IR 05000259/2013003

IR 05000259-13-003, 05000260-13-003, 05000296-13-003; 04/01/2013 - 06/30/2013; Browns Ferry Nuclear Plant, Units 1, 2 and 3; Equipment Operability Evaluations and Follow-up of Events

ML13226A550
Person / Time
Site:	Browns Ferry
Issue date:	08/14/2013
From:	Shaeffer S M Reactor Projects Region 2 Branch 6
To:	Shea J W Tennessee Valley Authority
References
IR-13-003
Download: ML13226A550 (63)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 August 14, 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 INTEGRATED INSPECTION REPORT 05000259/2013003, 05000260/2013003, AND 05000296/2013003

Dear Mr. Shea:

On June 30, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Browns Ferry Nuclear Plant, Units 1, 2, and 3. The enclosed inspection report documents the inspection results which were discussed on April 12, May 31, July 9, and August 5, 2013, with Mr. K. Polson, Site Vice President, 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, orders, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Two self-revealing findings of very low safety significance (Green) were identified during this inspection. Both of these findings were determined to involve violations of NRC requirements.

Further, one licensee-identified violation which was determined to be of very low safety significance is listed in this report. The NRC is treating the violations as non-cited violations (NCV) consistent with Section 2.3.2 of the Enforcement Policy. If you contest any of 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 20555-0001, with copies to: (1) the Regional Administrator, Region II; (2) the Director, Office of Enforcement, United States Nuclear Regulatory

Commission, Washington, DC 20555-0001; and (3) the NRC Resident Inspector at the Browns Ferry Nuclear Plant.

In addition, if you disagree with any cross-cutting aspect assignment in the 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 the Browns Ferry Nuclear Plant. 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 NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

Sincerely,/Gerald McCoy RA for/

Scott Shaeffer, Chief Reactor Projects Branch 6 Division of Reactor Projects

Docket Nos.: 50-259, 50-260, 50-296 License Nos.: DPR-33, DPR-52, DPR-68

Enclosure:

NRC Integrated Inspection Report 05000259/2013003, 05000260/2013003 and 05000296/2013003

cc: Distribution via Listserv

_________________________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRS SIGNATURE Via email Via telecom Via email Via email Via email Via email Via email NAME DDumbacher LPressley TStephen CStancil PNiebaum KMiller RHamilton DATE 08/14/2013 08/14/2013 08/14/2013 08/14/2013 08/13/2013 08/14/2013 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICE RII:DRS RII:DRS RII:DRP RII:DRS RII:DRS RII:DRP RII:DRP SIGNATURE Via email Via email Via email Via email Via email JXH /RA/ GJM /RA for/ NAME JRivera RKellner MRiches JMontgomery ASengupta JHeisserer SShaeffer DATE 08/14/2013 08/14/2013 08/14/2013 08/13/2013 08/13/2013 08/14/2013 08/14/2013 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO Letter to Joseph from Scott Shaeffer dated August 14, 2013

SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION REPORT 05000259/2013003, 05000260/2013003, AND 05000296/2013003

Distribution w/encl

C. Evans, RII L. Douglas, RII OE Mail RIDSNRRDIRS

PUBLIC RidsNrrPMBrownsFerry Resource Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION II

Docket Nos.: 50-259, 50-260, 50-296

License Nos.: DPR-33, DPR-52, DPR-68

Report No.: 05000259/2013003, 05000260/2013003, 05000296/2013003

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: April 1, 2013, through June 30, 2013

Inspectors: D. Dumbacher, Senior Resident Inspector L. Pressley, Resident Inspector T. Stephen, Resident Inspector C. Stancil, Resident Inspector P. Niebaum, Resident Inspector K. Miller, Resident Inspector R. Hamilton, Senior Health Physicist J. Rivera, Health Physicist R. Kellner, Health Physicist M. Riches, Project Engineer J. Montgomery, Reactor Inspector A. Sengupta, Reactor Inspector

Approved by: Scott Shaeffer, Chief Reactor Projects Branch 6 Division of Reactor Projects Enclosure

SUMMARY

IR 05000259/2013003, 05000260/2013003, 05000296/2013003; 04/01/2013 - 06/30/2013;

Browns Ferry Nuclear Plant, Units 1, 2 and 3; Equipment Operability Evaluations and Follow-up of Events.

The report covered a three month period of inspection by the resident inspectors and regional 6 inspectors. Two self-revealing finding of very low safety significance (Green) were identified. The significance of most findings is identified by their color (Green, White, Yellow, and Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP); and, the cross-cutting aspects were determined using IMC 0310, "Components Within the Cross-Cutting Areas". 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" Revision 4, dated December 2006.

List of Findings and Violations

Cornerstone: Mitigating Systems

Green.

A self-revealing Non-Cited Violation (NCV) of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensee's failure to establish a preventive maintenance program as required by procedure NPG-SPP-06.2, Preventive Maintenance. Specifically, the Residual Heat Removal Service Water Pump D1 Cross-Tie to Emergency Equipment Cooling Water (EECW) Valve (0-FCV-067-0048) was not maintained in a manner that ensured it would perform its design function. The valve was replaced on January 16, 2013, with a new valve with a stainless steel disk. Corrective actions were planned to develop preventive maintenance activities for this valve. The licensee entered this issue into their corrective action program.

This issue was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone objective and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the 0-FCV-067-0048 valve failed and could not perform its isolation function. The finding was previously characterized as "To Be Determined (TBD)" in the Browns Ferry inspection report number 05000259, 260, 296/2013002, dated May 14, 2013.

The issue was screened per IMC 0609, Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, and was determined to be a Green finding because it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensee's maintenance rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The cause of this finding was directly related to the cross-cutting aspect of Appropriately Coordinating Work Activities in the Work Control component of the Human Performance area, because maintenance activities for 0-FCV-067-0048 were more reactive than preventive. H.3(b), (Section 1R15.1)

Green.

A self-revealing non-cited violation (NCV) of Technical Specifications 3.5.1, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System was identified associated with the licensee's failure to perform a post maintenance test (PMT) as required by licensee procedure NPG-SPP-06.3, Pre-/Post Maintenance Testing. Specifically, the licensee failed to perform a PMT to verify the proper operation of the 2E RMOV Board alternate feeder breaker on July 10, 2012. The licensee entered this issue as an immediate corrective action into their corrective action program and the failed alternate feeder breaker to the Unit 2 2E 480V RMOV Board trip pushbutton was restored to the proper condition on July 19, 2012, at 0303.

This finding was more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the Unit 2, 2E 480V RMOV Board alternate feeder breaker was left in a tripped condition and unable to close for nine days. The issue was screened per IMC 0609, Appendix A, Exhibit 2 -

Mitigating Systems Screening Questions, and was determined to be a Green finding because it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensee's maintenance rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The cause of this finding was associated with the Human Performance area, Work Practices component, cross-cutting aspect of appropriate oversight of work activities because maintenance work activities for the alternate feeder breaker for the 2E 480V RMOV board did not include a check to ensure the breaker's nuclear safety function was supported. H.4(c). (Section 4OA3.2) Violations of very low safety significance or Severity Level IV that were identified by the licensee have been reviewed by the NRC. Corrective ac tions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 operated at full Rated Thermal Power (RTP) except for a planned downpower to 75 percent from June 15-17, 2013, for rod sequence exchange, scram time testing, and main condenser water box cleaning. The unit remained near full power the remainder of the quarter.

Unit 2 started the quarter shutdown due to refueling outage 17. The unit was restarted on May 4, 2013 and synchronized to the grid on May 5, 2013. Unit 2 operated at full RTP except for three planned down powers and two unplanned downpowers. On May 8, 2013, a planned downpower from 87 percent to 71 percent was performed for rod pattern adjustment. On May 19, 2013, a planned downpower to 85 percent was performed for reactor feedwater pump maintenance. On June 8, 2013, a planned down power to 75 percent was performed for rod pattern adjustment. On June 12, 2013, an unplanned downpower occurred to 94 percent due to feedwater heater level controller problems. On June 15, 2013, an unplanned downpower occurred to 97 percent due to automatic isolation of extraction steam to B1 and B2 high pressure feedwater heaters. The unit remained near full power the remainder of the quarter.

Unit 3 operated at full RTP except for five planned downpowers. On April 18, 2013, and April 20, 2013, planned downpowers to 95 percent were performed to remove and restore extraction steam from A1/A2 high pressure heaters for maintenance. From June 15-16, 2013, a planned downpower was performed to 92 percent for main condenser waterbox flushing. From June 22-23, 2013, a planned downpower was performed to 72 percent for main condenser waterbox flushing and cleaning. On June 25, 2013, a planned downpower to 80 percent was performed for main condenser waterbox cleaning and rod sequence exchange. The unit remained near full

power the remainder of the quarter.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

.1 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On April 11, 2013, a Tornado Warning was declared for Morgan County, approximately ten miles southeast of the power plant. The inspectors reviewed the licensee's overall preparations/protection for the unexpected onset of severe weather conditions and observed the licensee's implementation of abnormal operating instruction AOI 100-7, Severe Weather, including implementation of actions for Tornado Watch. Inspectors observed applicable contingency actions associated with Unit 2 in a refueling outage and potential elevation to shutdown risk level Orange. The inspectors also reviewed and discussed the implementation of 0-AOI-100-7 with the responsible Unit Supervisors and Shift Manager. Furthermore, the inspectors witnessed the licensee's execution of evacuation orders of vulnerable areas and buildings outside the power block, and the termination of work and evacuation of the turbine floor. The inspectors also toured the plant grounds for loose debris, which could become missiles during a tornado, and reviewed operator staffing and their accessibility to controls and indications for those systems required for safe control of the plant. This activity constituted one inspection sample.

b. Findings

No findings were identified.

.2 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

Prior to and during the onset of hot weather conditions, the inspectors reviewed the licensee's implementation of 0-GOI-200-3, Hot Weather Inspection, including applicable checklists - Attachment #1, Hot Weather Prep Annual Checklist; Attachment #2, and Hot Weather Operational Checklist; Attachment #3. The inspectors also reviewed the Hot Weather Discrepancy Log (PA-104); and discussed implementation of 0-GOI-200-3 with responsible Operations personnel and management. Inspectors also attended a Hot Weather readiness meeting. Furthermore, the inspectors monitored the status of risk significant equipment to cool the Unit 1, 2, and 3 control rooms and the status of the 7 cooling towers on site. This activity constituted one inspection sample.

b. Findings

No findings were identified.

