determine whether the transformer was adequately protected and whether it was susceptible to spurious tripping.

The team reviewed maintenance and corrective action histories to determine whether there have been any adverse operating trends. In addition, the team performed a walkdown of the installed equipment to determine whether the installed configuration is consistent with design documents including drawings, and calculations, and to assess the presence of hazards.

b. Findings

No findings were identified.

.2.11 Component Cooling System Pump Motor (1A-A)

a. Inspection Scope

The team reviewed maintenance schedules, and procedures, and completed work orders to determine whether the motor was being properly maintained. The team review protective relaying schemes and calculations to determine whether the motor was adequately protected and whether it was susceptible to spurious tripping.

The team reviewed maintenance and corrective action histories to determine whether there have been any adverse operating trends. The team also reviewed the component cooling system pump 1A-A motor circuit breaker. Corrective action and maintenance records were reviewed to determine if there had been any adverse operating trends. In addition, the team performed a walkdown of the installed equipment to determine whether the installed configuration is consistent with design documents including drawings, and calculations, and to assess the presence of hazards.

b. Findings

No findings were identified.

.2.12 Auxiliary Feedwater Level Control Valves (3-156, -164)

a. Inspection Scope

The TS, UFSAR, SDs, and P&IDs, were reviewed to establish an overall understanding of the design bases of the valves. Design calculations (e.g., differential pressure and required torque/thrust) were reviewed to verify that the design basis and design assumptions had been appropriately translated into these documents. Test procedures and results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses served to validate component operation under accident/event conditions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. The team also reviewed instrument loop diagrams, loop setpoints and scaling documents, and loop accuracy calculations to verify that the valve controls were consistent with design bases assumptions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with inservice/equipment qualification life. Component walkdowns were conducted to verify that the installed configurations would support their design basis function under accident conditions and had been maintained to be consistent with design assumptions.

b. Findings

No findings were identified.

.2.13 Refueling Water Storage Tank LCVs (62-135A, -136B)

a. Inspection Scope

The team reviewed applicable portions of the plant's TS, UFSAR and system descriptions to identify design basis requirements for LCV 135A/136B. The team reviewed the calculations that establish control circuit voltage drop, short circuit, and protection/coordination including thermal overload sizing and application to verify adequate protection during design bases scenarios. The team verified by review of control diagrams, that the operation of the LCV was consistent with the design basis and operational requirements. The team interviewed the system engineer to discuss the valve analysis as well as operational and maintenance history to verify that potentially degraded conditions were being appropriately addressed. Test procedures and recent test results were reviewed against design bases documents 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. The team examined maintenance rule documentation to verify that the valves were properly scoped, and monitored. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored, or prevented, and that scheduled component replacements were consistent with vendor recommendations and equipment qualification life. Component walkdown was conducted to verify that the installed configurations would support the design basis function under accident conditions and had been maintained to be consistent with design assumptions. The team conducted a non-intrusive visual inspection of LCV 135A/136B to verify that any potentially degraded material conditions were being appropriately addressed. Also, the team verified testing and calibration of instruments related to the valve.

b. Findings

No findings were identified.

.2.14 125 Volt Battery Boards (I and II)

a. Inspection Scope

The team reviewed the design basis documentation and UFSAR to identify the loading requirements for the vital batteries. The team reviewed the inputs to the battery sizing analysis and the battery voltage study, TS and maintenance allowable terminal load resistance, and panel load schedules to verify the adequate sizing of the battery. The battery voltage study was reviewed to verify adequate voltage was available to critical components. The vendor manual was reviewed to verify battery installation and operating instructions were implemented. Battery TS surveillance test and inspection results were reviewed to verify degradation was identified and anomalies were addressed and corrected. The equipment history as indicated by corrective work orders and condition reports was reviewed to verify that identified equipment problems were corrected. A field walkdown was performed to assess observable material conditions of the batteries.

Also, the team reviewed schematic diagrams and calculations for the normal supply breakers to determine whether equipment operation was consistent with the design bases. The team reviewed calculations for protective device settings to determine whether the breakers were subject to spurious tripping, and whether the breakers were selectively coordinated with upstream devices. Also, the team reviewed associated corrective action history to verify that degraded conditions were being appropriately addressed. In addition, the team interviewed the system engineer and performed a non-intrusive visual inspection of the direct current bus to assess the installation configuration and verify that degraded material conditions were being appropriately addressed.

b. Findings

No findings were identified.

.2.15 Hydrogen Igniters [large early release frequency]

a. Inspection Scope

The team reviewed applicable portions of the plant's TS, UFSAR and system descriptions to identify design basis requirements for the hydrogen igniters. The team interviewed the system engineer to discuss operation and maintenance history to verify that potentially degraded conditions were being appropriately addressed. Operation procedures for emergency power to the hydrogen igniters were reviewed to verify that component operation and power supply alignment were consistent with the design.

Test procedures and recent test results were reviewed against design bases documents 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. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and that scheduled component replacements were consistent with vendor recommendations and equipment qualification life.

b. Findings

No findings were identified.

.3 Operating Experience

a. Inspection Scope

The team reviewed four operating experience issues for applicability at Watts Bar Nuclear Plant. The team performed an independent review for these issues and where applicable, assessed the licensee's evaluation and dispositioning of each item. The issues that received a detailed review by the team included:

• NRC Generic Letter 1988-14, "Instrument Air Supply System Problems Affecting Safety-Related Equipment"

• NRC Regulatory Information Summary 2000-05, "Spring Actuated Safety and Relief Valve Reliability"

• NRC Information Notice 1992-29, "Potential Breaker Miscoordination Caused by Instantaneous Trip Circuitry"

• NRC Information Notice 1987-08, "Degraded Motor Leads in LIMITORQUE DC Motor Operators"

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

On May 15, 2012, the team presented the inspection results to Mr. David Gronek and other members of the licensee's staff. Proprietary information that was reviewed during the inspection was returned to the licensee or destroyed in accordance with prescribed controls.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

