Text

IR 05000259-13-007, 05000260-13-007, 05000296-13-007, 05000259-13-404, 05000260-13-404, 05000296-13-404; Tnv; 01/28/2013 - 02/28/2013; Browns Ferry Plant Units 1, 2, and 3, Component Design Bases Inspection Report
	ML13177A443
Person / Time
Site:	Browns Ferry
Issue date:	06/25/2013
From:	; NRC/RGN-II/DRS/EB1
To:	James Shea; Tennessee Valley Authority
References
	IR-13-007, IR-13-404
	Download: ML13177A443 (36)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION II

245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 June 25, 2013 Mr. Joseph Vice President, Nuclear Licensing Tennessee Valley Authority 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801 SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000259/2013007, 05000260/2013007, 05000296/2013007, 05000259/2013404, 05000260/2013404, AND 05000296/2013404

Dear Mr. Shea:

On, April 19, 2013, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Browns Ferry Nuclear Nuclear Plant, Units 1, 2, and 3. The enclosed inspection report documents the inspection results, which were discussed on April 19, 2013, with Mr. Groom and other members of your staff, and with M. Webb on June 3, 2013 The inspection examined activities conducted under your licenses as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your licenses. The team reviewed selected procedures and records, observed activities, and interviewed personnel.

Five NRC-identified findings of very low safety significance (Green), were identified during this inspection, and were determined to involve violations of NRC requirements. One of these NRC identified findings is associated with security and is documented in an attachment to this report.

The NRC is treating these violations as non-cited violations consistent with section 2.3.2 of the NRC Enforcement Policy. If you contest these violations or the significance of the 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 Browns Ferry. If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II; and the NRC Resident Inspector at Browns Ferry.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response, if any, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs 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). However, because of the security-related concerns contained in the enclosure, and in accordance with 10 CFR 2.390, a copy of Attachment 2 will not be available for public inspection.

The material enclosed herewith contains Security-Related Information in accordance with 10 CFR 2.390(d)(1) and its disclosure to unauthorized individuals could present a security vulnerability. Therefore, the material in attachment 2 will not be made available electronically for public inspection in the NRC Public Document Room or from the PARS component of NRC's ADAMS. If you choose to provide a response and Security-Related Information is necessary to provide an acceptable response, please mark your entire response Security-Related Information - Withhold from public disclosure under 10 CFR 2.390 in accordance with 10 CFR 2.390(d)(1) and follow the instructions for withholding in 10 CFR 2.390 (b)(1). In accordance with 10 CFR 2.390(b)(1)(ii), the NRC is waiving the affidavit requirements for your response.

Sincerely,

/RA/

Rebecca L. Nease, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-259, 50-260, and 50-296 License Nos.: DPR-33, DPR-52 and DPR-68

Enclosure:

Inspection Report 05000259/2013007, 05000260/2013007, 05000296/2013007, 05000259/2013404, 05000260/2013404, and 05000296/2013404 w/Attachment 1: Supplementary Information Attachment 2: Security Summary w/Supplementary Information (OUO)

REGION II==

Docket Nos.: 50-259, 50-260, 50-296 License Nos.: DPR-33, DPR-52, DPR-68 Report No.: 05000259/2013007, 05000260/2013007, 05000296/2013007, 05000259/2013404, 05000260/2013404, and 05000296/2013404 Licensee: Tennessee Valley Authority (TVA)

Facility: Browns Ferry Nuclear Plant, Units 1, 2, and 3 Location: Corner of Shaw and Nuclear Plant Roads Athens, AL 35611 Dates: January 28-February 28, 2013 (onsite)

Inspectors: Shakur Walker, Senior Reactor Inspector (Lead)

Geoffrey Ottenberg, Senior Reactor Inspector Robert Patterson, Reactor Inspector Marcus Riley, Reactor Inspector Craig Baron, Accompanying Personnel Stanley Kobylarz, Accompanying Personnel Approved by: Rebecca L. Nease, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY

IR 05000259/2013007, 05000260/2013007, 05000296/2013007, 05000259/2013404, 05000260/2013404, and 05000296/2013404; 1/28/2013-6/3/2013; Browns Ferry Nuclear Plant,

Units 1, 2 and 3; 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. Five Green non-cited violations (NCVs) were identified. The significance of inspection findings are indicated by their color (i.e.,

greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process dated June 2, 2011. Cross-cutting aspects are determined using IMC 0310, Components Within the Cross Cutting Areas dated October 28, 2011. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated January 28, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process revision 4.