.3 Offsite and Alternate AC Power Systems Readiness

a. Inspection Scope

Prior to the summer season, inspectors reviewed electrical power design features, and onsite risk and work management procedures to verify appropriate operational oversight and assurance of continued availability of offsite and alternate AC power systems. Inspectors verified that communications protocols existed between the transmission system operator and Browns Ferry Nuclear Plant for coordination of off-normal and emergency events affecting the plant, event details, estimates of return-to-service times, and notifications of grid status changes. Inspectors also reviewed procedures to verify that they included controls to adequately monitor both offsite AC power systems (including post-trip voltages) and onsite alternate AC power systems for availability and reliability. Furthermore, inspectors interviewed onsite licensed operators to determine their understanding and implementation of the power monitoring and assessment process. Inspectors reviewed the material condition of offsite AC power systems and onsite alternate AC power systems to the plant, including switchyard and transformers. This review included review of outstanding work orders affecting these systems and a walkdown of the switchyard with operati ons personnel to ensure the systems will continue to provide appropriate "as designed" capabilities. This activity constituted one inspection sample.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial Walkdown

a. Inspection Scope

The inspectors conducted five partial equipment alignment walkdowns to evaluate the operability of selected redundant trains or backup systems, listed below, while the other train or subsystem was inoperable or out of service. The inspectors reviewed the functional systems descriptions, Updated Final Safety Analysis Report (UFSAR), system operating procedures, and Technical Specifications (TS) to determine correct system lineups for the current plant conditions. The inspectors performed walkdowns of the systems to verify that critical components were properly aligned and to identify any discrepancies which could affect operability of the redundant train or backup system.

This activity constituted five Equipment Alignment Partial Walkdown inspection samples.

• Unit 2 RCIC system walkdown following pressure transient to 1337 psig on May 6, 2013

• Unit 1 Core Spray (CS) System, Loop II with Loop I out of service for Maintenance, May 9, 2013

• Standby Gas Treatment (SBGT) trains A and B with train C without an emergency backup power source due to Emergency Diesel Generator 3D maintenance, May 17, 2013

• Units 1, 2 and 3 Emergency Diesel Generators on May 28, 2013, (PER 731716, Diesel Generator A Fuel Pressure Sensing Line Broken)

• Unit 2 Residual Heat Removal (RHR) System, Loop II with Loop I out of service for Maintenance, May 30, 2013

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Routine Walkdowns

a. Inspection Scope

The inspectors reviewed licensee procedures, Nuclear Power Group Standard Programs and Processes (NPG-SPP)-18.4.7, Control of Transient Combustibles, and NPG-SPP-18.4.6, Control of Fire Protection Impairments, and conducted a walkdown of five fire areas (FA) and fire zones (FZ) listed below. Selected FAs/FZs were examined in order to verify licensee control of transient combustibles and ignition sources; the material condition of fire protection equipment and fire barriers; and operational lineup and operational condition of fire protection features or measures. Also, the inspectors verified that selected fire protection impairments were identified and controlled in accordance with procedure NPG-SPP-18.4.6. Furthermore, the inspectors reviewed applicable portions of the Fire Protection Report, Volumes 1 and 2, including the applicable Fire Hazards Analysis, and Pre-Fire Plan drawings, to verify that the necessary firefighting equipment, such as fire extinguishers, hose stations, ladders, and communications equipment, was in place. This activity constituted five inspection samples.

• Unit 1 Reactor building elevations 519 - 565' West (Fire Zone 1-1)

• Unit 1 Reactor building elevations 519 - 565' East (Fire Zone 1-2)

• Unit 3 Reactor Building elevations 519 - 565' East (Fire Zone 3-2)

• Unit 3 4KV Shutdown Board Room 3EC and 3ED, Diesel Generator Building, elevations 565' and 583' (Fire Area 23)

• Unit 3 4KV Bus Tie Board Room, Unit 3 Diesel Generator Building, EL 565' (Fire Area 24)

b. Findings

No findings were identified.

1R07 Heat Sink Performance

a. Inspection Scope

Unit 2 Residual Heat Removal (RHR) Heat Exchangers:

The inspectors examined activities associated with Unit 2 RHR Heat Exchangers. The inspectors also reviewed procedures used for testing flow rates; and reviewed design basis documents, calculations, test procedures, and results to evaluate the licensee's program for maintaining heat sinks in accordance with the licensing basis. Furthermore, the inspectors reviewed PERs and corrective actions to verify that the licensee was identifying issues and correcting them.

The inspectors performed walkdowns of key components of RHRSW systems to verify material conditions were acceptable and physical arrangement matched procedures and drawings. The inspectors observed replacement of the floating head and cleaning and inspection activities associated with the Unit 2 "A" RHR heat exchanger. Inspectors reviewed licensee compliance to commitments made based on their response to the NRC Generic Letter 89-13 for service water system problems that could affect heat exchanger performance. A review of previous licensee inspections and fouling determination testing of the Unit 2 heat exchangers was performed. Licensee corrosion and mollusk control chemical addition processes for heat exchangers were also reviewed. This activity constituted one Heat Sink Performance Inspection sample.

b. Findings

No findings were identified.

1R11 Licensed Operator Requalification

.1 Resident Inspector Quarterly Review

a. Inspection Scope

The inspectors observed licensed operator performance during requalification testing and training following several weeks performing shift-work away from a training environment. The inspectors also observed the TVA training staff adequately conducting this training and correcting performance deficiencies. The inspectors also observed the fidelity of the simulator environmen t verifying the similarity to the actual plant control rooms. This activity constituted two Resident Inspector Quarterly Inspection samples.

• May 13, 2013 observed Group 1 Crew D

• May 20, 2013 observed Group 3 Crew A

b. Findings

No findings were identified.

.2 Control Room Observations

a. Inspection Scope

Several times during the inspection quarter the inspectors observed and assessed licensed operator performance in the plant and main control room, particularly during periods of heightened activity or risk and where the activities could affect plant safety.

The inspectors reviewed various licensee policies and procedures such as OPDP-1, Conduct of Operations, NPG-SPP-10.0, Plant Operations and GOI-100-12, Power Maneuvering.

The inspectors utilized activities such as post maintenance testing, surveillance testing and refueling and other outage activities to focus on the following conduct of operations as appropriate;

• Operator compliance and use of procedures.

• Control board manipulations.

• Communication between crew members.

• Use and interpretation of plant instruments, indications and alarms.

• Use of human error prevention techniques.

• Documentation of activities, including initials and sign-offs in procedures.

• Supervision of activities, including risk and reactivity management.

• Pre-job briefs.

This activity constituted one Control Room Observation inspection sample.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

.1 Routine

a. Inspection Scope

The inspectors reviewed the specific structures, systems and components (SSC) within the scope of the Maintenance Rule (MR) (10CFR50.65) with regard to some or all of the following attributes, as applicable:

(1) Appropriate work practices;

(2) Identifying and addressing common cause failures;

(3) Scoping in accordance with 10 CFR 50.65(b) of the MR;

(4) Characterizing reliability issues for performance monitoring;

(5) Tracking unavailability for performance monitoring;

(6) Balancing reliability and unavailability;

(7) Trending key parameters for condition monitoring;

(8) System classification and reclassification in accordance with 10 CFR 50.65(a)(1) or (a)(2);

(9) Appropriateness of performance criteria in accordance with 10 CFR 50.65(a)(2); and

(10) Appropriateness and adequacy of 10 CFR 50.65 (a)(1) goals, monitoring and corrective actions (i.e., Ten Point Plan). The inspectors also compared the licensee's performance against site procedure NPG-SPP-3.4, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting; Technical Instruction 0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting; and NPG-SPP 3.1, Corrective Action Program. The inspectors also reviewed, as applicable, work orders, surveillance records, PERs, system health reports, engineering evaluations, and MR expert panel minutes; and attended several MR expert panel meetings to verify that regulatory and procedural requirements were met. This activity constituted two Maintenance Effectiveness inspection samples.

• May 23, 2013, Unit 2 RHR system Shift from a(1) status to a(2) status

• June 13, 2013, Units 1, 2, 3 degraded temperature indications for the control rod drive mechanisms

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

For planned online work and/or emergent work that affected the combinations of risk significant systems listed below, the inspectors examined four on-line and one off-line maintenance risk assessments, and actions taken to plan and/or control work activities to effectively manage and minimize risk. The inspectors verified that risk assessments and applicable risk management actions (RMA) were conducted as required by 10 CFR 50.65(a)(4) applicable plant procedures, and BFN Equipment to Plant Risk Matrix. Furthermore, as applicable, the inspectors verified the actual in-plant configurations to ensure accuracy of the licensee's risk assessments and adequacy of RMA implementations. This activity constituted five Maintenance Risk Assessment inspection samples.

• April 11, 2013; Entry into 0-AOI-100-7, Severe Weather for Tornado Watch with Unit 2 in Refueling Outage with both Shutdown Cooling Loops, Unit 1 Residual Heat Removal Loop 1, A3 RHR Service Water Pump, both Service Air Compressors, and D Control Air Compressor out-of-service (OOS)

• April 22, 2013; Entry into 2-POI-200.5 Operations with the Potential to Drain the Reactor Vessel (OPDRV) for maintenance that was performed on the Unit 2 2A and 2B Recirculating Water Pump seals and the 2-FCV-69-1 Reactor Water Clean-Up (RWCU) pump suction inboard isolation valve.

• May 20, 2013; Entry into the extended timeframe of the Emergency Diesel Generator (EDG) 3D outage to include availability of the other 7 EDGs and two Temporary Diesel Generators.

• May 23, 2013; EDG 3D, 3EN LPCI MG Set, A3 RHRSW/EECW, A Control Bay Chiller, all out of service with 3E RMOV BD to ALT.

• May 29, 2013; Unit 2 Residual Heat Removal (RHR) Loop I and A Control Bay Chiller out of service.

b. Findings

No findings were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the operability/functional evaluations listed below to verify technical adequacy and ensure that the licensee had adequately assessed Technical Specification operability. The inspectors also reviewed applicable sections of the Updated Final Safety Analysis Report (UFSAR) to verify that the system or component remained available to perform its intended function. In addition, where appropriate, the inspectors reviewed licensee procedure NEDP-22, Functional Evaluations, and NEDP-27, Past Operability Evaluations, to ensure that the licensee's evaluation met procedure requirements. Furthermore, where applicable, inspectors examined the implementation of compensatory measures to verify that they achieved the intended purpose and that the measures were adequately controlled. The inspectors also reviewed PERs on a daily basis to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. This activity constituted seven Operability Evaluation inspection samples.

