IR 05000390/2016011

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NRC Component Design Bases Inspection Report 05000390/2016011 and 05000391/2016011
ML16285A217
Person / Time
Site: Watts Bar  Tennessee Valley Authority icon.png
Issue date: 10/06/2016
From: Bartley J
NRC/RGN-II/DRS/EB1
To: James Shea
Tennessee Valley Authority
References
IR 2016011
Download: ML16285A217 (25)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ber 6, 2016

SUBJECT:

WATTS BAR NUCLEAR PLANT UNITS 1 AND 2 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000390/2016011 AND 05000391/2016011

Dear Mr. Shea:

On, August 26, 2016, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Watts Bar Nuclear Plant, Units 1 and 2, and discussed the results of this inspection with Mr. Pry and other members of your staff. Additional inspection results were discussed with Mr. Polickoski on September 26, 2016. The results of this inspection are documented in the enclosed report.

NRC inspectors documented two findings of very low safety significance (Green) and two Severity Level IV findings in this report. These findings involved violations of NRC requirements. The NRC is treating these violations as non-cited violations (NCVs)

consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC resident inspector at the Watts Bar Nuclear Plant.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC resident inspector at the Watts Bar Nuclear Plant. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Jonathan H. Bartley, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-390 and 50-391 License Nos. NPF-90 and NPF-96

Enclosure:

Inspection Report 05000390/2016011 and 05000391/2016011 w/Attachment: Supplementary Information

REGION II==

Docket Nos.: 50-390 and 50-391 License Nos.: NPF-90 and NPF-96 Report Nos.: 05000390/2016011, 05000391/2016011 Licensee: Tennessee Valley Authority (TVA)

Facility: Watts Bar Nuclear Plant, Units 1 and 2 Location: Spring City, TN 37381 Dates: July 25, 2016, to August 26, 2016 Inspectors: J. Eargle, Senior Reactor Inspector (Team Leader)

C. Smith, Reactor Inspector M. Greenleaf, Reactor Inspector H. Leake, Contractor M. Yeminy, Contractor C. Franklin, Trainee Approved by: Jonathan H. Bartley, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY

IR 05000390/2016011 and 05000391/2016011; 07/25/2016 - 08/26/2016; Watts Bar Nuclear Plant,

Units 1 and 2; Component Design Bases Inspection.

This inspection was conducted by a team of four Nuclear Regulatory Commission (NRC)inspectors from Region II and Region IV and two NRC contract personnel. Two Green NCVs and two SL IV NCVs were identified. The significance of the Unit 1 inspection findings is indicated by their color (Green, White, Yellow, Red) using the NRC Inspection Manual Chapter (IMC) 0609,

Significance Determination Process, dated April 29, 2015. The Unit 2 Mitigating Systems cornerstone has not yet transitioned to the Reactor Oversight Process, so the significance of the Unit 2 inspection findings is indicated by their severity level (IV, III, II, I) using traditional enforcement in accordance with IMC 2517 Watts Bar Unit 2 Construction Inspection Program, dated June 6, 2013. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated February 4, 2015. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 201

NRC-Identified and Self-Revealing Findings

Unit 1

Cornerstone: Mitigating Systems

Instructions, Procedures and Drawings, for the licensees failure to have adequate instructions and acceptance criteria in the emergency diesel generator surveillance instructions to ensure that the largest load rejection test bounds the power demand of the largest load. These issues were entered into the licensees corrective action program as condition reports 1201749 and 1199001. The licensee confirmed current operability and determined that likely corrective actions will include revisions to the surveillance instructions.

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

Specifically, the licensees SIs to implement TS SR 3.8.1.9 failed to ensure that the tested kW level of the rejected load bounded the largest predicted post-accident load. The team determined the finding to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of Technical Specification or Non-Technical Specification equipment. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Documentation in the area of Human Performance. [H.7] (Section 1R21.2.b.1)

Design Control, for the licensees failure to properly evaluate the available net positive suction head to the Unit 1 auxiliary feedwater pumps. These issues were entered into the licensees corrective action program as condition reports 1196925 and 1201623. The licensee confirmed current operability and had determined that likely corrective actions will include revisions to the net positive suction head calculation.