NRC identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green: The team identified a non-cited violation of 10 CFR Part 50, Appendix B,

Criterion XI, Test Control, for the failure to ensure that post-maintenance and post-modification testing of the high pressure cooling injection (HPCI) pump adequately demonstrated that it could achieve design basis flow within 30 seconds from a cold, non-oil-primed, turbine quick start under design basis conditions. This was a performance deficiency. The test configuration was less limiting than the design basis accident configuration, and the licensee had not verified by calculation or testing that the acceptance criteria in the test was adequate to demonstrate the HPCI pump could perform its function under design basis conditions. The licensee performed an operability review and documented the results in the corrective action program as Problem Evaluation Report 690086.

The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the HPCI pumps.

Specifically, using procedure 3-SR-3.5.1.7, the licensee failed to demonstrate that the HPCI pump could achieve the required flow and discharge pressure under accident conditions as required by the design basis. Additional analysis was required to verify system operability. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since the original design of the plant and was not indicative of current licensee performance.

(Section 1R21.2.1) 3

Green: The team identified a non-cited violation of 10 CFR 50, Appendix B,

Criterion III, Design Control, involving the failure to evaluate the effects of a postulated failure of the load center transformer non-safety-related, non-Class 1E cooling fans, which includes the fan power wiring and fan control equipment, on the safety-related Class 1E shutdown board load center transformers and 480V shutdown boards. This was a performance deficiency. The licensee tested the fans and performed an operability evaluation as documented in Problem Evaluation Report 682254 to provide reasonable assurance that the safety-related transformers would not be damaged from postulated failures from the non-safety-related fans and be capable of operating when required for the design basis accident conditions.

The performance deficiency was determined to be more than minor because the finding affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the load center transformers TS1A and TS1B and the 480V shutdown boards 1A and 1B respectively. Specifically, the licensee had not evaluated the effects of the failure of non-safety-related transformer cooling fans, on both the safety-related load center transformer and 480V shutdown board and resulted in a reasonable doubt of operability. The team used Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since November 2004; therefore, not indicative of current licensee performance. (Section 1R21.2.10)

Green: The team identified a non-cited violation of 10 CFR 50, Appendix B,

Criterion III, Design Control, for the licensees failure to perform analyses demonstrating that the degraded voltage relay (DVR) set points specified in technical specifications (TS) would ensure adequate voltage to safety-related equipment. This was a performance deficiency. The licensee entered this issue into their corrective action program as PERs 676678 and 696876. As immediate corrective actions, the licensee performed a sensitivity study to verify that the voltage at the DVR set points specified in TS could provide adequate starting voltage to a sample of limiting safety-related equipment.

The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the 4160 volts alternating current buses. Specifically, the finding challenged the assurance that safety-related loads had adequate motor starting voltage during required 4 degraded voltage scenarios. The team used Inspection Manual Chapter 0609,

Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not identified because this performance deficiency has existed since 1993 and was not indicative of current licensee performance. (Section 1R21.2.16)

Green: The team identified a non-cited violation of 10 CFR 50, Appendix B,

Criterion XVI, Corrective Action, for the licensees failure to promptly identify and take corrective actions to address a non-conforming condition adverse to quality related to three faulted strainers in the safety related Emergency Equipment Cooling Water system. This was a performance deficiency. The licensee initiated Problem Evaluation Report 677627 to perform a new operability evaluation since the operability evaluation in Problem Evaluation Report 208636 was found to be inadequate. The licensee concluded that there were no current operability issues.

The performance deficiency was determined to be more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the core spray system to respond to initiating events, in that, if left uncorrected could result in the plant not being able to sustain short-term heat removal under specific conditions. The team used Inspection Manual Chapter 0609,

Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that the finding was of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. The team evaluated the finding for cross-cutting aspects and determined the finding was associated with the corrective action program component of the problem identification and resolution area, because the licensee did not perform a thorough evaluation of identified problems such that the resolutions addressed the underlying causes and extent of condition.