C. Borelli, Manager, WBN PRA

K. Dutton, Director, Site Engineering

D. Gronek, Plant Manager

D. Guinn, Manager, Site Licensing

R. Kirkpatrick, Manager, Design Engineering

T. Morgan, Site Licensing

W. Nesmith, Design Engineering

NRC personnel

R. Nease, Chief, Engineering Branch Chief 1, Division of Reactor Safety, Region II

S. Schaeffer, Chief, Project Branch 6, Division of Reactor Project, Region II

S. Sandal, Senior Reactor Inspector, Division of Reactor Safety, Region II

B. Monk, Senior Resident Inspector, Division of Reactor Projects, Watts Bar Resident Office

R. Mathew, Team Leader, Division of Engineering, Office of Nuclear Reactor Regulation

G. Matharu, Senior Electrical Engineer, Division of Engineering, Office of Nuclear Reactor Regulation

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened and Closed

05000390/2012008-01 NCV Failure to Establish Test Procedures to Assure Satisfactory ACAS Performance during Design Basis Accidents [Section 1R21.2.4]

05000390/2012008-02 NCV Failure to Adequately Test the AFW Discharge Check Valves [Section 1R21.2.6]

05000390/2012008-03 NCV Inadequate Acceptance Criteria in Maintenance and Surveillance Procedures (5 Examples) [Section 1R21.2.9]

Opened

05000390/2012008-04 URI Effect of System Harmonics on Degraded Voltage Relay Function [Section 1R21.2.9]

LIST OF DOCUMENTS REVIEWED

Calculations

NAME TITLE REVISION/DATE 048018AFW System Level Review For Watts Bar Auxiliary Feedwater System Air Operated Valves 4048-018-BAL Evaluation Of Required Thrust For Balanced Disk Globe AFW Turbine Driven LCVs AT Watts Bar Nuclear Power Station 0048018MUL Evaluation Of Required Thrust For Multistage Unbalanced Disk Globe AOVs At Watts Bar Nuclear Power Station 2048018UNB Evaluation Of Required Thrust For AFW Motor Driven Large LCVs AT Watts Bar Nuclear Power Station 31-F-74-12 Demonstrated Accuracy Calculation for 1-F-74-12, 1-F-74-24 21-L-3-156A Setpoint and Scaling Document For MDAFW Pump 1A-A Level Control Valves 1-LCV-3-156 and 1-LCV-3-156A 11-LT-3-148 Demonstrated Accuracy Calculation For Steam Generator Narrow Range Level Control Loops 121-PD-3-132A Setpoint And Scaling Document For Pressure Control Valves 1-PCV-3-132-B 12-F-67-280 Emergency Diesel Heat Exchanger 2B2 Discharge Flow 085KS002 IPS ERCW Strainer Sliding Base/HPFP Strainer Anchorage and ERCW Discharge Overflow Structure Ladder 2E31

920108 300 Relay Setting Manual Tab 1A.27 2E31850221300 6.9kV Shutdown Board Normal And Alternate Feeders.

4EBASCO Services Inc. Equipment Seismic Qualification Of System 003 11/07/92EDQ00023620070003 125V DC Vital Battery System Analysis 4 & 13EPM-AF-080188 System Safety Limits For The Auxiliary Control Air Subsystem (ACAS) Safety Related Instrumentation 21EPMJJL081689 Auxiliary Building

EL 676 RHR Pump Rooms HVAC Cooling Load During Unit 1/Unit 2 LOCA 4EPMJN010890 Performance Of CCS Heat Exchangers 14EPM-MECB-111886A RHR SYS Net Positive Suction

Head (NPSH) Calculations 4EPM-MGF-080789 ACAS Compressor Capacity For Normal And DBA Conditions 5EPMOED070391 Equations For AFW Pump Performance Curves 9EPM-PFS-053191 Minimum Fluid Velocity Through Check Valves 02/11/11EPMPTC120594 Essential Raw Cooling Water System Pressure Drop Calculations 14EPMRM062289 Process Safety Limits For SG Narrow And Wide Range Level Instrumentation Setpoints 5HCGTBG091981 Design Parameters For Motor And Turbine Driven AFW Pumps 8MDQ00100320020070 Component Level Review Calculation For Watts Bar AFW Air Operated Valves 5

MDQ1003990045 Analysis Of Leaking Main And Auxiliary Feedwater Valves 2MQD 000 082 2003 0077 Emergency Diesel Generator Jacket Water Heat Exchanger Evaluation 2SSD 1-F-63-91B Setpoint And Scaling Document 1-F-63-91B 3SSD 1-F-63-92B Setpoint And Scaling Document 1-F-63-92B 3

SSD-1-LPL-62-129A Scaling And Setpoint Document Loop Cover Sheet 6Walworth Company

Seismic Calculation No. 281 1WAT-D-11490 TVA Watts Bar Nuclear Plant Unit 1 ECCS Analysis 08/22/06WB-DC-30-28 6.9KV SB RHR Pump Relay Setting 2WBN EEB 1-LT-3-148 Demonstrated Accuracy Calculation For Steam Generator Narrow Range Level Control Loops 12WBN EEB 1-PDT-3-132A-B Demonstrated Accuracy Calculation For MDAFW Pump Differential Control Loops 4WBN0SG4136 Steady State DBE LOCA Temperatures For The Auxiliary Building 21WBNEBMSTI110004 125V DC Voltage Analysis 64WBNEEBEDQ1999010001 AC Auxiliary Power System Analysis 72 & 76WBN-EEB-EDQ1999-010001 Unit 1 SI Phase B Grid Drop 164-153KV Shutdown CSST A &D Page 1 11/04/09WBN-EEB-MS-T108-0008 480V 1E Coordination/Protection 150WBNEEBMSTI060008 480V 1E Coordination/Protection 151WBNEEBMSTI060027 WBN Fast Transfer Analysis 0WBNEEBMSTI060029 Degraded Voltage Analysis 33WBNEEBMSTI070005 125V DC Protection And Coordination Calculation 60WBNOSG4242 6.9 kV and 480 V Board Room Temperature Analysis 3WBNSSG4002 WBN Hydrogen Volume Percent In Containment Following A LOCA 10WBPE2119202001 6.9kV Shutdown & Logic Boards Undervoltage Relay Requirements / Demonstrated Accuracy Calculation 9WCGACQ0127 Seismic Qualification for ERCW Strainers 1A-A/2 A-A/1B-B/2B-B 3WCGACQ543 Seismic Evaluation Of 125 Volt Battery Rack 0