• Unit 2 Residual Heat Removal (RHR) heat exchanger 2A past operability due to excessive Asiatic clams, April 14, 2009

• Unit 2 Emergency Equipment Cooling Water (EECW) leak on reactor building El 593 feet near the A and C RHR Heat Exchanger Rooms, April 3, 2013

• Unit 2 High Pressure Coolant Injection (HPCI) Main Pump Shaft Repair (PER 713047 / WO 114609716) April 18, 2013

• Unit 3 High Pressure Coolant Injection grease leak on coupler between booster pump and reduction gear box, April 27, 2013

• Unit 2 RCIC system pressure transient during normal quarterly surveillance

• RHR Service Water Pump D2 coupling bolts not lubricated prior to pump reassembly, May 21, 2013

• Various RCIC, RWCU, and HPCI MOV stroke time requirements used nominal voltage and design packing loads versus calculated DC voltage and measured packing loads, May 24, 2013.

.1 Failure to Implement Preventive Maintenance Program (closeout of AV 05000259, 260,

296/2013002-001)

Introduction:

A self-revealing Green non-cited violation (NCV) of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensee's failure to establish a preventive maintenance program to maintain the 0-FCV-067-0048, RHR Service Water Pump D1 Cross-Tie to Emergency Equipment Cooling Water (EECW) Valve, in a manner that ensured it would perform its design function as required by licensee procedure NPG-SPP-06.2, Preventive Maintenance.

Description:

The 0-FCV-067-0048 valve was a quarter-turn butterfly valve with a motor operator that allowed remote operation of the valve. The valve was the RHRSW/EECW cross tie valve installed in the discharge piping between the D1 RHRSW pump and the D3 EECW pump. The valve was required to be closed to maintain a boundary between these two interfacing systems. The licensee determined that the valve had not been replaced since its original installation during construction in August 1974. An inspection on January 10, 2013, revealed the 0-FCV-067-0048 cast iron valve disc separated from the valve stem with pieces of the valve disc found in the downstream pipe at the inlet of the 'D' EECW strainer. The licensee's root cause report (PER 671314) determined the direct cause of the valve failure was the cumulative effects of age and the pressure transients in the system. Licensee procedure NPG-SPP-06.2, Preventive Maintenance paragraph 3.2.1.B, required preventive maintenance programs to be structured to maintain components in a manner that permits them to perform their design functions.

According to BFN-50-7067, General Design Criteria Document for EECW, a design function of the 0-FCV-067-0048 valve was to isolate the EECW system from interfacing systems when necessary so that the EECW system may perform its required nuclear safety function. The inspectors concluded the lack of preventive maintenance prevented identification of valve degradation. No internal preventative maintenance was performed on the valve and the valve was left in service until it catastrophically failed and could not perform its design function. This valve failure resulted in unavailability of the D3 EECW pump and required the D1 RHRSW pump be aligned to supply the South EECW header.

Analysis:

The licensee's failure to establish a preventive maintenance program to ensure the 0-FCV-067-0048 valve would perform its design function as required by procedure NPG-SPP-06.2, Preventive Maintenance, was a performance deficiency.

This finding was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone objective and adversely affected the cornerstone objective to ensure availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the 0-FCV-067-0048 valve failed and could not perform its isolation function. The inspectors initially evaluated the significance of the finding using Inspection Manual Chapter (IMC) 0609 Appendix F, "Fire Protection Significance Determination Process" and assigned the finding a Moderate Degradation rating. Additional information provided by the licensee indicated that although the capability of the EECW system to provide flow to components required for safe shutdown during a fire scenario was reduced; the system was capable of providing sufficient flow for equipment to perform their intended safety functions. Therefore the performance deficiency was determined not to affect the ability to reach and maintain safe shutdown conditions in case of a fire and was it more appropriate to assess the risk significance of this finding using IMC 0609, Appendix A. The issue was screened using IMC 0609, Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, and was determined to be a Green finding because it did not represent an actual loss of function of one or more non-Technical Specification Trains of equipment designated as high safety-significant in accordance with the licensee's maintenance rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The cause of this finding was directly related to the Human Performance area Work Control component cross-cutting aspect of Appropriately Coordinating Work Activities because maintenance activities for 0-FCV-067-0048 were more reactive than preventive, H.3(b).

Enforcement:

10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, required, in part, that activiti es affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures. Procedure NPG-SPP-06.2, Preventive Maintenance, prescribed activities affecting quality related to the implementation of the preventive maintenance program. NPG-SPP-06.2 paragraph 3.2.1.B required the preventive maintenance program be structured to maintain

components in a manner that permits them to perform their design functions. Contrary to the above, since the original installation of 0-FCV-067-0048, the licensee failed to implement a preventive maintenance program as prescribed by NPG-SPP-06.2, Preventive Maintenance. Specifically, the licensee failed to ensure the preventive maintenance program was structured to maintain 0-FCV-067-0048 in a manner that

permitted it to perform its design function. This issue was entered in the licensee's corrective action program as problem event report (PER) 671314 and 735824. The failed valve was replaced on January 16, 2013, with a stainless steel disk. Further corrective actions were planned to develop adequate preventive maintenance activities for this valve. This finding was identified as a non-cited violation: NCV 05000259, 260, 296/2013002-01, Failure to Implement an Adequate Preventive Maintenance Program.

.2 Unresolved Item (URI).

The NRC identified an issue of concern associated with the April 14, 2009 2A Residual Heat Removal (RHR) Heat Exchanger (HX) raw water side inspection. It was determined that the licensee had failed to identify a degraded non-conforming condition of excessive fouling by Asiatic clams that exceeded the established tube plugging criteria.

Description:

The NRC working with organizations such as the Electric Power Research Institute (EPRI) had provided information to nuclear power plant license holders regarding fouling issues with raw water heat exchangers. In 1989, the NRC issued

Generic Letter (GL) 89-13 to provide approved means to address heat exchanger fouling. Browns Ferry's response to GL 89-13, originally committed the licensee to perform once per cycle inspection and cleaning of the RHR HXs and periodically perform thermal performance testing to prove continued operability per TVA letter dated March 16, 1990. Once per cycle during this period in time was once per 18 months. In this letter, TVA reported that they had observed issues with Asiatic clam fouling in their HXs.

Browns Ferry later shifted the inspection and cleaning frequency to once every two years and then to once every four years without providing a basis or thermal performance testing of all the RHR HXs.

On April 14, 2009, Browns Ferry performed a GL 89-13 inspection of the 2A Residual Heat Removal (RHR) Heat Exchanger (HX). TVA inspections documented results which included pictures of the raw water side of the HX. The pictures showed 245 tubes had shells that were either partially or fully obstructed by shells. This exceeded the maximum calculated 77 clogged tubes needed to maintain the HX operable. The TVA inspections failed to adequately document the visible Asiatic clam fouling, and did not enter the degraded non-conforming condition into the corrective action program, as required by NPG-SPP-03.1 Corrective Action Program section 3.1.R. This would have necessitated an operability evaluation, as required by NPG-SPP-09.14 Generic Letter 89-13 Implementation procedure.

On April 24, 2013, the NRC challenged past operability of the 2A RHR HX from 2005 to 2009. The licensee stated that implementing a new revision to the NPG-SPP 09.14 procedure would prove operability. The update to the NPG-SPP 09.14 procedure for analyzing clam fouling was inadequate to prove operability. TVA then utilized a new (2012) Electric Power Research Institute (EPRI) analysis for determination of the effect of fouling on heat exchanger tubes to determine that the 2A RHR HX remained operable from 2005 to 2009. This method gives full heat transfer credit to any tube not greater than 85 percent clogged.

Additional inspection of the licensee's application of the EPRI guidance and operability

determinations is required to provide additional information needed for the NRC to disposition this issue. This unresolved item was tracked as URI 05000259, 260, 296/2013003-03.

1R18 Plant Modifications

.1 Permanent Plant Modifications

a. Inspection Scope

The inspectors reviewed the maintenance and repair of the Unit 2 High Pressure Coolant Injection (HPCI) turbine pump shaft (WO 114609716) including related documents and procedures. The inspectors reviewed licensee procedures NPG-SPP-09.3, Plant Modifications and Engineering Change Control, NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments, NPG-SPP-09.20, Vendor Manual Control and NPG-SPP-06.9.3, Post-Modification Testing, and observed part of the licensee's activities to implement this possible modification.

The inspectors reviewed the associated 10 CFR 50.59 screening against the system design bases documentation to verify that the modifications had not affected system operability/availability. The inspectors reviewed selected ongoing and completed work activities to verify that installation was consistent with the design control documents. This activity constituted one Permanent Plant Modification sample.

b. Findings

No findings were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors witnessed and reviewed the eight post-maintenance tests (PMT) listed below to verify that procedures and test activities confirmed Structures, Systems, and Components (SSC) operability and functional capability following the described maintenance. The inspectors reviewed the licensee's completed test procedures to ensure any of the SSC safety function(s) that may have been affected were adequately tested, that the acceptance criteria were consistent with information in the applicable licensing basis and/or design basis documents, and that the procedure had been properly reviewed and approved. The inspectors also witnessed and/or reviewed the test data, to verify that test results adequately demonstrated restoration of the affected safety function(s). The inspectors verified that PMT activities were conducted in accordance with applicable WO instructions, or licensee procedural requirements. Furthermore, the inspectors verified that problems associated with PMTs were identified

and entered into the CAP. This activity constituted eight Post Maintenance Test inspection samples.