The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 74%. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating SSC that maintained its operability. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance. [H.6] (Section 1R21.2.b.2)

Unit 2

Instructions, Procedures and Drawings, for the licensees failure to have adequate instructions and acceptance criteria in the emergency diesel generators surveillance procedures to ensure that the largest load rejection test bounded the power demand of the largest load. These issues were entered into the licensees corrective action program as condition reports 1201749 and 1199001. The licensee confirmed current operability and determined that likely corrective actions will include revisions to the surveillance instructions.

The performance deficiency was determined to be more than minor because it represented an inadequate procedure that, if left uncorrected, could adversely affect the quality of the testing of a safety-related SSC. Specifically, the licensees procedures to implement TS SR 3.8.1.9 failed to ensure that the tested kW level of the rejected load bounded the largest predicted post-accident load. The team determined this finding to be of very low safety significance, SL IV, because it represented a failure to meet a regulatory requirement, including one or more Quality Assurance criteria that had more than minor safety significance. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Documentation in the area of Human Performance. [H.7] (Section 1R21.2.b.3)

Design Control, for the licensees failure to properly evaluate the available net positive suction head to the Unit 2 auxiliary feedwater pumps. These issues were entered into the licensees corrective action program as condition report 1196925. The licensee confirmed current operability and had determined that likely corrective actions will include revisions to the net positive suction head calculation.

The performance deficiency was determined to be more than minor because it represented an inadequate quality oversight function that, if left uncorrected, could adversely affect the quality of the analysis of a safety related SSC. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 57%. The team determined this finding to be of very low safety significance, SL IV, because it represented a failure to meet a regulatory requirement, including one or more Quality Assurance criteria that had more than minor safety significance. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance. [H.6] (Section 1R21.2.b.4)

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R21 Component Design Bases Inspection

Main article: IP 71111.21

.1 Inspection Sample Selection Process

The team selected risk-significant components and related operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included risk significant structures, systems, and components (SSCs) that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1E-6. The sample included 10 SSCs, 1 SSC associated with containment large early release frequency (LERF), and 3 operating experience (OE) items. Additionally, 3 SSCs were selected and reviewed during a partial completion of inspection procedure 71111.21 that was documented in inspection reports 05000390/2015007 and 05000391/2015603.

This brings the total sample of SSCs for this inspection procedure to 13.

The team performed a margin assessment and a detailed review of the selected risk-significant components and associated 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). 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, Inspection Manual Chapter 0326 conditions, NRC Resident Inspector input regarding problem equipment, system health reports, industry OE, and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, OE, and the available defense-in-depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.

.2 Component Reviews

a. Inspection Scope

SSCs

  • Unit 1/2 Shut Down Board Room Heating Ventilation And Air Conditioning
  • Unit 2 Motor Operated Valves 2-FCV-063-025-B and 2-FCV-063-026A
  • Unit 1 Motor Operated Valves 2-FCV-003-116-A/B, 2-FCV-003-126-A/B
  • Unit 1/2 480 Volt Shut Down Boards
  • Unit 1/2 Essential Raw Cooling Water Pump Motors
  • Unit 1/2 Containment Sump & Refueling Water Storage Tank Level Instrumentation
  • Unit 1/2 Eagle 21 Components with LERF Implications

For the 11 components listed above, the team reviewed the plant technical specifications (TS), UFSAR, design bases documents, and drawings to establish an overall understanding of the design bases of the components. Design calculations and procedures were reviewed to verify that the design and licensing bases had been appropriately translated into these documents and that the most limiting parameters and equipment line-ups were used. Logic and wiring diagrams were also reviewed to verify that operation of electrical components conformed to design requirements. Test procedures and recent test results were reviewed against design bases documents to verify the adequacy of test methods and that acceptance criteria for tested parameters were supported by calculations or other engineering documents, and that individual tests and analyses served to validate component operation under accident conditions.