Completed ProceduresNAME TITLE REVISION/ DATE 08-820491 Perform Procedure 1-SI-74-901-A, Residual Heat Removal Pump 1A-A Quarterly Performance Test 02/01/0908-820919 Perform Procedure 1-SI-63-907, RHR Hot Leg And Cold Leg Injection Check Valve Testing During Refueling 10/03/090-SI-236-21 125VDC Vital Battery I Quarterly Inspection 10/20/110-SI-236-22 125VDC Vital Battery II Quarterly Inspection 09/28/110-SI-236-31 125VDC Vital Battery I Annual Inspection 06/04/110-SI-236-32 125VDC Vital Battery II Annual Inspection 06/29/10

0-SI-236-41 125VDC Vital Battery I 18 Months Service Test and 125VDC Vital Charger I Test 06/13/110-SI-236-42 125VDC Vital Battery

II 18 Months Service Test and 125VDC Vital Charger II Test 04/07/110-SI-236-51 125VDC Vital Battery I 60 Months Service Test and 125VDC Vital Charger I Test 05/31/110-SI-236-52 125VDC Vital Battery

II 60 Months Service Test and 125VDC Vital Charger II Test 06/09/1110-814341 Perform Procedure

TI-104, Instrument Air Quality, (Rev. 11) 12/09/1010-815772 Perform Procedure 1-SI-63-907, RHR Hot Leg and Cold Leg Injection Check Valve Testing During Refueling 04/04/11111591196 Perform Procedure 1-SI-74-901-B, RHR Pump 1B-B Quarterly Performance Test 03/09/11112016541 Perform Procedure

TI-104, Instrument Air Quality, (Rev. 11) 09/19/111-SI-3-906-A Valve Testing during Hot Standby-Auxiliary Feedwater System (Train A) 05/18/111-SI-3-923-A Auxiliary Feedwater Pump 1A-A Comprehensive Pump Test 05/11/111-SI-3-923-A Auxiliary Feedwater Pump 1A-A Comprehensive Pump Test 10/13/09PTI-063-03 Safety Injection System Charging, SI, and RHR Flow Balance Test, (Pre-operational Test, 12/16/94) 4TI-79.821 Diesel Generator 1A-A Jacket water Cooler Performance Test 05/13/10TI-79.824 Diesel Generator 2B-B Jacket Water Cooler Performance Test 05/09/09 Completed Work Orders NAME TITLE REVISION/ DATE 03-005702-000 OVHL Inspection of GE

AK-50 Circuit Breakers PMUG M0566W 04/05/0403-020188-000 OVHL Inspection of GE

AK-50 Circuit Breakers PMUG M0566W 06/18/0504-813376-000 6.9KV Circuit Breaker Inspection Perform

MI-57.001 07/06/0405-816107-000 Provide Substitute Breaker during PM Performance This Breaker is the Normal Feeder for RHR Pump 1B-B 07/24/0605-820315-000 Periodic Required Replacement Of MEQ Belts 05/04/0705-822154-000 Replace Belts And Each Fan Shaft Bearing 10/11/0606-822047-000 6.9KV Circuit Breaker Inspection and Overhaul

MI-57.006 Perform

MI-57.006. 6.9KV SD BD 1A-A RHR PMP 1A-A

05/03/09

Completed Work Orders NAME TITLE REVISION/ DATE 07-815593-000 Disassemble Clean & Inspect CCs HX B Perform M4049F &

MI-70.02WM4049F 03/25/0907-816062-000 ESFA System Slave Relay Go Test Train A 03/21/0807-816064-000 ESFA System Slave Relay Go Test Train A 03/17/0807-817941-000 6.9KV Circuit Breaker Inspection and Overhaul

MI-57.006 Perform

MI-57.006. 6.9KV SD BD 1B-B CMPT 14 RHR PMP 1B-B

07/30/0807-820240-000 Relay and Indication Instrument Test Perform

MI-236010 11/24/0807-820818-000 GE AK50 Circuit Breaker Inspection Per

MI-57.004. 125V Vital Batt BD 2 CMPT 109 Vital Batt II Tie to Vital Batt BD

II 11/06/0907-820818-001 GE AK50 Circuit Breaker Inspection Performance

MI-57.004 11/06/0908-812842-000 Inspect / Lube of Valve Stems for MOV's EQ and 89-10 Includes Gear Case Lube Check for 10 Specific MOV's 10/09/0908-812883-000 Bridge Megger & Hypot Testing MPM WM1992V 09/07/0908-814211-000 480V Breaker Inspection Perform

MI-57.002 04/23/10

08-816283-000 GE AK50 Circuit Breaker Inspection Per

MI-57.004 Load: Discharge Test Breaker 0-BKR-236-4/108-G 06/12/0908-820009-000 125V Vital Battery Charger and BD Inspection and Cleaning Train B 09/02/0908-820443-000 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 01/29/0908-820596-000 125V Vital Battery Charger and BD Inspection and Cleaning Train A Wang 0406F Multi-Equip 11/18/0908-820624-000 ESFA System Slave Relay Go Test Train A 10/23/0908-820822-000 ESFA System Slave Relay Go Test Train B 10/12/09

08-820923-000 AFW Check Valve Testing During hot Standby TDAFW 10/18/0908-820924-000 AFW Pump 1A-S Comprehensive Pump Test 10/18/0908-820939-000 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 10/15/0908-820949-000 Inspection Of MDAFW Train Check Valves 09/23/0908-821019-000 Inspection Of TDAFW Train Check Valves 09/23/0908-821030-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 10/15/0908-821120-000 Component Cooling System Heat Exchanger C Performance Test 10/22/0909-810430-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 04/01/0909-810731-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 05/04/0909-812024-000 1-SI-211-4-B 92 Day Trip Actuating Device Operational Test on Undervoltage Relays for 1B-B 6.9 kV Shutdown Board 07/01/0909-812029-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 07/02/0909-812383-000 Dry Boron Packing Leak 03/12/0909-812719-000 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 07/31/09