• April 24, 2013, Unit 2 Core Spray Loop I Comprehensive Testing following hand-switch replacement and valve disassembly and inspection, WO 113870368

• April 26, 2013, Unit 2 Leakage Test of the Reactor Pressure Vessel and Associated Piping per 2-SI-3.3.1.A, following Unit 2 Outage, WO 113759995

• April 29, 2013, Unit 2 Residual Heat Removal (RHR) Loop I auto initiation test following maintenance performed during Refueling Outage 17, 2-SR-3.5.1.9 (RHR I)

• April 30, 2013, Unit 2 Reactor Core Isolation Cooling (RCIC) System Rated Flow at Low RPV Pressure following repairs to the RCIC Turbine Exhaust hand control valve (2-71-14), 2-SR-3.5.3.4 (RCIC)

• April 29, 2013, Unit 2 Standby Liquid Control (SLC) System Pump Functional Test, 2-SR-3.1.7.7

• May 5, 2013, Unit 2 High Pressure Coolant Injection System comprehensive pump testing following system maintenance and modifications performed during Unit outage, 2-SR-3.5.1.7(COMP)

• May 10, 2013, Unit 1 Core Spray Flow Rate Loop I following preventative maintenance items performed, 1-SR-3.5.1.6(CS1)

• May 24, 2013, Unit 3 Emergency Diesel Generator 3C Load Test following completion of the modifications to the lube oil system, WO 114747436

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

.1 Unit 2 Scheduled Refueling Outage (U2R17)

a. Inspection Scope

From March 14 to May 4, 2013, the inspectors examined critical outage activities to verify that they were conducted in accordance with Technical Specifications (TS), applicable procedures, and the licensee's outage risk assessment and management plans. Activities observed March 14 through March 31 are documented in Inspection Report 2013-002. Significant inspection activities conducted by the inspectors were as follows: Outage Risk Assessment The inspectors also reviewed the daily U2R17 Refueling Outage Reports, including the Outage Risk Assessment Management (ORAM) Safety Function Status, and regularly attended the twice a day outage status meetings. These reviews were compared to the requirements in licensee procedure NPG-SPP-07.2, Outage Management, and TS. These reviews were also done to verify that for identified high risk significant conditions, due to equipment availability and/or system configurations, contingency measures were identified and incorporated into the overall outage and contingency response plan.

Furthermore, the inspectors frequently discussed risk conditions and designated protected equipment with Operations and outage management personnel to assess licensee awareness of actual risk conditions and mitigation strategies. The inspectors reviewed licensee compliance with 10 CFR 26 Nuclear Fatigue Rule through interviews with the site coordinator and reviews of 12 fatigue assessments and hours worked for 11 TVA employees and contractors.

Decay Heat Removal The inspectors reviewed licensee procedures 2-OI-74, Residual Heat Removal System (RHR); 2-OI-78, Fuel Pool Cooling and Cleanup System; and Abnormal Operating Instruction 0-AOI-72-1, Alternate Decay Heat Removal System Failures; and conducted a main control room panel and in-plant walk downs of system and components to verify correct system alignment. During planned evolutions that resulted in an increased outage risk condition of "Orange" for shutdown cooling, inspectors verified that the plant conditions and systems identified in the risk mitigation strategy were available. In addition, the inspectors reviewed controls implemented to ensure that outage work was not impacting the ability of operators to operate spent fuel pool cooling, RHR shutdown cooling, and/or Alternate Decay Heat Removal (ADHR) system. Furthermore, the inspectors conducted several walkdowns of the ADHR system during operation with the fuel pool gates removed.

Critical Outage Activities The inspectors examined outage activities to verify that they were conducted in accordance with TS, licensee procedures, and the licensee's outage risk control plan.

Some of the more significant inspection activities accomplished by the inspectors were as follows:

• Walked down selected safety-related equipment clearance and associated with tagout numbers.

o 2-071-0010C, 2-071-0012A, 2-071-0013, 2-071-0023D, 2-071-0029A, 2-071-0035; Reactor Core Isolation Cooling System o 2-073-0011A, 2-073-0012B, 2-073-0024, 2-073-0026; High Pressure Coolant Injection System o 2-074-0027; Residual Heat Removal System o 2-075-0003A; Core Spray System o 2-074-0018; Residual Heat Removal System o 2-075-0018B; Core Spray System o 2-075-0020; Core Spray System

• Verified Reactor Coolant System (RCS) inventory controls, especially during evolutions involving operations with the potential to drain the reactor vessel (OPDRV) controlled per 2-POI-200.5

• Verified electrical systems availability and alignment

• Monitored important control room plant parameters (e.g., RCS pressure, level, flow, and temperature) and TS compliance during the various shutdown modes of operation, and mode transitions

• Evaluated implementation of reactivity controls

• Reviewed control of containment penetrations and overall integrity

• Examined foreign material exclusion controls particularly in proximity to and around the reactor cavity, equipment pit, and spent fuel pool

• Routine tours of the control room, reactor building, refueling floor and drywell Reactor Vessel Disassembly and Refueling Activities The inspectors witnessed selected activities associated with reactor vessel reassembly, and reactor cavity flood-up in accordance with 2-GOI-100-3A, Refueling Operations (Reactor Vessel Disassembly and Floodup). Also, the inspectors witnessed fuel handling operations during the two Unit 2 reactor core fuel shuffles performed in accordance with TS and applicable operating procedures, such as 0-GOI-100-3A, Refueling Operations (In Vessel), 0-GOI-100-3B, Operations in the Spent Fuel Pool, and 0-GOI-100-3C, Fuel Movement Operations During Refueling. The inspectors verified specific fuel movements as delineated by the Fuel Assembly Transfer Sheets (FATF).

Furthermore, the inspectors also witnessed and examined the video verification of the final completed reactor core conducted per Attachment 6, of 0-GOI-100-3C.

Drywell Closeout On May 1 and 2, the inspectors reviewed the licensee's conduct of 2-GOI-200-2, Drywell Closeout, and performed an independent detailed closeout inspection of the Unit 2

drywell.

Restart Activities The inspectors specifically conducted the following:

• Witnessed heatup and pressurization of Unit 2 reactor pressure vessel in accordance with 2-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor pressure Vessel and Associated Piping

• Reviewed and verified completion of selected items of 0-TI-270, Refueling Test Program, Attachment 2, Startup Review Checklist

• Witnessed Unit 2 approach to criticality and power ascension per 2-GOI-100-1A, Unit Startup, and 2-GOI-100-12, Power Maneuvering

• Reactor Coolant Heatup/Pressurization to Rated Temperature and Pressure per 2-SR-3.4.9.1, Reactor Heatup and Cooldown Rate Monitoring Corrective Action Program The inspectors reviewed PERs generated during U2R17 and attended management

review committee (MRC) meetings to verify that initiation thresholds, priorities, mode holds, operability concerns and significance levels were adequately addressed. Resolution and implementation of corrective actions of several PERs were also reviewed for completeness.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors witnessed portions of, and/or reviewed completed test data for the following surveillance tests of risk-significant and/or safety-related systems to verify that the tests met Technical Specification surveillance requirements, UFSAR commitments, and in-service testing and licensee procedure requirements. The inspectors' review confirmed whether the testing effectively demonstrated that the SSCs were operationally capable of performing their intended safety functions and fulfilled the intent of the associated surveillance requirement. This activity constituted eight inspection samples, three in-service, four routine tests, and one reactor coolant system leakage detection test.

In-Service Tests

• April 23, 2013, 2-SR-3.6.1.3.8(6), Instrument Line Excess Flow Check Valve Operability Test, (Bench Testing)

• May 7, 2013, 2-SR-3.5.3.3 COMP, RCIC Comprehensive Full Flow Test; WO 113858689

• May 14, 2013, 3-SR-3.5.1.7, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Rx Pressure Routine Surveillance Tests

• April 26, 2013, 2-SR-3.6.1.3.8(4), Instrument Line Excess Flow Check Valve Operability Test

• May 3, 2013, 2-SR-3.5.1.1(HPCI), Maintenance of Filled High Pressure Coolant Injection (HPCI) Discharge Piping

• May 6, 2013, 1-SR-3.5.1.6(RHR I), Quarterly Residual Heat Removal (RHR) system rated flow test - Loop I

• June 1, 2013, 3-SR-3.8.1.1(3A), Diesel Generator "3A" Monthly Operability Test Reactor Coolant System Leak Detection Tests

• June 17, 2013, 2-SR-3.4.4.1, Manual Calculation of Unidentified, Identified and Total Leakage

b. Findings

No findings were identified.

Cornerstone:

Emergency Preparedness

1EP6 Drill Evaluation

a. Inspection Scope

During the report period, the inspectors observed an Emergency Preparedness (EP) drill that contributed to the licensee's Drill/Exercise Performance (DEP) and Emergency Response Organization (ERO) performance indicator (PI) measures on May 8, 2013, to identify any weaknesses and deficiencies in classification, notification, dose assessment and protective action recommendation (PAR) development activities. The inspectors observed emergency response operations in the simulated control room and certain Emergency Response Facilities to verify that event classification and notifications were done in accordance with EPIP-1, Emergency Classification Procedure and other applicable Emergency Plan Implementing Procedures. The inspectors also attended the post-drill critique to compare any inspector-observed weakness with those identified by the licensee in order to verify whether the licensee was properly identifying weaknesses.

This inspection activity satisfied one inspection sample for the Drill Evaluation of emergency preparedness.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Occupational Radiation Safety and Public Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Control

a. Inspection Scope

Radiological Hazard Assessment The inspectors reviewed a number of radiological surveys, including those performed for airborne areas, of locations throughout the facility including the Unit 2 drywell, Unit 1, Unit 2, and Unit 3 reactor buildings, the turbine building, and the Independent Spent Fuel Storage Installation (ISFSI). The inspectors also walked down many of the same areas and select radioactive material storage locations with a survey instrument, evaluating material condition, postings, and radiological controls. The inspectors observed jobs in radiologically risk-significant areas including high radiation areas and areas with, or with the potential for, airborne activity.

The inspectors evaluated the surveys in relation to the identified hazards for sufficient detail and frequency.

Instructions to Workers During plant walk downs, the inspectors observed labeling and radiological controls on containers of radioactive material. The inspectors also reviewed radiation work permits (RWP) used for accessing high radiation areas and airborne areas, verifying that appropriate work control instructions and electronic dosimeter (ED)setpoints had been provided and to assess the communication of radiological control requirements to workers. The inspectors reviewed selected ED dose and dose rate alarms, to verify workers properly responded to the alarms and that the licensee's review of the events was appropriate. The inspectors observed pre-job RWP briefings and health physics technician coverage of workers. The inspectors reviewed the various methods being used to notify workers of changing or changed radiological conditions.

Contamination and Radioactive Material Control The inspectors observed the release of potentially contaminated items from the r adiologically controlled area (RCA) and from contaminated areas such as the drywell. The inspectors also reviewed the procedural requirements for, and equipment used to perform, the radiation surveys for release of personnel and materiel. During plant walk downs, the inspectors evaluated radioactive material storage areas and containers, including satellite RCAs, assessing material condition, posting/labeling, and control of materials/areas. In addition, the inspectors reviewed the sealed source inventory and verified labeling, storage conditions, and leak testing of selected sources. The inspectors verified if Category 1 and 2 sealed sources had been appropriately reported to the National Source Tracking System and physically verified the presence and controls of these sources. The sources were verified to be physically present and in proper working order.