P.1(c) (Section 1R21.4)

Cornerstone: Security

Green: A security finding is documented in Attachment

Licensee-Identified Violations

None.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Security

1R21 Component Design Bases Inspection

Main article: IP 71111.21

.1 Inspection Sample Selection Process

The team selected risk significant components and related operator actions for review using information contained in the licensees probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1 X10-6. The sample included 17 components (including one associated with containment large early release frequency and one associated with security documented in Attachment 2). In addition, the team reviewed six 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.

Documents reviewed are listed in Attachment 1.

.2 Component Reviews

.2.1 Unit 3 High Pressure Coolant Injection (HPCI) Pump

a. Inspection Scope

The team reviewed UFSAR, technical specifications (TS), TS Bases, and System Design Criteria (SDC) documents to establish an overall understanding of the design bases of the HPCI Turbine. The team reviewed design calculations to verify the capability of the HPCI turbine and pump to provide the required flow and head under accident conditions, and to verify that the periodic testing was adequate to demonstrate the capability of the components. The team reviewed operating and test procedures related to the component, as well as recently test results to verify the actual performance of the component. The team also performed a detailed walk-down of the component and related equipment, reviewed 6 HPCI system health reports, and conducted interviews with the system and design engineers to verify the current condition of the component.

b. Findings

Failure to Verify the Capability of HPCI to Achieve Required Flow and Pressure within 30 Seconds Under Accident Conditions

Introduction:

The team identified a finding of very low safety significance (Green)involving a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XI, Test Control, for the failure to ensure that post-maintenance and post-modification testing of the HPCI pump adequately demonstrated that it could achieve design basis flow within 30 seconds from a cold, non-oil-primed, turbine quick start under design basis conditions. The test configuration was less limiting than the design basis accident configuration, and the licensee had not verified by calculation or testing that the acceptance criteria in the test was adequate to demonstrate the HPCI pump could perform its function under design basis conditions.

Description:

The team identified a failure to ensure that testing of the HPCI pump adequately demonstrated that it could achieve 5,000 gpm design flow within 30 seconds from a cold, non-oil-primed, turbine quick start under the most limiting accident conditions. Specifically, UFSAR section 6.4.1 stated, The HPCI controls automatically start the system and bring it to design flow rate within 30 seconds from receipt of a reactor vessel low-low-water-level signal or a primary containment (drywell) high-pressure signal.

Procedure 3-SR-3.5.1.7, rev. 66, was used for post-maintenance testing following HPCI governor control corrective maintenance, and post-modification testing (excluding pump, impeller, or casing replacement). Procedure step 5.2, Technical Specification Requirements, stated that the HPCI System achieve 5,000 gpm flow at a minimum discharge pressure 110 psi above reactor pressure within 30 seconds from a cold, non-oil-primed, turbine quick start. Attachment 3 of the procedure, steps 1.0 [20] and 1.0 [22]

verified that the required pump flow and discharge pressure were achieved in less than or equal to 30 seconds. However, the test was performed under the following conditions, which are less limiting than the design basis conditions: reactor vessel test pressure of 960 - 1035 psig vs. design basis pressure of 1120 psig; and pump discharge test pressure of 1070 - 1145 psig vs. pump discharge design basis pressure of 1220 psig. The licensee did not have a test or calculation that demonstrated that the test as performed, would verify HPCI pump design basis requirements. This condition was applicable to all three units, as well as the Reactor Core Isolation Cooling (RCIC) pumps for all three units In response to this concern, on March 3, 2013, the licensee initiated PER 690086, which addressed the HPCI condition, as well as a similar condition affecting RCIC pump testing. In the PER, the licensee concluded that this condition was not an immediate operability concern based on past test results for both HPCI and RCIC being well-within design basis requirements (30 seconds for HPCI), thus providing some margin between the less-liming test conditions and the design basis conditions.