09-815903-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 10/15/0909-816178-000 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 10/15/0909-818104-000 1-SI-211-6-B Channel Calibration Test on Loss of Voltage Relays for 6.9kV Shutdown Board 1B-B 12/28/0909-818105-000 1-SI-211-5-B Channel Calibration Test on Degraded Voltage Relays for 6.9kV Shutdown Board 1B-B 12/28/0909-818107-000 1-SI-211-7-B Channel Calibration Test on Load Shed and Diesel Start Undervoltage Relays for 6.9kV Shutdown Board 1B-B 12/29/0909-818110-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 12/30/0909-818846-000 1-SI-268-1-A 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 02/01/1009-820542-000 Received Alarm Window 1B-B 125 DC Vital Battery BD

II 11/05/0909-820846-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 03/31/1009-821129-000 Relay and Indication Instrument Test Perform

MI-236011 04/05/1009-821285-000 Sample and Lube of Comp Cool Water Pump Mtr

MPM 4694F 04/23/1009-821285-000 Sample and Lube of Comp Cool Water Pump Mtr

MPM 4694F 04/23/1009-821286-000 Long term Bend Radius Program Plan Meggering 480V Class 1E Greater Than or Equal to 100HP Motors 04/23/1009-821291-000 Relay and Indication Instrument Test Perform

MI-236010 08/20/1010-810289-000 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 04/29/1010-811575-000 Troubleshoot/ Correct Cause of Low Voltage on 'A' Phase 02/03/1010-812051-000 Disassemble Clean & Inspect CCs HX A Perform M4049F &

MI-70.02WM4049F 03/06/1110-812051-000 Disassemble Clean & Inspect CCs HX C Perform M4049F &

MI-70.02WM4049F 07/20/0910-812115-000 Disassemble Clean & Inspect CCS HX B PerformM4049F &

MI-70.02WM4049F

03/06/1110-812310-000 Inspect / Lube of Valve Stems for MOV's EQ and 89-10 Includes Gear C 04/22/1110-812441-000 Inspection of Switchgear Bus and MCC Perform

MI-57.200 04/21/1110-812648-000 125V Vital Battery Charger and BD Inspection and Cleaning Train B Wang 0516W Multi-Equip 12/29/1010-812985-000 OD- Add Baker Explorer

EP-2 Accessory Device to Allow Condition -10-813381-000 1-SI-211-7-B Channel Calibration Test on Load Shed and Diesel Start Undervoltage Relays for 6.9kV Shutdown Board 1B-B 06/29/1010-813384-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 07/01/1010-813689-000 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 07/28/1010-814471-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 09/30/1010-814767-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 11/01/10

10-815526-000 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 01/06/1110-815683-000 ESFA System Slave Relay Go Test Train B 05/12/1110-815685-000 ESFA System Slave Relay Go Test Train A 05/11/1110-815706-000 18 Month Channel Calibration of Volume Control Tank Level Loop 1-LPL-62-129A 04/21/1110-815736-000 Check Valve Testing During Hot Standby Auxiliary Feedwater System (Train A) 04/04/1110-815769-000 ESFA System Slave Relay Go Test Train A 05/14/1110-815790-000 18 Month Permanent Hydrogen Mitigation System Train A Operability Test 05/09/1110-815875-000 18 Month Permanent Hydrogen Mitigation System Train B Operability Test 05/10/1110-815878-000 18 Month Channel Calibration Of Steam Generator 1 Auxiliary Feedwater Level Loop 1-LPL-164 08/16/1110-815878-000 18 Month Chanel Calibration Of Steam Generator 1 Auxiliary Feedwater Level Loop 05/08/1110-816014-000 Component Cooling System Heat Exchanger A Performance Test 05/16/1110-816045-000 Auxiliary Feedwater Pump 1A-A Comprehensive Pump Test 05/12/1110-8813379-000 1-SI-211-5-B Channel Calibration Test on Degraded Voltage Relays for 6.9kV Shutdown Board 1B-B 06/28/10111057630 Inspect Clean and Lube of RHR Motor 03/08/11111202479 0-BD-236-0001 125 Vital Battery Board 10/21/10111429781 125V Vital Battery Charger and BD Inspection and Cleaning Train A Wang 0406F Multi-Equip 07/18/1111-1516647-000 Check Valve Setup For Problems & Correct As Required 11/17/10111552771 1-SI-268-1-A 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 01/24/11111608627 1-SI-268-1-A 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 05/09/11111691841 1-SI-268-1-B 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 04/01/11111984456 Relay and Indication Instrument Test Perform

MI-236011 09/30/11112015903 1-SI-268-1-B 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 07/01/11112015907 1-SI-268-1-B 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 12/28/11112016754 1-SI-268-1-A 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 07/27/11112016755 1-SI-268-1-A 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 11/03/1111-2017336-000 1-SI-3-901-B MDAFW Pump 1B-B Quarterly Performance Test 11/04/11112164566 Sample and Lube of Comp Cool Water Pump Mtr

MPM 4694F 11/14/11112210971 Failure 1-LCV-62-136 04/29/1111-2260678-000 1-SI-3-906-A Check Valve Testing During Hot Standby AFW System (Train A) 05/18/1196-124835-000 MDAFW Pump 1B-B Quarterly Performance Test 09/11/96980855500

TI-50.018, Auxiliary Air Compressor B-B Capacity and Check Valve Test (Train B) 03/27/99

WBN Master

PM 1013W Perform Motor Current Signature Analysis on Safety Related Motors 0 Corrective Action Documents NAME TITLE REVISION/ DATE WBSCA910278 Significant Corrective Action Report Westinghouse DS Circuit Breaker Discriminator Circuit 12/03/91II-S-91-120 Inadequate Circuit Breaker Coordination 01/02/92129486 ASME Code Class Boundary Deficiency -165582 Battery Cell 60 Not Monitored During 0-SI-236-44 -