Radiological Hazards Control and Work Coverage The inspectors evaluated licensee performance in controlling worker access to radiologically significant areas and monitoring jobs in-progress associated with the Unit 2 refueling outage. Established radiological controls were evaluated for selected tasks including control rod drive removal and reinstallation activities, radioactive waste processing, fuel handling, and closeout inspections for Unit 1 restart. The inspectors evaluated the effectiveness of radiation exposure controls, including air sampling, barrier integrity, engineering controls, and postings through a review of both internal and external exposure results.

During walk downs with a radiation survey meter, the inspectors independently verified if ambient radiological conditions were consistent with licensee performed surveys, RWPs, and pre-job briefings; observed the adequacy of radiological controls; and observed controls for radioactive materials stored in the spent fuel pool. ED alarm set points and worker stay times were evaluated against area radiation survey results for drywell and refueling floor activities. The inspector did an independent radiological survey of the Browns Ferry ISFSI installation.

Risk-Significant High Radiation Area and Very High Radiation Area Controls The inspectors discussed the controls and procedures for locked-high radiation areas (LHRAs) and very high radiation areas (VHRAs) with health physics supervisors and the radiation protection manager. During plant walk downs, the inspectors verified the posting/locking of LHRA/VHRA areas.

Radiation Worker Performance and Radiation Protection Technician Proficiency-The inspectors observed radiation worker performance through direct observation, via remote camera monitoring, and via telemetry. These jobs were performed in high radiation, airborne, and/or contaminated areas. The inspectors also observed health physics technicians providing field coverage of jobs and providing remote coverage.

Problem Identification & Resolution - Licensee Corrective Action Program (CAP) documents associated with radiation monitoring and exposure control were reviewed and assessed. This included review of selected Problem Evaluation Reports (PERs)related to radworker and health physics technician performance. The inspectors evaluated the licensee's ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-3.1, Corrective Action Program, Revision 5. The inspectors also evaluated the scope of the licensee's internal audit program and reviewed recent assessment results. Documents reviewed are listed in the Attachment.

Radiation protection activities were evaluated against the requirements of Updated Final Safety Analysis Report (UFSAR) Section 12; TS Sections 5.4 and 5.7; 10 Code of Federal Regulations (CFR) Parts 19 and 20; and approved licensee procedures.

Radiological control activities for ISFSI areas were evaluated against 10 CFR Part 20, 10 CFR Part 72, and TS details. Documents reviewed are listed in the Attachment.

The inspectors completed 1 sample, as described in Inspection Procedure (IP) 71124.01.

b. Findings

No findings were identified.

2RS2 Occupational ALARA Planning and Controls

a. Inspection Scope

Radiological Work Planning Inspectors evaluated As Low As Reasonably Achievable (ALARA) program guidance and implementation for ongoing tasks associated with U2R17. Inspectors also evaluated tasks and review of post-outage ALARA activities associated with U3R15 refueling outage. A list was obtained from the licensee of work activities for their current outage. Inspectors selected four work activities to evaluate the ALARA Plan and associated documentation for jobs, including under-vessel maintenance; refuel floor maintenance activities, disassembly and refurbishing of Reactor Water Cleanup (RWCU) system isolation valve 2-FCV-069-0001, and other valve work activities. Inspectors evaluated dose mitigation features, dose goals and other factors that went into planning the dose goal for each task, including the review of TEDE ALARA evaluations for the decrease of worker efficiency from the use of respiratory protective devices. Selected RWPs were reviewed by inspectors to verify the integration of ALARA requirements into the documents for worker instruction. Inspectors followed the progression of available work activities to compare dose rates accrued and work evolution to the ALARA planning. Since inspectors were not at the site through the end of the outage, post job reviews from the previous U3R15 outage were reviewed.

Verification of Dose Estimates and Exposure Tracking Systems Three ALARA work packages and the assumptions and basis for the current collective exposure estimates

were reviewed by inspectors. The inspectors reviewed ALARA procedures, had discussions with ALARA personnel, reviewed daily exposure graphs and outage reports that tracked and trended the dose of ongoing work, and reviewed monthly Station ALARA Committee Meeting Minutes. The use of Work-In-Progress reviews for ALARA trigger points were also evaluated by the inspectors.

Source Term Reduction and Radiation Worker Performance The inspectors evaluated source term reduction methods through the review of licensee documents and records, and discussions with ALARA personnel. Inspectors reviewed actions already executed by the licensee to reduce source term, including replacing various plant components with Stellite free components, performing ultrasonic cleaning on fuel bundles, electro-polishing of various valves, pumps, and piping, and utilizing the Radiation Protection Closed Circuit Television (CCTV) Remote Monitoring System. The inspectors also reviewed future plans for source reduction, including the reduction of hotspots and soluble cobalt concentrations, and the use of permanent shielding.

The inspectors observed radiation worker performance through CCTV remote monitoring

and direct observations. This included Control Rod Drive (CRD) removals, a PSC Head Tank hotspot flush, and the attendance of ALARA, High Radiation Area (HRA), and pre-job briefs.

Problem Identification and Resolution The inspectors reviewed licensee corrective action documents associated with ALARA planning and controls. This included review of selected Problem Evaluation Reports (PERs) and self-assessments. The inspectors evaluated the licensee's ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-3.1 Revision 5. Radiation worker performance was evaluated against the requirements found in TS Sections 5.4 and 5.7; Title 10 Code of Federal Regulations (CFR) Parts 19 and 20; and approved licensee procedures. Documents reviewed are listed in the Attachment.

The inspectors completed 1 sample, as described in Inspection Procedure (IP)71124.02.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation

a. Inspection Scope

Engineering Controls The inspectors reviewed the use of temporary and permanent engineering controls to mitigate airborne radioactivity during refueling outage U2R17. In addition, during observations of jobs in-progress and containment walk-downs, inspectors observed the placement and use of HEPA negative pressure units, and air sampling equipment.

Use of Respiratory Protection Devices & Self-Contained Breathing Apparatus for Emergency Use Inspectors reviewed the use of respiratory protection devices to limit the intake of radioactive material, including devices used for routine tasks and devices stored for use in emergency situations. Inspectors observed the physical condition of Self-Contained Breathing Apparatus (SCBA) units, negative pressure respirators (NPRs), powered air purifying respirators (PAPRs) and device components staged for routine and emergency use throughout the plant. SCBA bottle air pressure, the number of units, and the number of spare masks and air bottles available was also evaluated by inspectors. The inspectors reviewed maintenance records for selected SCBA units for the past year and evaluated SCBA and NPR compliance with National Institute for Occupational Safety and Health certification requirements. The inspectors also reviewed records of Grade D (or better) air quality testing for supplied-air devices and SCBA bottles. The inspectors reviewed the status and surveillance records of SCBAs staged for in-plant use during emergencies through review of records and walk-down of SCBA staged in the control room and selected locations.

The inspectors verified the licensee had procedures in place to ensure that the use of respiratory protection devices was ALARA when engineering controls were not practicable. Control room operators and fire brigade were interviewed on the use of the devices including SCBA bottle change-out and use of corrective lens inserts. In addition, qualifications for individuals responsible for testing and repairing SCBA vital components were evaluated through review of training records. Selected maintenance records for SCBA units and air cylinder hydrostatic testing documentation were reviewed.

The inspectors verified that the licensee has procedural requirements in place for evaluating air samples for the presence of alpha emitters and reviewed airborne radioactivity and contamination survey records for selected plant areas to ensure air samples are screened and evaluated per the procedure requirements.

The inspectors walked-down the respirator issue and storage locations and verified that the equipment was appropriately stored and maintained. Records of monthly and quarterly inventory and inspection of the equipment were also reviewed by the inspectors. The inspectors discussed the process for issuing respirators, and verified that selected individuals qualified for respirator and/or self-contained breathing apparatus (SCBA) use had completed the required training, fit-test, and medical

evaluation.

Problem Identification and Resolution Licensee CAP documents associated with the control and mitigation of in-plant radioactivity were reviewed and assessed. This included review of selected SRs related to use of respiratory protection devices including SCBA. The inspectors evaluated the licensee's ability to identify, characterize, prioritize, and resolve the identified issues in accordance with procedure NPG-SPP-03.1, Corrective Action Program, Revision 5. The inspectors also evaluated the scope of the licensee's internal audit program and reviewed recent assessment results. Documents reviewed are listed in the Attachment.

Radiation protection activities were evaluated against the requirements specified in 10 CFR Parts 19 and 20; and approved licensee procedures. Documents reviewed are listed in the Attachment.

The inspectors completed all specified line-items detailed in IP 71124.03 (sample size of 1).

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment

a. Inspection Scope

External Dosimetry: The inspectors reviewed National Voluntary Laboratory Accreditation Program (NVLAP) certification data and discussed program guidance for storage, processing, and evaluation of results for active and passive personnel dosimeters currently in use. Comparisons between Electronic Dosimeter (ED) and Thermo Luminescent Dosimeter (TLD) data were discussed in detail. The inspectors reviewed ED alarm logs and reviewed licensee's dosimeter incident reports and assessment actions for selected alarm events.

Internal Dosimetry: Program guidance and assessment results for internally deposited radionuclides were reviewed. The inspectors reviewed selected Whole Body Count (in vivo) analyses from January 2012 to March 2013. Capabilities for collection and analysis of special bioassay samples were discussed with licensee staff, there were no dose assessments based on biological samples to review.

Special Dosimetric Situations: The inspectors evaluated the licensee's use of multi-badging, extremity dosimetry, and dosimeter relocation within non-uniform dose rate fields and reviewed assessments. Worker monitoring in neutron areas was discussed with licensee staff. The inspectors also reviewed records of monitoring for declared pregnant workers from September 2011 to March 2013 and discussed monitoring guidance with dosimetry staff. In addition, shallow dose assessments for selected Personnel Contamination Events occurring between September 2011 and March 2013 were reviewed and discussed.

Problem Identification and Resolution: The inspectors reviewed and discussed selected Corrective Action Program (CAP) documents associated with occupational dose assessment. The inspectors evaluated the licensee's ability to identify and resolve the issues in accordance with procedure NPG-SPP-03.1, "Corrective Action Program", Rev. 5. The inspectors also discussed the scope of the licensee's internal audit program and reviewed recent assessment results.

Occupational dose assessment activities were evaluated against the requirements of 10 CFR Parts 19 and 20; and approved licensee procedures. Documents reviewed are listed in the Attachment.

The inspectors completed 1 sample as required by IP 71124.04.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation

a. Inspection Scope

The inspectors reviewed the licensee's radiation monitoring instrumentation programs to

verify the accuracy and operability of radiation monitoring instruments used to monitor areas, materials, and workers to ensure a radiologically safe work environment and to detect and quantify radioactive process streams and effluent releases.