Analysis:

The team determined that the licensees failure to ensure that post-maintenance and post-modification testing of the HPCI pump adequately demonstrated 7 that it could meet its design basis requirements was a performance deficiency. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the HPCI pumps. Specifically, using procedure 3-SR-3.5.1.7, the licensee failed to demonstrate that the HPCI pump could achieve the required flow and discharge pressure under accident conditions as required by the design basis. Additional analysis was required to verify system operability. The team determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, both issued June 19, 2012. The finding screened as very low safety significance (Green),because it was a design deficiency that did not result in the loss of functionality or operability. The licensee performed an immediate operability determination that demonstrated the HPCI pumps, as well as the RCIC pumps, could achieve design flow.

This operability determination was based on past test results, and was documented in the corrective action program as PER 690086. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of current licensee performance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, 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 the above, since plant startup, the licensee failed to ensure that HPCI post-maintenance and post-modification testing verified acceptance limits contained in applicable design documents. The licensee performed an immediate operability evaluation to provide a reasonable expectation of operability based on actual test data. The licensee initiated PER 609086 to evaluate the finding and determine the appropriate final corrective actions. This violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. The violation was entered in the licensees corrective action program as PER 690086. (NCV 05000259, 260, 296/2013007-01, Failure to Verify the Capability of HPCI to Achieve Required Flow and Pressure Under Accident Conditions)

.2.2 Unit 3 HPCI Steam Isolation Valves (FCV-73-2, FCV-73-3)

a. Inspection Scope

The team reviewed UFSAR, TS, TS Bases, and SDC documents to establish an overall understanding of the design bases of the motor operated HPCI steam isolation valves.

The team reviewed design basis calculations, operating procedures, test procedures, and recent test results to verify the capacity of the valves to perform their design functions. The review included the design pressure differential across the motor operated valves and the available voltage to the valve motors under the most limiting conditions. The team also reviewed HPCI system health reports and interviewed the motor operated valve (MOV) engineer to verify the current condition of the components.

b. Findings

No findings were identified.

.2.3 Unit 3 HPCI Turbine Stop Valve and Governor Valve (FCV-73-18, FCV-73-19)

a. Inspection Scope

The team reviewed UFSAR, TS, TS Bases, and SDC documents to establish an overall understanding of the design bases of the HPCI turbine stop valve and governor valve.

These valves were provided as part of the HPCI turbine package. The team reviewed HPCI design basis calculations, operating procedures, test procedures, and recent test results to verify the capacity of the valves to perform their design functions. The team also reviewed HPCI system health reports and interviewed the HPCI system and design engineers to verify the current condition of the components.

b. Findings

No findings were identified.

a. Inspection Scope

The team reviewed the plants TS, UFSAR, SDC and P&IDs to establish an overall understanding of the design bases of the. The team reviewed the cooling specifications, design bases information and supporting calculations to identify system flow requirements. The team reviewed the procedures and results of the chiller inspections and cleanings, flow balancing and trending to verify that degraded conditions were being appropriately addressed. Component related PERs, corrective maintenance activities, and system health reports were reviewed to evaluate the licensees capability for detection, monitoring, and correcting potential degradation. A field walk-down was performed with the system engineer to assess observable material conditions and verify that the system configuration was consistent with the design basis assumptions, system operating procedures, and plant drawings.

b. Findings

No findings were identified.

.2.10 4160 - 480V Shutdown Board 1A Load Center Transformer TS1A

a. Inspection Scope

The UFSAR was reviewed to establish an overall understanding of the design bases of the shutdown board 1A load center transformer TS1A. The team reviewed load flow and short circuit current calculations to determine the maximum load and short circuit current requirements and vendor documents to verify that transformer ratings were in conformance with the design analyses. The team also reviewed the coordination and protection calculation for the transformer and the shutdown board to verify the adequacy of transformer protection. The team reviewed surveillance tests on the transformer feeder breaker for adequacy of results in accordance with the design basis setting requirements. The team reviewed transformer cooling fan operation and preventive maintenance procedures to verify the capability to satisfy the basis load requirement.