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

Additionally, the team performed the following specific reviews:

  • The team reviewed the Unit 1 and 2 refueling water storage tank level instrumentation and Unit 1 and 2 Eagle 21 process protection system to ensure that it met appropriate equipment qualification.
  • The team reviewed the Unit 1 and 2 containment sump level instrumentation to ensure that it was environmentally qualified in accordance with 10 CFR 50.49.
  • The team reviewed the seismic design calculations of the Unit 1 and Unit 2 auxiliary feedwater piping from the condensate storage tank to ensure that they were done according to licensing commitments.
  • The team reviewed the licensees electrical calculations to ensure that the calculations adequately demonstrated that CSST A or B could meet all of the licensees commitments in order to perform as GDC 17 sources.

b. Findings

.1 Failure To Ensure Adequate Unit 1 Emergency Diesel Generator Surveillance

Instructions

Introduction:

The NRC identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have adequate instructions and acceptance criteria in the emergency diesel generator (EDG) surveillance instructions (SI) to ensure that the largest load rejection test bounded the power demand of the largest load.

Description:

Technical Specification (TS) Surveillance Requirement (SR) 3.8.1.9, required verification of the EDG response following rejection of a load greater than or equal to its associated single largest post-accident load. The licensees SIs, 0-SI-82-3, 18 Month Loss of Offsite Power with Safety Injection Test - DG 1A-A, revision 62, and 0-SI-82-4, 18 Month Loss of Offsite Power with Safety Injection Test - DG 1B-B, revision 62, required verification, in Step 5.4, that the EDGs can reject a load of 640 kW (800 Hp) while meeting the specified voltage and frequency limits. However, the instructions for performing the test, in Section 6.1.5, did not require verification that the rejected load is operating at 640 kW. Additionally, calculation MDQ00299920110380, Evaluation of the Impact of Diesel Generator (DG) Frequency and Voltage Limits, revision 5, issued October 6, 2014, identified that the essential raw cooling water (ERCW) pump motors, which are the largest post-accident loads, can operate as high as 813 HP. This value exceeded the procedures acceptance criterion of 800 HP. Thus, the procedure allowed an EDG rejected load test value that did not bound the magnitude of the generators single largest load (813 HP). These issues were entered into the licensees corrective action program (CAP) as condition reports (CR) 1201749 and 1199001 and the licensee verified current operability by confirming the EDGs were capable of performing as required by SR 3.8.1.9 by evaluating the results of SR 3.8.1.10. The licensee had determined that likely corrective actions will include revisions to the SIs.

Analysis:

The licensees failure to have adequate instructions and acceptance criteria in the EDG SIs to ensure that the largest load rejection test bounded the power demand of the largest load was determined to be a performance deficiency (PD) and violation of Title 10 of CFR Part 50, Appendix B, Criterion V. The PD was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the licensees SIs to implement TS SR 3.8.1.9 failed to ensure that the tested kW level of the rejected load bounded the largest predicted post-accident load.

The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of Technical Specification or Non-Technical Specification equipment.

The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Documentation in the area of Human Performance because the licensee failed to create and maintain complete, accurate, and up-to-date documentation by not maintaining the SIs in alignment with the applicable design output documentation. [H.7]

Enforcement:

Title 10 of CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, required, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, since October 6, 2014, the licensee failed to ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances and failed to ensure that written test procedures incorporated appropriate quantitative acceptance criteria. Specifically, the licensees procedures that implement TS SR 3.8.1.9 failed to ensure that the kW level of the rejected load bounded the largest predicted post-accident load. This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy. The violation was entered into the licensees CAP as CRs 1201749 and 1199001. The licensee had determined that likely corrective actions will include revisions to the SIs. This violation is identified as NCV 05000390/2016011-01, Failure To Ensure Adequate Unit 1 Emergency Diesel Generator Surveillance Instructions.