Walkdowns and Observations: The inspectors walked down effluent and process monitoring systems, including the Main Stack Radiation Monitoring System (0-RE-90-147 and 148), Unit 1 and (U1 and U2) Reactor Building Vent Exhaust (1-RM-249 and 250), and Liquid Radwaste Monitor (0-RE-90-130), evaluating material condition and verifying configurations were consistent with Offsite Dose Calculation Manual (ODCM) descriptions. The inspectors also evaluated the material condition and location of area radiation monitors (ARMs) 1-RE-90-1, 2-RE-90-4, 2-RE-90-17, and continuous air monitors. For selected effluent monitors and ARMs, the inspectors verified in-field responses were consistent with readings obtained in the control room.

During plant tours and observations in the calibration lab, the inspectors assessed material condition and operability of portable survey instruments in addition to verifying calibration and source checks were current. The inspectors reviewed records of survey instrument function/source checks and observed and discussed performance of required checks with calibration lab personnel. Material condition of source check devices, device operation, and establishment of source check acceptance range were also discussed with calibration lab personnel.

The inspectors evaluated material condition and observed performance of source checks on personal contamination monitors and small article monitors located at the RCA exit and discussed differences in source check geometries for portal monitors located at the protected area exit.

Calibration and Testing Program: The inspectors reviewed the last two calibration records for the following effluent, process, area radiation, and post-accident monitors:

2-RM-90-250 (Unit 2 Reactor Building Vent Exhaust), 1-RM-90-272A and 273A (Unit 1 Containment High Range Radiation Monitors), 0-RE-90-130 (Liquid Radwaste Monitor), and 0-RE-90-147/148 (Main Stack). In addition to evaluating the calibration procedures, calibration geometry, functional tests, and calibration sources, the inspectors verified monitor set-points were consistent with and/or changed in accordance with ODCM and/or site procedures.

Instrumentation used in the chemistry and health physics counting rooms was evaluated for material condition, operability, and use. Daily background and quality control charts for select high-purity germanium spectroscopy, low background counting systems, and alpha counting systems were reviewed. The inspectors also reviewed the cross-check analysis results for several quarters of calendar year 2011 and 2012.

For the whole body counter, the inspectors reviewed the most recent calibration, assessed the isotope library, reviewed and discussed performance of daily quality control (QC) checks, and verified appropriate check and calibration sources were used. In addition, the inspectors reviewed calibrations of, and observed performance of source checks on select portal monitor, personnel monitor, and small article monitor equipment.

Documents reviewed are listed in the Attachment.

The inspectors reviewed performance of the portable instrument calibration lab through review and discussion of instrument calibrations, direct observation of source and response checks, review of instrument calibration records, assessment of the established source check ranges of the Shepherd calibrator (geometry, sources, etc.),

and review of the annual recertification of the Western Area Radiological Laboratory (WARL) high level gamma well calibrator. Portable instrument calibration records review included three ion chamber instruments, two neutron instruments, four low volume air samplers, and three friskers.

Operability and reliability of selected radiation detection instruments were reviewed against details documented in the following: 10 CFR Part 20; NUREG-0737, Clarification of TMI Action Plan Requirements; UFSAR Chapters 7 and 13; TS Sections 3.3.3.1, Post Accident Monitoring, 3.3.6.2, Secondary Containment Isolation Instrumentation, 5.4, Procedures, and 5.5 Programs and Manuals; and applicable licensee procedures.

Document reviewed are listed in the Attachment.

Problem Identification and Resolution: Selected corrective action program documents associated with radiation monitoring instruments, including condition reports and audits, were reviewed and assessed. This review of corrective action documents included evaluating the licensee's response to indications of degraded count room instrument performance. The inspectors verified that problems were being identified at an appropriate threshold and resolved in accordance with procedures NPG-SPP-03.1, Corrective Action Program.

The inspectors completed the specified line-item samples detailed in Inspection Procedure (IP) 71124.05. Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA1 Performance Indicator (PI) Verification

.1 Cornerstone: Mitigating Systems

Mitigating Systems Performance Indicator (MSPI)- Heat Removal (Reactor Core Isolation Cooling and High Pressure Coolant Injection)

a. Inspection Scope

The inspectors reviewed the licensee's procedures and methods for compiling and reporting the following Performance Indicators (PIs), including procedure NPG-SPP-02.2 Performance Indicator Program. The inspectors examined the licensee's PI data for the specific PIs listed below for the second quarter 2012 through first quarter of 2013. The inspectors reviewed the licensee's data and graphical representations as reported to the NRC to verify that the data was correctly r eported. The inspectors also validated this data against relevant licensee records (e.g., PERs, Daily Operator Logs, Plan of the Day, Licensee Event Reports, etc.), and assessed any reported problems regarding implementation of the PI program. Furthermore, the inspectors met with responsible plant personnel to discuss and go over licensee records to verify that the PI data was appropriately captured, calculated correctly, and discrepancies resolved. The inspectors also used the Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, to ensure that industry reporting guidelines were appropriately applied. This activity cons tituted six mitigating systems performance indicator inspection samples.

• Unit 1 Mitigating Systems Performance Index - Reactor Core Isolation Cooling

• Unit 2 Mitigating Systems Performance Index - Reactor Core Isolation Cooling

• Unit 3 Mitigating Systems Performance Index - Reactor Core Isolation Cooling

• Unit 1 Mitigating Systems Performance Index - High Pressure Coolant Injection

• Unit 2 Mitigating Systems Performance Index - High Pressure Coolant Injection

• Unit 3 Mitigating Systems Performance Index - High Pressure Coolant Injection

b. Findings

One finding is documented as a licensee identified violation in Section 4OA7.

.2 Radiation Protection

a. Inspection Scope

Occupational Radiation Safety Cornerstone: The inspectors reviewed Performance Indicator (PI) data collected from April 14, 2012, through May 1, 2013, for the Occupational Exposure Control Effectiveness PI. For the reviewed period, the inspectors assessed PER records to determine whether HRA, VHRA or unplanned exposures, resulting in TS or 10 CFR 20 non-conformances, had occurred during the review period. The inspectors reviewed radiologically controlled area exit transactions with exposures greater than 100 mrem to determine if they were consistent with the requirements of the RWP. The reviewed data were assessed against guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Rev. 6. Documents reviewed are listed in the Attachment.

Public Radiation Safety Cornerstone: The inspectors reviewed the Radiological Control Effluent Release Occurrences PI results for the Public Radiation Safety Cornerstone from June 22, 2012, through March 1, 2013. For the assessment period, the inspectors reviewed cumulative and projected doses to the public and PER documents related to Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual issues.

Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Review of items entered into the Corrective Action Program:

As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"

and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensee's CAP. This review was accomplished by reviewing daily PER and Service Request (SR) reports, and periodically attending Corrective Action Review Board (CARB) and PER Screening Committee (PSC) meetings.

.2 Annual Follow-up of Selected Issues - Operator Aids and Permanent Information

Postings

a. Inspection Scope

The inspectors reviewed the deficiencies from an audit of operator aids (OAs) and permanent information postings (PIPs) conducted in May 2011. The audit was one of the corrective actions associated with PER 344224, which addressed that an audit of OAs and PIPs had not been performed since October 2004. Procedure 0-TI-414, Component Labeling, Signs, Operator Aids, and Permanent Information Postings, requires an audit to be performed every 24 months. In addition to directing the

performance of the audit, the PER captured the need to documentation of any deficiencies and schedule a subsequent audit by May 2013. The audit was completed in May 2011. The inspectors reviewed the results of the audit and did a sampling of the identified deficiencies to verify the deficiencies had been corrected. The inspectors also verified that the audit scheduled for May 2013 had been conducted and reviewed the results. The inspectors verified the deficiencies identified from the May 2013 audit have been entered into the corrective action program. This activity constituted one in-depth selected issue. Documents reviewed are listed in the Attachment.

b. Findings and Observations

No findings were identified. However, the inspectors had the following observations:

PER 344224 The PER addressed the failure to perform an audit of operator aids (OAs) and permanent information postings (PIPs). The corrective actions included completing the audit, identifying any deficiencies with the OAs /PIPs, and scheduling a subsequent audit within two years of the previous audit. The inspectors performed a walkdown of twenty-four OAs / PIPs from the list of deficiencies identified during the 2011 audit to verify the deficiencies had been addressed. Of the OAs/PIPs sampled, one PIP had not been installed and another PIP was unmounted laying on a piece of equipment approximately 10 feet from its associated component. A review of the recent audit, completed in May 2013, determined that while this audit identified several deficiencies it failed to identify either the missing PIP or the improperly mounted PIP identified during the inspectors' walkdown. The information was provided to the licensee who entered the issue into the corrective action program and took immediate actions to correct the deficiencies identified during the inspectors' walkdown.

.3 Semiannual Review to Identify Trends

a. Inspection Scope

As required by Inspection Procedure 71152, the inspectors performed a review of the licensee's corrective action program (CAP) implementation and associated documents to identify trends that could indicate the existence of a more significant safety issue. The inspectors' review included the results from daily screening of individual PERs (see Section 4OA2.1 above), licensee trend reports and trending efforts, and independent searches of the PER database and WO history. The inspectors' review nominally considered the six-month period of January 2013 through June 2013, although some searches expanded beyond these dates. Additionally, the inspectors' review also included the Integrated Trend Reports (ITR) from October 1, 2012, to March 31, 2013.

Furthermore, the inspectors verified that adverse or negative trends identified in the licensee's PERs, periodic reports and trending efforts were entered into the CAP. Inspectors interviewed the appropriate licensee staff and also reviewed procedures, NPG-SPP-02.7, PER Trending and NPG-SPP-02.8, Integrated Trend Review.

The purpose of the licensee's integrated trend reviews was to identify the top site and departmental issues (gaps to excellence)requiring management attention. Other objectives were to provide status of the top issues and their progress to resolution, identify continuing issues, emerging trends and issues to be monitored, review progress towards resolving past top issues, review issues identified by external organizations such as the NRC, Nuclear Safety Review Board (NSRB), Quality Assurance, etc., and determine why they were not identified by line organizations.

In addition to reviewing the site's progress on the above issues, the inspectors conducted an independent review of the licensee's CAP to independently identify potential adverse trends.

b. Findings and Observations

No findings were identified. However, the inspectors had the following observations discussed below:

Inspectors noted licensee-identified issues and trends in both the first and second quarter trend reports. The licensee continued to trend weaknesses related to human performance. This issue was identified as site wide and was a major area of concern to the site. The licensee developed a performance improvement plan and actions were initiated to eliminate the performance gap in this area.