The team reviewed Service Requests (SR) and PERs for recurring issues affecting reliability. The team reviewed alarm response procedures for the transformer to assess the adequacy of operator actions. The team performed a walk-down of the installed equipment to assess the observable material conditions, to verify transformer nameplate data, to determine whether the installed configuration is consistent with design documents including drawings and calculations, and to assess the presence of hazards.

b. Findings

Failure to Evaluate the Effects of the Failure of Non-Class 1E Load Center Transformer Cooling Fans on the Class 1E 4160-480V Load Center Transformers and 480V Shutdown Boards

Introduction:

The team identified a finding of very low safety significance (Green)involving a NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to evaluate the effects of a postulated failure of the non-safety-related (non-10 CFR 50.49-qualified, non-Class 1E) shutdown board transformer cooling system (which includes the fans, fan power wiring, and fan control equipment) on the safety-related (10 CFR 50.49 qualified, Class 1E) shutdown board load center transformers and 480V shutdown boards.

Description:

The team reviewed design change notice 51216, which was initiated to perform electrical work for Browns Ferry Unit 1 Recovery (including work to replace the supply transformers to the shutdown boards). The transformers are located in a harsh high energy line break (HELB) environment outside primary containment and were seismically and environmentally qualified by the vendor, ASEA Brown Boveri, for 1000 kVA for post-accident LOCA and HELB conditions. The TVA staff provided the team with an additional environmental qualification analysis for operating the transformer for up to 1330 kVA (with the shutdown board transformer cooling system operating), to meet the requirements for design basis load flow for post-accident conditions.

However, TVA classified the shutdown board transformer cooling system (hereinafter referred to as transformer cooling system) as non-safety-related, which cannot be credited during accident scenarios.

The team found that TVA General Design Criteria Document, BFN-50-727, Environmental Qualification, Section 5.1.1, stipulated, the evaluation of safety equipment in a harsh environment shall consider effects of all associated devices whether in a mild or harsh environment, safety-related or not (C/R BFNBEII1167). On review of design change notice 51216, the team found that contrary to this requirement, the licensee did not consider and evaluate potential adverse effects of a postulated failure of the non-safety transformer cooling system on the safety-related load center transformer power circuits with the postulated failure of the and the safety-related 480V shutdown board, which powers the transformer cooling system.

The team questioned operability of the shutdown transformers to perform their intended design function. On March 1, 2013, TVA engineering provided a functional evaluation that provided reasonable assurance that the 480V shutdown boards were adequately protected by the safety-related fuse that isolates the shutdown board power from the non-safety-related circuits up to the load center transformer control panel. On April 17, 2013, TVA provided additional information and an engineering evaluation that provided reasonable assurance that the safety-related load center transformers would not be adversely affected by the postulated failure of the transformer cooling system and would be capable of operating as required for the design basis accident conditions.

Analysis:

The team determined that the licensees failure to evaluate the effects of a postulated failure of the non-safety-related shutdown board transformer cooling system, on the safety-related load center transformers TS1A and TS1B and the 480V shutdown boards as required by the licensees General Design Criteria Document was a 12 performance deficiency. The finding also applies to 480V shutdown board transformers TS2A, TS2B, TS3A and TS3B in units 2 and 3. The performance deficiency was determined to be more than minor because the finding affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and operability of the load center transformers TS1A and TS1B and the 480V shutdown boards 1A and 1B respectively. Specifically, prior to the inspection, TVA had not evaluated the effects of the failure of non-safety-related shutdown board transformer cooling system on both the safety-related load center transformer coils and connections and also on the fans safety-related power source that was provided from the 480V shutdown board. This resulted in a reasonable doubt of operability of the transformers to perform their intended design function. The team determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, both issued June 19, 2012.

The finding screened as very low safety significance (Green) because it was a design deficiency that did not that did not result in the loss of functionality or operability. This finding was not assigned a cross-cutting aspect because the issue has existed since November 2004, and the underlying cause was not indicative of current licensee performance.

Enforcement:

The team reviewed the modification to replace the EDG governor to a digital control system to verify that the replacement did not introduce new failure modes that placed the licensee in an unanalyzed condition and that the modification package was consistent with the requirements in 10CFR50.59. The team reviewed completed testing procedures to verify that the governor was capable of meeting the requirements specified in TS and the design basis delineated in the UFSAR. The team performed a walk-down to assess the observable material condition and to determine whether the installed configuration was consistent with design documents and to assess the presence of hazards.

b. Findings

No findings were identified.