.2 Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 1 AFW

Pumps

Introduction:

The NRC identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate the available net positive suction head (NPSH) to the Unit 1 auxiliary feedwater (AFW) pumps.

Description:

Calculation EPMJKJ011191, WBN AFW System Pump Net Positive Suction Head Available, revision 11, which was revised for dual unit operation on August 11, 2015, evaluated the available NPSH to the Unit 1 and Unit 2 AFW pumps. The teams review of the calculation revealed that the calculation failed to evaluate the available NPSH for the Unit 1 AFW pumps during the most limiting operational alignment where suction is swapped from the Condensate Storage Tank (CST) to ERCW. Additionally, the team determined that the available NPSH for the Unit 1 pumps during suction from ERCW was not evaluated properly because the calculation assumed that Unit 2 was still in construction.

Without considering an operational Unit 2, the calculation failed to evaluate the Unit 1 AFW system when the flow from the only available train of ERCW must be shared with Unit 2, reducing the available NPSH. The licensee performed an analysis to determine whether the available NPSH was greater than the required NPSH during the swap over process and that the margin was reduced from 12.07 feet to 3.16 feet, or approximately 74%. These issues were entered into the licensees CAP as CRs 1196925 and 1201623 and the licensee had determined that likely corrective actions will include revisions to the NPSH calculation.

Analysis:

The licensees failure to properly evaluate the available NPSH to the Unit 1 AFW Pumps was determined to be a PD and violation of Title 10 of CFR Part 50, Appendix B, Criterion III. The PD was determined to be more than minor because, it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 74%.

The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0609, App. A, The SDP for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating SSC that maintained its operability. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance because the licensee failed to carefully guard and change design margins of the NPSH of the AFW system through a systematic and rigorous process.

[H.6]

Enforcement:

Title 10 of CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, that design control measures shall provide for verifying or checking the adequacy of design. Contrary to the above, since August 11, 2015, the licensee did not implement design control measures to verify the adequacy of the design of the AFW system by failing to evaluate the available NPSH for the Unit 1 AFW pump during the most limiting operational alignment and considering an operational Unit 2. This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy.

The violation was entered into the licensees CAP as CRs 1196925 and 1201623 and the licensee had determined that likely corrective actions will include revisions to the NPSH calculation. This violation is identified as NCV 05000390/2016011-02, Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 1 AFW Pumps.

.3 Failure To Ensure Adequate Unit 2 Emergency Diesel Generator Surveillance

Instructions

Introduction:

The NRC identified a SL IV NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the licensees failure to have adequate instructions and acceptance criteria in the EDG surveillance procedures to ensure that the largest load rejection test bounded the power demand of the largest load.

Description:

TS SR 3.8.1.9 required verification of the EDG response following rejection of a load greater than or equal to its associated single largest post-accident load. The licensees SIs, 0-SI-82-5, 18 Month Loss of Offsite Power with Safety Injection Test - DG 2A-A, revision 38, and 0-SI-82-6, 18 Month Loss of Offsite Power with Safety Injection Test - DG 2B-B, revision 41, required verification, in Step 5.4, that the emergency diesel generators, can reject a load of 640 kW (800 Hp) while meeting the specified voltage and frequency limits. However, the instructions for performing the test, in Section 6.1.5, did not require verification that the rejected load is operating at 640 kW. Additionally, calculation MDQ00299920110380, Evaluation of the Impact of Diesel Generator (DG)

Frequency and Voltage Limits, revision 5, issued October 6, 2014, identifies that the ERCW pump motors, which are the largest post-accident loads, can operate as high as 813 HP. This value exceeded the procedures acceptance criterion of 800 HP. Thus, the procedure allowed an EDG rejected load test value that did not bound the magnitude of the generators single largest load (813 HP). These issues were entered into the licensees CAP as CRs 1201749 and 1199001 and the licensee verified current operability by confirming the EDGs are capable of performing as required by SR 3.8.1.9 by evaluating the results of SR 3.8.1.10. The licensee has determined that likely corrective actions will include revisions to the SIs.