171904 Unqualified Cable Connectors - C and D -200971 125V DC Vital Battery System Analysis Calculation -203658 1-LT-62-129A Was Found Out of As Found Tolerance -210812 CDBI Self-Assessment PER Regarding Typo in TRM Table 3.3.2-1 -220754 Install TACF Vertical Support Plates On All 4 ERCW Strainer Base Plates 03/12/10221018 Strainer Anchorage Qualification Cannot Be Located 03/15/10223677 A Calculating Evaluating the ERCW Strainers Was Not Included in CCRIS 04/01/10231297 MDAFW Pump B-B Discharge Pressure Control Valve Found Closed With MDAFW Pump Running Post Trip 05/25/10246447 Increasing Trend In 1-PCV-3-132 Stroke Times 08/26/10294604 1-PCV-3-132 Spring Rate 12/08/10297829 1-LSV-3-164-A Has The Wrong UNID Was Hanging On The Solenoid 12/15/10320932 1-PCV-3-132 Failed Stroke Time During 1-SI-3-901-B 02/09/11328153 Inspect Sheave On Motor And Fan And Replace As Necessary 02/24/11330125 Did Not Get To Start

PMT 02/28/11345893 Vital Charger II Not Functioning Properly -349207 Noticed The Flow To The #3 And #4 S/G Drop Dramatically 04/04/11

357399 PER To Document Presence Of 3 Live Clams Approximately 5/8 Inch In Diameter In CCS HTX A 04/20/11361423 After restoration of 1B-1B RXMOV BD 1-LCV-62-136 had no indication of being energized and would not operate from the control room hand switch -363217 C CCS Heat Exchanger Will Not Allow Adequate Flow 05/03/11370468 Controller Will Not Control In Automatic 05/17/11415215

TI-104 gaps, size of particulates > 25 micron needs to be documented 08/10/11436594 Inboard Motor Bearing Degraded 09/22/11438578 Cover Installed On Wrong Heat Exchanger 09/27/11441625 Received alarm window 17-B and could not determine the cause of the alarm after board walkdown. Troubleshoot to determine cause and resolve issue to clear MCR alarm. When a WO is created it should be categorized as a Control Room Deficiency -

444994 All AFW LCW Will Not Maintain SG Levels Within Specified Control Bands Per Procedures In Auto 10/11/11445835 AFW Level Control Valves Will Not Control In Auto 10/12/11445838 Aux Feedwater Level Control Valve Will Not Control In Auto 10/12/11445839 Aux Feedwater Level Control Valve Will Not Control In Auto 10/12/11446464 AFW Level Control Valves Will Not Control In Auto 10/13/11

468950 WBN has adverse capacity degradation for Vital Battery III and IV -500697 Margin Management: Increasing Trend In 1-PCV-3-132 Stroke Times 02/03/12930397 Overvoltage Condition on 6.9kV Shutdown Bds. 10/28/93

Drawings

NAME TITLE REVISION/ DATE 0-75W1508-1 500kV Switchyard Main Single Line 11-15E500-1 Key Diagram Station Aux Power System 351-15E500-2 Station Auxiliary Power System 41

1-15E500-3 Transformer Taps & Voltage Limits Aux Power System 231-45W700-1 Key Diagram 120V AC & 125V DC Vital Plant Control Power System 311-45W703-1 Wiring Diagrams 125V DC Vital Battery Board I, SH 1 511-45W703-10 Wiring Diagrams 125V DC Vital Battery Board I Instrument Loading, SH 1 41-45W703-11 Wiring Diagrams 125V DC Vital Battery Board I Instrument Loading, SH 2 61-45W703-12 Wiring Diagrams 125V DC Vital Battery Board II Instrument Loading, SH 3 31-45W703-13 Wiring Diagrams 125V DC Vital Battery Board II Instrument Loading, SH 4 61-45W703-2 Wiring Diagrams 125V DC Vital Battery Board II Single Line, SH 2 381-45W703-5 Wiring Diagrams 125V DC Vital Battery Board I Panel 4 Connection Diagram, SH 5 341-45W703-5A Wiring Diagrams 125V DC Vital Battery Board I Panel 4 Connection Diagram, SH 5A 281-45W703-6 Wiring Diagrams 125V DC Vital Battery Board II Panel 4 Connection Diagram, SH 6 211-45W703-6A Wiring Diagrams 125V DC Vital Battery Board II Panel 4 Connection Diagram, SH 6A 251-45W724-1 6900V Shutdown Board 1A-A Single Line 251-45W724-2 6900V Shutdown Board 1B-B Single Line 241-45W749-1 Wiring Diagrams 480V Shutdown Board 1A1-A Single Line 581-45W749-2 Wiring Diagrams 480V Shutdown Board 1A2-A Single Line 541-45W749-3 Wiring Diagrams 480V Shutdown Board 1B1-B Single Line 57

1-45W749-4 Wiring Diagrams 480V Shutdown Board 1B2-B Single Line 631-45W751-1 Wiring Diagrams 480V Reac MOV BD 1A1-A Single Line, SH 1 501-45W751-2 Wiring Diagrams 480V Reac MOV BD 1A1-A Single Line, SH 2 30

1-45W751-3 Wiring Diagrams 480V Reac MOV BD 1A1-A Single Line, SH 3 441-45W751-7 Wiring Diagrams 480V Reac MOV BD 1B1-B Single Line, SH 2 541-45W751-8 Wiring Diagrams 480V Reac MOV BD 1B1-B Single Line, SH 1 271-45W751-9 Wiring Diagrams 480V Reac MOV BD 1B1-B Single Line, SH 3 391-45W760-211-1 Wiring Diagram 6900V Shutdown Power Schematic Diagram 181-45W760-211-17 Wiring Diagram 6900V Shutdown Power-Train B Schematic Diagram 101-45W760-211-20 Wiring Diagram 6900V Shutdown Power Schematic Diagram 71-45W760-211-21 Wiring Diagram 6900V Shutdown Power Schematic Diagram 91-45W760-211-7 Wiring Diagram 6900V Shutdown Power-Train A & B Schematic Diagram 151-45W760-62-7 Wiring Diagrams Chemical & Volume Control System Schematic Diagrams 131-45W760-70-1 Wiring Diagrams Component Cooling System Schematic Diagrams 251-45W760-70-9 Wiring Diagrams Component Cooling System Schematic Diagrams 141-45W760-74-1 Wiring Diagram RHR System Schematic Diagrams 121-47W610-32-2 Electrical Control Diagram Control Air System 221-47W610-62-3 Electrical Control Diagram Chemical & Volume Control System 251-47W610-74-1 Electrical Control Diagram RESIDUAL HEAT REMOVAL SYSTEM 171-47W611-62-4 Electrical Logic Diagram Chemical and Volume Control System 121-47-W611-65-1 Electrical Logic Diagram Emergency Gas Treatment System 111-47W611-74-2 Electrical Logic Diagram RHR System 5