.2.16 Units 1 and 2 4160V Shutdown Board Bus B

a. Inspection Scope

The team reviewed the plants TS, UFSAR, system descriptions, and electrical drawings to establish an overall understanding of the design bases for the 4160 VAC emergency shutdown board bus. The team reviewed AC load flow calculations to verify that equipment needed to mitigate a design basis accident had adequate voltage to start and run under accident conditions. The team reviewed functional testing and maintenance procedures to verify that the degraded voltage relays and loss of voltage relays were maintained in an operable condition and operated within the design bases. The team also reviewed uncertainty calculations and completed calibration procedures of the degraded voltage relays to verify that the assumptions in the calculation were conservative with respect to the operation of the relays. A field walk-down of the 4160 VAC shutdown board bus was conducted to assess the observable material condition of the degraded and loss of voltage relays, associated cables, and DC control circuits to determine whether the installed configuration was consistent with design documents and to assess the presence of hazards. The team also conducted interviews with responsible licensee personnel to answer questions that arose during the inspection pertaining to the methodology applied for determining the degraded voltage relay set point.

b. Findings

Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations

Introduction:

The team identified a finding of very low safety significance (Green)involving a NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to perform analyses demonstrating that the degraded voltage relay (DVR) set points specified in TS would ensure adequate voltage to safety-related equipment.

Description:

The purpose of calculation EDQ0057920034, 4.16KV and 480V Busload, Voltage Drop and Short Circuit Calculation, revision 079, was to demonstrate that the design of the Browns Ferry Nuclear Plant Auxiliary Power system was in conformance with the description of the degraded voltage protection configuration described in UFSAR Section 8.4 and to confirm the basis for the degraded voltage set points and time delays. The NRC required all licensees to install degraded voltage protection systems as described in NRC letter dated June 3, 1977, Statement of Staff Positions Relative to Emergency Power Systems for Operating Reactors. Staff Position 1.a of this letter, which the licensee is committed to in UFSAR Section 8.4.8.1.3, states that the selection of voltage and time set points shall be determined from an analysis of the voltage requirements of the safety-related loads at all onsite systems distribution levels.

The DVR settings at Browns Ferry are in accordance with TS Table 3.3.8.1-1 which states the values to be as follows: Allowable Values 3940 VAC and 3900 VAC. The nominal trip set point is 3920 VAC.

The team noted that calculation EDQ0057920034 included a methodology that credited non-safety-related load tap changers to improve voltage to the maximum reset set point of 3983 VAC. The TVA methodology of assuming minimum expected grid voltage and 16 proper operation of non-safety-related load tap changers is acceptable for the purpose of optimizing system voltages for normal operation. However, these assumptions are not appropriate for evaluating the adequacy of the DVR set points with respect to

(1) the starting and running voltage requirements of Class 1E motors, and

(2) the minimum voltage requirements for the most limiting safety related component as delineated in Staff Position 1.a. The licensees failure to perform an analysis at the minimum value (3900 VAC) allowed by TS challenged the assurance that postulated voltages greater than 3900 VAC and less than 3983 VAC would be adequate for safety-related equipment to perform their required safety function during degraded voltage scenarios.

As a result of this concern, the team lacked reasonable assurance that the DVR set points specified in TS could provide the required motor starting voltages for safety-related loads during required degraded voltage scenarios. The licensee entered this issue into their corrective action program as PERs 676678 and 696876. As immediate corrective actions, the licensee performed a sensitivity study to verify that the voltage at the DVR set points specified in TS could provide adequate starting voltage to a sample of safety-related equipment. As a result of this sensitivity study, 19 MOVs had to be re-evaluated to verify that they were capable of performing their required safety function.

Analysis:

The licensees failure to perform analyses demonstrating that the DVR set points specified in TS would ensure adequate voltage to safety-related equipment as required by the design basis was a performance deficiency. This performance deficiency was more than minor because it affected the Design Control attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of the 4160 VAC buses. Specifically, the finding challenged the assurance that safety-related loads had adequate motor starting voltage during required degraded voltage scenarios. The team determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, both issued June 19, 2012. The finding was determined to be of very low safety significance (Green) because the finding was not a design deficiency resulting in the loss of functionality or operability. A cross-cutting aspect was not assigned because this performance deficiency has existed since 1993 and was not indicative of current licensee performance.