Inspectors continue to see increased focus and improvement on the licensee's trending efforts and trending related products. The station's departmental accountability in terms of responding to trending issues is slowly improving.

Inspectors identified three issues that were either not clearly identified or were not being trended by the licensee's PI group.

Inspectors identified a potential adverse trend with the adequacy of briefings conducted to support complex infrequently performed tests or evolutions (CIPTE). The licensee initiated PER 746111 to address this trend. The following PER's were provided as data points support this trend:

• PER 707725, Briefing for an operation with potential to drain the reactor vessel (OPDRV) for repair of 2-FCV-69-1.

• PER 718099, CIPTE briefing for reactor pressure vessel pressure test.

• PER 724721, CIPTE test director and manager not assigned in a timely manner.

• PER 702110, Quality Assurance (QA) identified gaps in CIPTE brief for control rod (CR) exchange brief.

• PER 702679, QA identified CIPTE issues during CR brief.

4OA3 Follow-up of Events

.1 (Closed) Licensee Event Report (LER) 05000296/2013-002-00; 05000296/2013-002-01 , Manual Actuation of Reactor Core Isolation Cooling (RCIC) System During Reactor

Shutdown

a. Inspection Scope

The inspectors reviewed LER 05000296/2013-002-00 dated April 12, 2013 and revised

LER 05000296/2013-002-01 dated June 7, 2013.

Inspectors reviewed the information from the Root Cause Analysis for PER 710216 related to this event.

The RCIC system was required due to a failure of the Reactor Feedwater recirculation piping separation which caused a loss of condenser vacuum.

The Condensate System is currently in a(1) status for Maintenance Rule tracking. The corrective actions associated with the a(1) plan are being implemented to minimize future recurrences of this issue. The use of RCIC and the Safety Relief Valves to control reactor level and pressure during the planned shutdown and cooldown of the reactor was performed in accordance with Browns Ferry procedures. These LER's were closed.

b. Findings

No findings were identified.

.2 (Closed) Licensee Event Reports (LER) 05000260/2012-003-00 and 05000260/2012-003-01 Reactor Motor Operated Board Transfer Failure

a. Inspection Scope

The inspectors reviewed LERs 05000260/2012-003-00 dated September 17, 2012, and revised LER 05000260/2012-003-01 dated March 15, 2013.

Inspectors reviewed the information from the Root Cause Analysis for PER 581478 related to this event.

The cause of the Reactor Motor Operated Valve Board transfer failure was due to a failure to perform a post maintenance test which subsequently caused a manual trip pushbutton to remain stuck in the trip position. Corrective actions were to revise the maintenance test procedures with verification steps to ensure the trip pushbuttons had returned from the depressed position. One violation of regulatory requirements was identified and is discussed below. These LERs were closed.

b. Findings

Introduction:

A self-revealing Green non-cited violation (NCV) Technical Specifications (TS) Limiting Conditions for Operation (LCO) 3.5.1, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System was identified associated with the licensee's failure to perform a post maintenance test (PMT) which rendered an ECCS injection subsystem power supply inoperable for greater than TS allowed outage time.

Description:

On July 19, 2012, while Operations personnel were performing a manual transfer of the Browns Ferry Unit 2, 480V RMOV board 2E to its alternate power supply, the alternate feeder breaker failed to close. Operations personnel found the manual trip pushbutton in the depressed position. Section 3.2.2 A.1 of procedure NPG-SPP-06.3, Pre-/Post Maintenance Testing required, in part, post maintenance testing (PMT) to be sufficiently comprehensive to ensure that the maintenance performed does not adversely affect the equipment's ability to perform its intended function, and that no new or related problems were created by the maintenance activity. The licensee failed to perform a PMT to test the Unit 2 2E 480V RMOV Board alternate feeder breaker on July 10, 2012. As a result, the breaker was left in a tripped condition and unable to close when the 2E RMOV board was returned to operable status. It was subsequently discovered, nine days later, that the trip push button was stuck during a manual transfer attempt from the normal power source to the alternate power source on July 19, 2012.

Analysis:

Failure to perform a post maintenance test before returning the 2E 480V RMOV board to operable status as required by licensee procedure NPG-SPP-06.3, Pre-/Post Maintenance Testing was a performance deficiency. This finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone objective and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the Unit 2 2E 480V RMOV Board alternate feeder breaker was left in a tripped condition and unable to close for nine days. The significance of this finding was evaluated using Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings and IMC 0609, Appendix A of the Significance Determination Process (SDP)

The issue was screened using IMC 0609, Appendix A, Exhibit 2 - Mitigating Systems Screening Questions, and was determined to be a Green finding because it did not represent an actual loss of function of one or more non-Technical Specification Trains of equipment designated as high safety-significant in accordance with the licensee's maintenance rule program for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The cause of this finding was directly related to the cross-cutting aspect of appropriate oversight of work activities in the Work Practices component of the Human Performance area because maintenance work activities for the alternate feeder breaker for the 2E 480V RMOV board did not include a check to ensure the breaker nuclear safety function was supported, H.4(c).

Enforcement:

Technical Specifications(TS) 3.5.1, Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System required, in part, that while the plant is in Mode 1, 2 and 3, each ECCS injection/spray subsystem shall be OPERABLE. The TS ACTION statement requires that with "One low pressure ECCS injection/spray subsystem inoperable", restore low pressure ECCS injection/spray

subsystem(s) to OPERABLE status within seven days or be Mode 3 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Contrary to the above, while the Unit was in Mode 1, from July 10 until July 19, 2012, division 2 of low pressure ECCS was inoperable due to the Unit 2, 480V RMOV board 2E being unable to perform its intended transfer function as described in Technical Specification Surveillance Requirement 3.5.1.12 and action was not taken to either restore the system to operable status within seven days or place the unit in Mode 3 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Mode 4 within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This issue was captured in the licensee's corrective action program as PER 581478. The failed alternate feeder breaker to the Unit 2 2E 480V RMOV Board trip pushbutton was restored to the proper condition on July 19, 2012, at 0303. This violation was applicable to Unit 2 and is identified as NCV 05000260/2013003-01, Failure to Perform Post Maintenance Testing on the 2E 480V RMOV board.

.3 (Discussed) Licensee Event Report (LER) 05000296/2013-003-00 , Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System from a Turbine Trip.

Licensee Event Report (LER) 05000259/2013-002-00, Manual Reactor Shutdown Due to Decreasing Condenser Vacuum

a. Inspection Scope

On February 25, 2013 at 1313 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.995965e-4 months <br />, Unit 3 automatically scrammed due to a turbine trip. The turbine trip was caused by low condenser vacuum as a result of a failure of a long cycle return line piping connection to the miscellaneous drain header. The licensee initiated PER 687732 to enter the event into the corrective action program. Inspectors reviewed LER 05000296/2013-003-00 and all associated documentation which included the Root Cause Analysis (RCA) for PER 687732. This LER was Open.

On March 19, 2013 at approximately 0402 hours0.00465 days <br />0.112 hours <br />6.646825e-4 weeks <br />1.52961e-4 months <br />, Unit 1 was manually scrammed due to decreasing main condenser vacuum. The degrading condenser vacuum was caused by the failure of a vent and drain pipe joint to a miscellaneous drain header which connected to the main condenser. The licensee initiated PER 698870 to enter the event into the corrective action program. Inspectors reviewed event report LER 05000259/2013-002-00 and all associated documentation which included the Root Cause Analysis (RCA) for PER 698870. The direct cause of the event was attributed to cyclic fatigue of the drain piping to header connection. The root cause was attributed to upstream valve leakage and the station's failure to consider and address the possible failure of the piping due to the valve leakby. This LER was open.

b. Findings

Introduction:

Unresolved Item (URI). A self-revealing issue of concern was identified with the Unit 3 automatic reactor shutdown due to the turbine trip from loss of condenser vacuum on February 25, 2013 and the Unit 1 manual reactor shutdown due to a decreasing vacuum manual scram on March 19, 2013.

Description:

On February 25, 2013, Unit 3 was operating at approximately 92 percent power following a forced midcycle outage for repairs on the condenser circulating water (CCW) system. At 1313 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />4.995965e-4 months <br /> the unit automatically scrammed due to a turbine trip. The turbine trip was caused by low condenser vacuum as a result of a failure of a long cycle return line piping connection to the miscellaneous drain header which connects with the main condenser. Main steam isolation valves (MSIVs) were manually closed and main turbine bypass valves were unavailable due to the loss of condenser vacuum.

During the transient all available mitigating equipment performed as designed. The licensee initiated PER 687732 to enter the event into the corrective action program.

Licensee investigation revealed the failure of a feedwater long cycle return line connection to the miscellaneous drain header was a result of seat leakage from the reactor feedwater system long cycle return flow control valves (3-FCV-003-0071, 72, and 73). This seat leakage caused the water within the pipe to flash to steam which translated to excessive vibration on the pipe and resulted in fatigue failure of the connection between the 8 inch pipe and 24 inch miscellaneous drain header.

On March 19, 2013, Unit 1 was operating at approximately 80 percent power. At 0402 hours0.00465 days <br />0.112 hours <br />6.646825e-4 weeks <br />1.52961e-4 months <br /> the unit was manually scrammed due to decreasing main condenser vacuum.

Following the manual scram condenser vacuum recovered and all available mitigating

equipment was available and performed as designed.

The vacuum degradation was caused by the separation of a 4 inch vent and drain pipe from a 24 inch miscellaneous drain header which connected to the main condenser.

The separation was caused by vibration induced cyclic fatigue as a result of the combination of leaking drain valves and repeated operation of dump valves associated with feedwater heaters. The licensee initiated PER 698870 to enter the event into the corrective action program.

Additional inspection of the maintenance history, service requests (SRs), problem evaluation reports (PERs), cause analyses and mitigating system responses concerning these two reactor trips are required to provide additional information into the issue. This unresolved item was tracked as URI 05000259, 260, 296/2013003-02.