1-47W803-1 Flow Diagram Feedwater 581-47W803-2 Flow Diagram Auxiliary Feedwater 591-47W803-3 Flow Diagram Main and Auxiliary Feedwater 311-47W810-1 Flow Diagram Residual Heat Removal System 191-47W811-1 Flow Diagram Safety Injection System 55

1-47W813-1 Flow Diagram Reactor Coolant System 431-47W845-1 Mechanical Flow Diagram Essential Raw Cooling Water System (IPS, Yard, DGB) 571-47W845-2 Mechanical Flow Diagram Essential Raw Cooling Water System (Aux Bldg) 821-47W845-4 Flow Diagram Essential Raw Cooling Water 321-47W848-1 Mechanical Flow Diagram Control Air 26

1-47W850-2 Flow Diagram Fire Protection Raw Service Water 351-47W859-1 Flow Diagram Component Cooling System 521-47W881-4 Flow Diagram Diesel Cooling Water System 325402-011-V1A-MPVT-00086-

001 SULZER Pump Curve, (100022172 Final Performance Test Results) 02/16/11

37VD206-46-1 Valve Data Sheet, Figure

CVIG-2015-SCE-N 137W206-2 Mechanical Intake Pumping Station & Equipment 2947A373-7 Heating Ventilation & Air Conditioning Air Cooling Units 08/08/7447-W920-1 Heating Ventilating & Air Conditioning 37 6036F04001 Schematic Diagram-Jacket Water Heat Exchanger, (Morrison Knudsen Power Systems Division Drawing) -6036F04002 Heat Exchanger Location w/pipe details, (Morrison Knudsen Power Systems Division Drawing) -720002869 Control Valve 4" Model 37-20721

BE-272992 4" Class 900 Valve 4N04-2215-500-1 Assembly-20"-150 ANSI CVIB Check Valve ASME Sec. III, Class 3

BT-36290-1 ERCW Pump B-A Original Performance Curve, (BYRON JACKSON TEST) 11/22/76T-36307-1 ERCW Pump D-A Original Performance Curve, (BYRON JACKSON TEST) 12/02/76 Modifications NAME TITLE REVISION/ DATE

DCN 18152-A Disable Discriminator Circuit Device ADCN 57975 1-PCV-003-0122, 0132/003B Partially Open 05/20/11 Miscellaneous NAME TITLE REVISION/ DATE 1

FE 610 RHR MINIFLOW Watts Bar Unit No. 1 SP. SHT. 6.52, (Orifice Plates and Flanges) PAGE 27

017460-25 Final Qualification Plan for Tayco Hydrogen Igniters for Tennessee Valley Authority for use in Sequoya and Watts Bar Nuclear Power Generating Stations 07/12/821FE 971 RHR HX. NO. 1 Outlet Fl. Watts BAR No. 1

SP.SHT 9.52, (Orifice Plates and Flanges) PAGE 42 03-OT-JPMA047A Local Control of SG Levels (AFWT A-S) 13-OT-JPMA098 Alignment of ERCW to 1A-A CCP Lube Oil Heat Exchanger per

AOI-15 7Atlas Industrial Mfg. Co. Jacket Water Cooler Heat Exchanger Specification Sheet 04/16/74DS-1935-2584 Check and Isolation Valves for High-Pressure and ERCW Valves 4ES-1.1 SI Termination 17GL 88-14 Instrument Air Supply System Problems Affecting Safety-Related Equipment 08/08/88IN 89-26 Instrument Air Supply to Safety-Related Equipment 04/07/89IN 92-29 Potential Breaker Miscoordination Caused by Instantaneous Trip Circuitry 04/17/92IN 95-05 Undervoltage Protection Relay Settings Out of Tolerance Due to Test Equipment Harmonics 01/20/95

IN 97-90 Use of Non-conservative Acceptance Criteria in Safety-Related Pump Surveillance Tests 12/30/97L44

890223 305 TVA Response to

GL 88-14 02/23/89L44

900712 802 Revised Response to

GL 88-14 07/12/90N3-32-4002 Compressed Air System 8N3-63-4001 Safety Injection System 27

N3-67-4002 Essential Raw Cooling Water System 25N3-74-4001 Residual Heat Removal System 4N3-82-4002 Standby Diesel Generator System 16NER 92 0038 Inadequate Circuit Breaker Coordination 01/13/92NP5479 Application Guide for Check Valves in Nuclear Power Plants 1

NPG-DCD-WB-DC-30-27 AC and DC Control Power System - (Unit 1/ Unit 2) 33NRC Bulletin No. 88-04 Potential Safety-Related Pump Loss 05/05/88NRC letter to TVA Watts Bar Nuclear Station - Response to Generic Letter (GL) 89-13 - Service Water System Problems Affecting Safety-Related Equipment 07/09/90OE 5548 Operating Experience Program Information Transmittal 09/17/92SSD-1-SPF-63-91B Scaling Setpoint Document, RHR Pump A-A to RCS 2 & 3 CL 5SSD-1-SPF-63-91C Scaling Setpoint Document, RHR Pump A-A to RCS 2 & 3 CL 3SSD-PI-74-18-D RHR Pump Outlet Pressure 0SSD-PI-74-22-D RHR Pump Inlet Pressure 0TI-100.001 Appendix A, 09-PV-0013, RHR Pump 1A-A, Pump Reference Value Worksheet 17TI-100.001 Appendix A, 09-PV-0013, RHR Pump 1B-B 10TI-100.11 Inservice Testing Bases Document 23TI-79.821 Diesel Generator 1A-A Jacket water Cooler Performance Test 2TI-79.824 Diesel Generator 2B-B Jacket water Cooler Performance Test 2TTL No.