Enforcement:

Title 10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, since 1993, the licensee failed to assure that Staff Position 1.a committed to in the UFSAR, was correctly translated into documents used to establish DVR set points. Specifically, Browns Ferry design calculation EDQ0057920034, used to support the TS degraded voltage set points, credited non-safety-related voltage regulation equipment to ensure adequate voltage to all class 1E equipment, in lieu of demonstrating that the set points for the DVRs specified in TS could provide adequate starting voltage to safety-related equipment. The licensee performed a sensitivity study to verify that the voltage at the DVR set points specified in TS could provide adequate starting voltage to a sample of safety-related equipment. As a result of this sensitivity study, 19 MOVs had to be re-evaluated to verify that they were capable of performing their required safety function.

This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. The violation was entered into the licensees corrective action 17 program as PERs 676678 and 696876 to address recurrence. (NCV 05000259, 260, 296/2013007-03, Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations)

.2.17 Security Uninterruptible Power Supply (UPS) and Standby Diesel

See Attachment 2.

.3 Review of Low Margin Operator Actions

.4 Operating Experience (Six Samples)

a. Inspection Scope

The team reviewed seven operating experience issues for applicability at Browns Ferry 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 Information Notice 2010-03, Failures of MOVs Due to Degraded Stem Lubricant 19

NRC Information Notice 2012-16, Preconditioning of Pressure Switches Prior to Surveillance Test

NRC Information Notice 2012-12, HVAC Design Control Issues Challenge Safety System Function

NRC Information Notice 2011-01, Commercial-Grade Dedication Issues Identified During NRC Inspections

NRC Information Notice 2011-22, Instrumentation and Control Module Hardware, Configuration, and Procedure Issues

NRC Browns Ferry Component Design Bases Inspection (CDBI) Report 05000259, 260, 296/2009008

b. Findings

Failure to Promptly Identify and Correct the EECW Strainers Degraded/Non-Conforming Condition

Introduction:

The team identified a finding of very low safety significance (Green)involving a NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and take corrective actions to address a non-conforming condition adverse to quality related to three faulted strainers in the safety related Emergency Equipment Cooling Water (EECW) system.

Description:

During the 2009 NRC CDBI at Browns Ferry (Inspection Report 05000259, 260, 296/2009008), the inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for an inadequate procedure used for flow balancing of the EECW system. That violation stated that the perforated holes in the installed strainers for the EECW system were capable of filtering debris greater than 0.125 inches (1/8 inch); however, allowed debris less than 0.125 inches to pass through into the strainers. The team determined at that time, the inlet throttle valves to the 2A and 2B Core Spray room coolers, 2-THV-067-0551 and 2-THV-067-0594, had disc to seat clearances of less than 0.125 inches. With the clearance less than that of the inlet screen, flow blockage in these valves due to debris passing through the EECW strainers could have occurred resulting in inoperability of these safety related room coolers. In addition, while visually inspecting the EECW strainers, the team identified a number of perforated cone holes that were greater than the 0.125 inches design. Section 10.10 of the UFSAR states that the EECW strainers will have a screen size of 0.125 inches.

These strainer cones are created by cutting and bending/folding a sheet with perforated holes into the cone shape. This process results in a seam in the cone where the two sheet edges join. Due to the fabrication process, there were random locations in some of the strainer cones which would allow two of the 1/8 inch holes to line up on opposite sides of the seam such that an opening is formed where the major axis was larger than 1/8 inch but less than 1/4 inch, and the minor axis less than 1/8 inch. In November 2009, the licensee generated PER 208636 in response to the NCV, which concluded that this condition did not constitute a degraded non-conforming condition without specifically addressing the non-conformance to the current licensing and design basis. Corrective action for PER 208636 included initiation of work orders to observe and replace any affected EECW strainer perforated cones, where the fabrication method of the perforated 20 strainer cone could allow a hole size larger than specified in the UFSAR. By allowing debris greater than the design allowed 1/8 inch, it could impact the Core Spray room cooler valves which were throttled to 1/8 inch. At the time of this inspection, only one of four EECW strainers had been partially corrected.