The licensees failure to have adequate instructions and acceptance criteria in the emergency diesel generator surveillance procedure to ensure that the largest load rejection test bounded the power demand of the largest load was determined to be a PD and violation of Title 10 of CFR Part 50, Appendix B, Criterion V. The PD was determined to be more than minor because it represented an inadequate procedure that, if left uncorrected, could adversely affect the quality of the testing of a safety-related SSC.

Specifically, the licensees procedures to implement TS SR 3.8.1.9 failed to ensure that the tested kW level of the rejected load bounded the largest predicted post-accident load.

The team determined this finding to be of very low safety significance, SL IV, in accordance with Section 6.5 of the Enforcement Policy. Specifically, the finding was a SL IV violation because it represented a failure to meet a regulatory requirement, including one or more Quality Assurance criteria that had more than minor safety significance. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Documentation in the area of Human Performance because the licensee failed to create and maintain complete, accurate, and up-to-date documentation by not maintaining the SIs in alignment with the applicable design output documentation.

[H.7]

Enforcement:

Title 10 of CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, required, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, since October 6, 2014, the licensee failed ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances to assure that written test procedures incorporated appropriate quantitative acceptance criteria. Specifically, the licensees procedures that implement TS SR 3.8.1.9 failed to ensure that the kW level of the rejected load bounded the largest predicted post-accident load. This finding was determined to be a SL IV violation using Section 6.5 of the NRC Enforcement Policy. This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy. The violation was entered into the licensees CAP as CRs 1201749 and 1199001. The licensee has determined that likely corrective actions will include revisions to the SIs. This violation is identified as NCV 05000391/

2016011-03, Failure To Ensure Adequate Unit 2 Emergency Diesel Generator Surveillance Instructions.

.4 Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 2 AFW

Pumps

Introduction:

The NRC identified a SL IV NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate the available NPSH to the Unit 2 AFW pumps.

Description:

Calculation EPMJKJ011191, WBN AFW System Pump Net Positive Suction Head Available, Revision 11, which was revised for dual unit operation on August 11, 2015, evaluated the available NPSH to the Unit 1 and Unit 2 AFW pumps. The teams review of the calculation revealed that the calculation failed to evaluate the available NPSH for the Unit 2 AFW pumps during the most limiting operational alignment where suction is swapped from the CST to ERCW. The licensee performed an analysis to determine whether the available NPSH was greater than the required NPSH during the swap over process and determined that the margin was reduced from 3.3 feet to 1.43 feet, or approximately 57%. This issue was entered into the licensees CAP as CR 1196925 and the licensee has determined that likely corrective actions will include revisions to the calculation.

The licensees failure to properly evaluate the available NPSH to the Unit 2 AFW Pumps was determined to be a PD and violation of Title 10 of CFR Part 50, Appendix B, Criterion III. The PD was more than minor because, it represented an inadequate quality oversight function that, if left uncorrected, could adversely affect the quality of the analysis of a safety related SSC. Specifically, the licensees inadequate evaluation of the available NPSH for the AFW pumps resulted in a significant margin reduction of approximately 57%. The team determined this finding to be of very low safety significance, SL IV, in accordance with Section 6.5 of the Enforcement Policy. Specifically, the finding was a SL IV violation because it represented a failure to meet a regulatory requirement, including one or more Quality Assurance criteria that had more than minor safety significance. The team determined the finding was indicative of current licensee performance and assigned a cross-cutting aspect of Design Margin in the area of Human Performance because the licensee failed to carefully guard and change design margins of the NPSH of the AFW system through a systematic and rigorous process. [H.6]

Enforcement:

Title 10 of CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, that design control measures shall provide for verifying or checking the adequacy of design. Contrary to the above, since August 11, 2015, the licensee did not implement design control measures to verify the adequacy of the design of the AFW system by failing to evaluate the available NPSH for the Unit 2 AFW pump during the most limiting operational alignment. This finding was determined to be a SL IV violation using Section 6.5 of the NRC Enforcement Policy. This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy. The violation was entered into the licensees CAP as CR 1196925 and the licensee has determined that likely corrective actions will include revisions to the calculation. This violation is identified as NCV 05000391/2016011-04, Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 2 AFW Pumps.