IRA-03 Seismic Qualification of Air Receiver, (for TVA, Watts Bar, Ingersoll-Rand Air Power Compressor Division) 10/11/77TVA letter to NRC Sequoyah, Browns Ferry and Watts Bar Nuclear Plants-Response to Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-Related Equipment 01/26/90TVA letter to NRC Watts Bar Nuclear Station - Response to Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-

Related Equipment 07/09/90TVA letter to NRC Watts Bar Nuclear Plant Unit 1s-Response to Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-Related Equipment, Request for Additional Information (RAI) Concerning TVA's Amended Commitment 03/23/94TVA letter to NRC Watts Bar Nuclear Plant Unit 1s-Response to Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-

Related Equipment, Revised

Methodology and Schedule 07/30/97TVA letter to NRC Watts Bar Nuclear Plant Unit 1-Response to Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-Related Equipment, Revised

Methodology and Schedule 11/05/97

TVA Valve Specification No. 38-83015 Attachment B, Rev.2 08/20/76WAT-D-9911 RHR Pump Kmin ECCS Evaluation 01/13/95WBN Cables Standard Report Cable 1V207A 07/15/92WBNEQ-MOV-001 Limitorque Motorized Valve Operator with Type RH Insulated Motor, Sheet 5a 02/10/89WBN-SDD-N3-3B-4002 Auxiliary Feedwater System 16WBN-SDD-N3-82-4002 Standby Diesel Generator System 16WBN-VTD-AS04-0080 Instructions for Asea Brown Boveri Single Phase Voltage Relay Type 27N High Accuracy Undervoltage Relay and Type 59N High Accuracy Overvoltage Relay 0WBN-VTD-C770-0020 Cutler Hammer Instruction Sheet Bulletin 10250 Heavy Duty Oil Tight Control Units Renewal Parts 0WBN-VTD-G080-1170 GE Instructions for Low Voltage Power Circuit Breakers (PUB. #

GEK-7303C) 4WBN-VTD-O015-0020 CCS Heat Exchanger Data Sheets -WBN-VTD-W030-0020 Walworth Company Maintenance Manual Cast Steel Bolted Bonnet Gate, Globe and Swing Check Valves 01/31/86WBN-VTD-W120-0590 Instruction Book Motor Equipment RHR Pump Motor 0WBN-VTD-W120-2356 Westinghouse Instructions for Determination of Dryness and Drying Out of Transformers 1WBN-VTD-W120-2958 Westinghouse Motor Operated Gate Valves Manually Operated Gate Valves Swing Check Valves 9WBN-VTD-W120-3023 Westinghouse

MPM-DS Maintenance Program Manual for Safety Related Type DS Low Voltage Metal Enclosed Switchgear (Pub #MPM-DS) 12WBN-VTM-W090-0080 Weschler Electric Co. K-241 Switchboard Instruments 1WBN-VTM-W120-0480 Westinghouse 125VDC Vital Battery Board 1 thru IV and Kemco Unit Control Inc. 125 VDC Vital Battery Board V and Distribution Panels A, B and O 17WBN-VTM-W120-0490 Instructions for Westinghouse AB

DE-ION Circuit Breakers Types EB, EHB, FB and Mark 75 Type HFB 0WBN-VTM-W120-0500 Instructions for Westinghouse AB

DE-ION Circuit Breakers Types LBB, LB and Mark 75 Type HLB 0WBPER

148716 Review NRC

IN-1997-21 03/10/09WPN-DS-501433-0201 Check and Isolation Valves for ERCW System, (includes seismic evaluation) 5 Procedures NAME TITLE REVISION/ DATE 0-MI-0.026 Heat Exchanger Cleaning And Inspection 1

0-MI-0.34 Non-Intrusive Check Valve Testing 0

0-MI-57.036 Electric Motor/Generator Replacement and Bearing Maintenance 20-MI-57.108 Insulation Resistance And Continuity Tests For Rotating Machinery Valves Cables Buses and Transformers 30-SI-32-902-B Aux Air Compressor Cooling Water Valve Full Cycle Exercising During Normal Operation - Train B 121-SI-211-5-B Channel Calibration Test on Degraded Voltage Relays for 6.9kv Shutdown Board 1B-B 51-SI-268-1-A 92 Day Permanent Hydrogen Mitigation System Train A Igniter Availability Test 31-SI-268-1-B 92 Day Permanent Hydrogen Mitigation System Train B Igniter Availability Test 31-SI-268-2-A 18 Month Permanent Hydrogen Mitigation System Train A Operability Test 41-SI-268-2-B 18 Month Permanent Hydrogen Mitigation System Train B Operability Test 41-SI-3-906-A Check Valve Testing During Hot Standby - Auxiliary Feedwater System (Train A) 131-SI-3-906-B Check Valve Testing During Hot Standby - Auxiliary Feedwater System (Train B) 131-SI-3-911 Non-Intrusive Testing Of Motor Driven AFW Pump Check Valves During Refueling 41-SI-3-912 AFW Check Valve Testing During Hot Standby Turbine Driven Flowpath 151-SI-3-917 Non-Intrusive Testing TD AFW Check Valves During Refueling 41-SI-3-923-A Auxiliary Feedwater Pump 1A-A Comprehensive Pump Test 61-SI-3-923-S Auxiliary Feedwater Pump 1A-S Comprehensive Pump Test 5

1-SI-63-907 Residual Heat Removal Hot Leg and Cold Leg Injection Check Valve Testing During Refueling 24AOI-10 Loss of Control Air 40AOI-13 Loss of Essential Raw Cooling Water 39AOI-15 Loss of Component Cooling Water (CCS) 33

AOI-40 Station Blackout 15 & 16AOI-7.06 Alignment of HPFP Water to the Steam Generators 4ARI-223-229 ERCW 7ARI-230-240 CLRS ERCW

SUP 15ARI-241-253 CCS 11

ARI-254-264 CCS 4ARI-95-101 Reactor Coolant Pumps 33ECA-0.0 Loss of Shutdown Power 22EDMG-2 Serious Event Mitigation 0ES-1.2 Post LOCA Cooldown and Depressurization 15ES-1.2 Post LOCA Cooldown and Depressurization, (HP-WOG Document)