Based on questions by the inspection team, the licensee initiated PER 677627 to perform a new operability evaluation since it was determined the previous operability evaluation in PER 208636 was inadequate. The licensee concluded that the issue constituted a degraded/non-conforming condition, though there were no current operability issues identified. The licensees basis in PER 677627, Rev. 1 was that there was no history with clogging of the Core Spray room cooler throttled valves since being re-positioned in 2010 after the condition was identified. Additionally, the licensee stated that the radial gap/clearance around the perimeter of the Alvco globe valve disk for both valves is greater than 1/4; therefore, any debris that passed through a strainer cone with the larger gap is capable of passing through the valve opening. Furthermore, turbulent system flows in these throttle valves also tends to flush debris and prevent accumulation of these potentially larger particles.

Analysis:

The licensees failure to promptly identify and take corrective actions to address a non-conforming condition adverse to quality related to three faulted strainers in the safety related EECW system was a performance deficiency. The performance deficiency was determined to be more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the Core Spray system to respond to initiating events, in that, it could result in the plant not being able to sustain short-term heat removal under specific conditions and resulted in the reasonable doubt of operability.

The team determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, both issued June 19, 2012. The finding was determined to be of very low safety significance (Green) because the finding was a design deficiency that did not result in the loss of functionality or operability. The inspectors determined that this finding represented current licensee performance and directly involved the cross-cutting area of Problem Identification and Resolution, component of the Corrective Action Program because the licensee did not perform a thorough evaluation of identified problems such that the resolutions addressed the underlying causes and extent of condition. P.1(c)

Enforcement:

Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, states in part, that measures shall be established to assure that conditions adverse to quality, such as failures, deficiencies, and non-conformances, are promptly identified and corrected.

Contrary to the above, from November 2009 to February 2013, the licensee failed to promptly identify and correct a non-conforming condition adverse to quality, in that three of the strainers on the safety related EECW system allowed greater than the design 1/8 debris to pass through the system, increasing the likelihood of clogging downstream components. The licensee performed an immediate operability evaluation to establish reasonable assurance of operability for the system. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. The violation was entered into the licensees corrective action program as PER 677647. (NCV 05000259, 260, 296/2013007-04, Failure to Promptly Identify and Correct the EECW Strainers Degraded/Non-conforming Condition) 21

4OA6 Meetings, Including Exit

On February 28, 2013, the team presented the inspection results to Mr. Jensen and other members of the licensees staff. On April 19, 2013, the team discussed the results of the inspection with Mr. Groom and members of the licensees staff, and with M. Webb on June 3, 2013. Proprietary information that was reviewed during the inspection was returned to the licensee or destroyed in accordance with prescribed controls.

ATTACHMENT 1:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

Steve Bono, Plant Manager

James Emens, Site Licensing Manager

Kevin Groom, Engineering Team Manager

Tim Mingus, Mechanical Design Lead

Mike Oliver, Site Licensing

Keith Polson, Site Vice President

Don Robertson, Operations

Rick Sampson - Electrical & I&C Lead

Marianne Webb, Site Licensing

NRC personnel

D. Dumbacher, NRC Senior Resident

L. Pressley, NRC Resident Inspector

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

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened and Closed

05000259, 260, 296/2013007- NCV Failure to Verify the Capability of HPCI to Achieve Required Flow and Pressure within 30 Seconds Under Accident Conditions (Section 1R21.2.1)

05000259, 260, 296/2013007- NCV Failure to Evaluate the Effects of the Failure of Non-Class 1E Load Center Transformer Cooling Fans on the Class 1E 4160-480V Load Center Transformers and 480V Shutdown Boards (Section 1R21.2.10)

05000259, 260, 296/2013007- NCV Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations (Section 1R21.2.16)

05000259, 260, 296/2013007- NCV Failure to Adequately Identify, Evaluate, and Correct the EECW Strainers Degraded/Non-

conforming Condition (Section 1R21.4) Attachment 1

LIST OF DOCUMENTS REVIEWED