.5 URI - Common Service Station Transformers A and B General Design Criteria 17

analyses

Introduction:

The team identified an unresolved item (URI) related to the licensees analyses done to evaluate the use of common service station transformers (CSST) A and B as qualified offsite circuits that satisfy general design criteria (GDC) 17. This URI is to determine if a performance deficiency exists.

Description:

The Class 1E power system is normally supplied from offsite power through CSST C and D. Watts Bar applied for and was issued a license amendment to, in part, add an allowance to use CSST A or B as qualified offsite circuits that satisfy GDC 17. During the review of the license amendment, the NRC requested additional information about when CSST A or B is being used as a GDC 17 source and the auxiliary systems for both units are powered from the main generator. By letter dated January 29, 2015, the licensee replied by stating, in part, that their analysis evaluated:

a.) A dual unit trip as a result of abnormal operational occurrence; and c.) Accident in one unit and spurious ESF actuation in the other unit.

The team has requested the licensees analyses for scenarios a.) and c.) from the licensees response and has questions about how the licensee accounted for the voltage drop due to fast transfer of loads from the main generator to CSST A or B. This issue will remain open pending the licensees response to the teams questions and subsequent review in order to determine if a PD exits. This URI is identified as 05000390,391/2016011-05 Common Service Station Transformers A and B General Design Criteria 17 Analyses)

.3 Operating Experience

a. Inspection Scope

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

  • NRC Generic Letter 2006-02: Reliability and the Impact on Plant Risk and the Operability of Offsite Power

b. Findings

No findings were identified.

OTHER ACTIVITIES

4OA6 Meetings, Including Exit

On August 26, 2016, the team presented the inspection results to Mr. Pry and other members of the licensees staff. Additional inspection results were discussed with Mr.

Polickoski on September 26, 2016. The team verified that no proprietary information was retained by the inspectors, or documented in this report.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

G. Pry, Plant Support Director
G. Arent, Director of Site Licensing
J. Polickoski, Senior Program Manager of Corporate Nuclear Licensing
D. Lee, Senior Manager, Design Engineering
J. Ware, Manager, Mechanical Design
R. Cox, Manager, Electrical Design
B. Cusick, Manager, Civil Design
R. Proffitt, Watts Bar Licensing

NRC personnel

J. Bartley, Chief, Division of Reactor Safety
A. Blamey, Chief, Division of Reactor Projects
J. Nadel, Unit 1 Senior Resident Inspector, Division of Reactor Projects
J. Jandovitz, Unit 2 Senior Resident Inspector, Division of Reactor Projects
J. Hamman, Unit 1 Resident Inspector, Division of Reactor Projects.

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened &

Closed

05000390/2016007-01 NCV Failure To Ensure Adequate Unit 1 Emergency Diesel Generator Surveillance Instructions (Section 1R21.2.b.1)
05000390/2016011-02 NCV Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 1 AFW Pumps (Section 1R21.2.b.2)
05000391/2016011-03 NCV Failure To Ensure Adequate Unit 2 Emergency Diesel Generator Surveillance Instructions (Section 1R21.2.b.3)
05000391/2016011-04 NCV Failure To Adequately Evaluate Available Net Positive Suction Head To The Unit 2 AFW Pumps (Section 1R21.2.b.4)
05000390,391/2016011-05 URI Common Service Station Transformers A and B General Design Criteria 17 Analyses (Section 1R21.2.b.5)

LIST OF DOCUMENTS REVIEWED

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