HP-Rev. 2FR-H.1 Loss Of Secondary Heat Sink -GO-2 Reactor Startup 40MI-17.018 Flood Preparation-High Pressure Fire protection System Spool Pieces 10MI-70.002 Component Cooling Heat Exchanger Maintenance And Testing 13

NEDP-22 Operability Determinations And Functional Evaluations 11NETP-107 Medium Voltage Motor Testing and Maintenance Program 4NPG-SPP-03.1.4 Corrective Action Program Screening And Oversight 3OPDP-8 Limiting Conditions For Operation Tracking 5SOI-2&3.01 Condensate And Feedwater System 116

SOI-3.02 Auxiliary Feedwater System 50SOI-67.01 Essential Raw Cooling Water System 114SOI-84.01 Flood Mode Boration Make-Up System 9TI-100.001 Inservice Testing of Pumps 17TI-104 Instrument Air Quality 11

TI-31.02 Plant Equipment Vibration Monitoring & Vibration Diagnostics Program 22TI-79.000 Program for Implementing NRC Generic Letter 89-13 11TI-79.703 Component Cooling System Heat Exchanger C Performance Test 8TI-8.06 Check Valve Program 2WOG

Document

ERG-EOP Step Deviation Document EOP Document Form, (EOP No.

ES-1.2) 12/13/10 Service Requests/Problem Evaluation Reports Initiated Due to CDBI Activity NAME TITLE REVISION/ DATE

499225 NRC Identified-Housekeeping in Unit 1 Terry Turbine Room 02/01/12499275 NRC Inspector Observed Unsecured Wrench during CDBI Walkdown at the

IPS 02/01/12499950 CDBI 2012; IST Program Check Valve Classification 02/02/12500684 2012 CDBI Identified, Evaluate Auxiliary Control Air System (ACAS) testing 03/03/12500697 Margin Management: Increasing Trend In 1-PCV-3-132 Stroke Times 02/03/12500726 2012

CDBI-HPFP System Description Deficiency 02/03/12501737 CDBI Walkdown 219-Unsecured Unlabeled Ladder on Floor of

AB 692' Pen Room 02/06/12501792 2012 CDBI 47W821 Series Drawings have Extraneous Outdated Information Depicted 02/06/12501941 2012 CDBI NRC Identified Evaluate Auxiliary Control Air System (ACAS) Testing 02/06/12502543 Issue relating from an NRC CDBI 2012 Question on ERCW Check Valves, (minimum flow required to maintain the ERCW pump discharge check valve discs in the fully open position) 02/07/12503052 NRC Identified- Unrestrained Temporary Equipment 02/08/12503187 Issue Resulting from an NRC CDBI 2012 Question on ERCW Check Valves 02/08/12506577

CDBI-2012; IST Program ERCW Check Valve Classification 02/15/12507667

CDBI-2012 IST Program ERCW Check Valve Classification 02/16/12507675 CDBI - NRC Identified Degraded Breaker Control Voltage Issues with Maintenance Procedures 02/12/12508099 CDBI Misc HO Lock Tabs Identified on Floor behind 1-L-565/C 02/17/12508115 CDBI Identified Housekeeping Issues during CCS Walkdown 02/17/12

Service Requests/Problem Evaluation Reports Initiated Due to CDBI Activity NAME TITLE REVISION/ DATE

510299

CDBI-Housekeeping Issue Identified during Walkdown of

AB 737' 02/22/12510729

CDBI-The IPS Floor Drain Trench is Clogged with Trash 02/06/12510956 NRC CDBI Identified -

Generic Letter 89-13 testing requirements not per Program intent 02/23/12511010 CDBI -gap between DS breaker procedure documentation and Westinghouse service life limitations 02/22/12511469 CDBI - Acceptance criteria in 0-MI-57.200 appears to be inconsistent with vendor information. 02/22/12511507 NRC CDBI Identified - Diesel Generator Jacket Water Heat Exchanger Performance Test Results 02/24/12511679 CDBI Identified-Generic Letter 89-13 Testing Requirements not per Program Intent 02/24/12511701 NRC CDBI Identified-Diesel Generator Jacket Water Heat Exchanger Performance Test Results 02/24/12513893 CDBI 2012 Issue Discrepancy between the vendor test recommendations and SI required M&TE 02/27/12514049 CDBI Concern on EOP Transition Points Lacking Specific Guidance on Values for Decision Making 02/29/12514360

NRC-Identified (during CDBI) - Rust Stains on Support Steel for RHR Pump 1B-B Seal Water Cooler 03/01/12514586

NRC-Identified (during CDBI) Issue - Condensation from CCS piping creates water puddles on floor. 03/01/12514730 2012

CDBI-Wording on CCS Heat Exchanger Test Procedures 03/01/12514798 CDBI Question Concerning Number of PER's on MDAFW LCV Behavior 03/01/12514799 2012 CDBI Stow Flexible Remote Operator Concerns Bend Radius Maintenance Clamps 03/01/12515302 CDBI Concern on EOP Transition Points Lacking Specific Guidance on Valves for Decision Making 03/02/12515385 CDBI Issue-Rust Stains on Support Steel for RHR Pump 1B-B Seal Water Heat Exchanger 03/02/12515389

CDBI-No Documentation of Megger Testing of Cable to Connect

MA-1 Diesel Generator 02/02/12515407

CDBI-2012-NRC identified Maintenance Procedure acceptance criteria issues 03/01/12515411 During the 2012 CDBI Inspection the Inspector Determined that 1PCV-3-132 was Returned to Operable Status W/O Meeting

NEDP-22 02/12/12515413 2012 CDBI Issue NRC Identified Harmonic Distortion Issue 03/01/12516562 CDBI - MDAFW Pump 1B-B Anomalous Condition 03/02/12

517095 CDBI -2012-NRC Apparent Discrepancy between 6.9KV Breaker Closure Time 03/06/12605645

CDBI-NRC Identified Degraded Breaker Control Voltage Issues with Maintenance Procedures 02/16/12