.4 (Closed) Licensee Event Report (LER) 05000259/2010-001-01, Unit 1, 2, and 3

Appendix R Safe Shutdown Instruction Procedures Contain Incorrect Operator Manual Actions (Closed) Licensee Event Report (LER) 05000259/2012-001-01, Unanalyzed Conditions Discovered During NFPA 805 Transition Review (Closed) Licensee Event Report (LER) 05000259/2012-002-01, Fault Propagation During a Postulated Appendix R Event Could Result in an Inability to Close Motor Operator Valves

(Closed) Licensee Event Report (LER) 05000259/2012-003-01: Reactor Protection System Circuit Could Potentially Remain Energized During An Appendix R Fire

a. Inspection Scope

On December 26, 2012, the licensee submitted revisions to the subject LERs to reflect that conditions that were originally submitted are not considered as safety system functional failures, as described in NEI 99-02, "Regulatory Assessment Performance Indicator Guidelines." The regulatory aspects of LER 2010-001-01 were reviewed and

documented in NRC Inspection report 05000259, 260, 296/2012-007. Similarly, the regulatory aspects of LERs 2012-001-01, 2012-002-01, and 2012-003-01 were reviewed and documented in NRC Inspection Report 05000259, 260, 296/2012-004.

Inspectors reviewed the subject LERs to determine if the facts supporting the original LER reviews were still applicable, and that no new conditions that had not been reviewed by the NRC were associated with this revision. Inspectors also reviewed NEI 99-02 to verify the licensee's basis for determining that these conditions were not safety system functional failures was consistent with the guidance in NEI 99-02.

b. Findings

No findings were identified.

4OA5 Other Activities

.1 Temporary Instruction (TI) -2515/182 - Review of the Implementation of the Industry Initiative to Control Degradation of Underground Piping and Tanks, Phase 1

a. Inspection Scope

Leakage from buried and underground pipes has resulted in ground water contamination incidents with associated heightened NRC and public interest. The industry issued a guidance document, Nuclear Energy Institute (NEI) 09-14, "Guideline for the Management of Buried Piping Integrity," (ADAMS Accession No. ML1030901420), to describe the goals and required actions (commitments made by the licensee) resulting from this underground piping and tank initiative. On December 31, 2010, NEI issued Revision 1 to NEI 09-14, "Guidance for the Management of Underground Piping and Tank Integrity," (ADAMS Accession No. ML110700122), with an expanded scope of components which included underground piping that was not in direct contact with the soil and underground tanks. On November 17, 2011, the NRC issued TI-2515/182 "Review of the Industry Initiative to Control Degradation of Underground Piping and Tanks," to gather information related to the industry's implementation of this initiative.

The inspectors reviewed the licensee's programs for buried pipe and underground piping and tanks in accordance with TI-2515/182 to determine if the program attributes and completion dates identified in Sections 3.3 A and 3.3 B of NEI 09-14, Revision 1 were contained in the licensee's program and implementing procedures. For the buried pipe and underground piping program attributes, with completion dates that had passed, the inspectors reviewed records to determine if the attribute was in fact complete and to determine if the attribute was accomplished in a manner which reflected good or poor practices in program management.

b. Observations The licensee's buried piping and underground piping and tanks program was inspected in accordance with paragraphs 03.01.a through 03.01.c of TI-2515/182 and was found to meet all applicable aspects of NEI 09-14 Revision 1, as set forth in Table 1 of the TI. Based upon the scope of the review described above, Phase I of TI-2515/182 was completed.

c. Findings

No findings were identified.

.2 Independent Spent Fuel Storage Installation (ISFSI) Change Evaluations.

a. Inspection Scope

Under the guidance of IP 60857, the inspectors reviewed the licensee's evaluations of the changes to the Independent Spent Fuel Storage Installation (ISFSI) in accordance with 10 CFR 72.48, Changes, Tests, and Experiments, as well as the licensee's procedure for implementing 72.48 evaluations. The review focused on the changes that were implemented since the last inspection. The review determined that the evaluations were consistent with the requirements of 10 CFR 72.48 and the evaluations were documented in accordance with NPG-SPP-09.9, "10 CFR 72.48 Evaluations of Changes, Tests and Experiments for Independent Spent Fuel Storage Installation". The inspectors also reviewed the changes to10CFR72.212, Report of Evaluations since the last inspection.

b. Findings

No findings were identified.

.3 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours. These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status reviews and inspection activities.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

.1 Exit Meeting Summary

On April 12, 2013, the health physics inspectors discussed results of the onsite radiation protection inspections with Mr. Phil Summers, Director of Safety & Licensing, and other responsible staff. The inspectors noted that one proprietary document was reviewed during the course of the inspection that would be properly disposed of when no longer needed.

On May 31, 2013, an exit meeting was conducted by phone with Steve Norris, Component Engineering Manager, members of the buried pipe program and licensee staff. The inspectors verified that all proprietary information was returned to the licensee.

On July 9 and August 5, 2013, the resident inspectors presented the quarterly inspection results to Mr. Keith Polson, Site Vice President, and other members of the licensee's staff, who acknowledged the findings.

4OA7 Licensee-Identified Violations

The following Severity Level IV violation was identified by the licensee and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation.

The licensee identified a violation of Title 10 CFR Part 50.9(a), "Completeness and Accuracy of Information," which required, in part, that information provided to the NRC

by all licensees be complete and accurate in all material respects. Contrary to the above, on January 21, 2013, the licensee failed to provide to the NRC fourth quarter 2012 performance indicator data for the Unit 2 MSPI - emergency AC power systems that were complete and accurate in all material respects. Specifically, the licensee failed to include the December 2012 failure of D EDG as an MSPI "run" failure. Once the information was corrected, the Unit 2 MSPI performance indicator changed from Green to White. This violation was characterized as a Severity Level IV non-cited violation, consistent with Example 6.9.d.11 of the Enforcement Policy. The violation was entered into the licensee's corrective action program as PERs 704392, 669462, and 740285.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

S. Austin, Licensing Engineer

E. Bates, Licensing Engineer

A. Bolduc, I&C Supervisor

T. Cagle-Jaudon, RP Supervisor - Instruments

P. Campbell, System Engineer

J. Covey, Radiation Protection - ALARA

S. Cowan, Superintendant Radiation Protection Operations

D. Drummonds, Underground and Buried Piping Program Owner

M. Ellet, BFN Maintenance Rule Engineer

J. Emens, Site Licensing Manager

J. Ferguson, Radiation Protection Manager

M. Floyd, Nuclear Fatigue Rule Manager

R. Givens, Senior Reactor Operator

D. Green, Licensing Contractor

J. Guthrie, System Engineer

P. Hermann, Technical Consultant, Licensing

L. Hughes, Manager Operations

S. Jeffers, Radiation Protection - Dosimetry

J. Lacasse, System Engineer

B. McNutt, Shift Manager

T. Mingus, Engineering

S. Norris, Components Mgr.

J. Oates, System Engineer - I&C

M. Oliver, Licensing Engineer

K. Polson, Site Vice President

B. Rinne, Components Engineer

M. Roy, System Engineer

J. Solley, I&C Supervisor

P. Summers, Director of Safety and Licensing

S. Wentzel, System Engineer

A. Yarborough, System Engineer

LIST OF ITEMS

Opened

05000259, 260, 296/2013003-02 URI LER

05000296/2013-003-00 (Section 4OA3.3) and

05000259/2013-002-00 (Section 4OA3.3)

05000260/2013003-03 URI Residual Heat Removal (RHR) Heat Exchanger (HX) excessive fouling (1R15.2)

Opened and Closed

05000259, 260, 296/2013002-01 NCV Failure to Implement Preventive Maintenance Program (Section 1R15.1)

05000260/2013003-01 NCV Failure to perform Post Maintenance Testing on the

2E 480V RMOV board (Section 4OA3.2)

Closed

05000296/LER-2013-002-00 LER Manual Actuation of Reactor Core Isolation Cooling (RCIC) System During Reactor Shutdown (Section 4OA3.1)

05000296/LER-2013-002-01 LER Manual Actuation of Reactor Core Isolation Cooling (RCIC) System During Reactor Shutdown (Section 4OA3.1)

05000260/LER-2012-003-00 LER Reactor Motor Operated Valve Board transfer Failure (Section 4OA3.2)

05000260/LER-2012-003-01 LER Reactor Motor Operated Valve Board transfer Failure (Section 4OA3.2)

05000259, 260, 296/2010-001-01 LER Unit 1, 2, and 3 Appendix R Safe Shutdown Instruction Procedures Contain Incorrect Operator Manual Actions (Section 4OA3.4)

05000259, 260, 296/2012-001-01 LER Unanalyzed Conditions Discovered During NFPA

805 Transition Review (Section 4OA3.4)

05000259, 260, 296/2012-002-01 LER Fault Propagation During a Postulated Appendix R Event Could Result in an Inability to Close Motor

Operator Valves (Section 4OA3.4)

Attachment

05000259, 260, 296/2012-003-01 LER Reactor Protection System Circuit Could Potentially Remain Energized During An Appendix R Fire (Section 4OA3.4)

05000259, 260, 296/2013002-01 AV Failure to Implement Preventive Maintenance

Program, (Section 1R15.1)

Discussed

05000296/2013-003-00 LER Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System from a Turbine

Trip (Section 4OA3.3)

05000259/2013-002-00 LER Manual Reactor Shutdown Due to Decreasing Vacuum (Section 4OA3.3)

Section 1R01:

Adverse Weather Protect ion - Readiness for Impending Adverse Weather Conditions - Offsite Power System Readiness

0-GOI-200-3 Appendix A dated 4/17/2013 0-GOI-200-3 Attachment 1 dated 5/30/2013

0-GOI-200-3 Hot Weather Operations Rev 13 Meeting minutes from Hot Weather Preparedness meeting dated 5/28/2013

NRC

BL 2012-01 Design Vulnerability in Electric Power System

NRC

RIS 2011-12 Adequacy of Station Electric Distribution System Voltages Revision 1

Operating Experience

Smart Sample (OpESS) 2012/01, High Wind Generated Missile Hazards, dated 12/29/2011

PER 516502, Loss of Both Madison and Maury 500KV Lines During Severe Weather

PER 516503, U3 Power Ascension Suspended Tornado Warning and Loss of (2) 500KV Lines

PER 517126, Inadequate Actions and Procedures for Severe Weather at BFN

PER 623173, Procedure Does Not Exist for Tie-Down of Material Stored in Outside Areas

PER 728812 Summer Readiness Milestone #20 not met

TVA Response to NRC

BL 2012-01 dated Oct 25, 2012

WO 04-720020-000 Replace pump

BFN-2-PMP-024-0092

WO 06-710860-000 Troubleshoot Turbine Bldg Exhaust Fan 3-FAN-064-0048J

WO 06-710876-001 Troubleshoot ground on 4kV cooling tower switchgear B and D

WO 08-713069-000 Troubleshoot/repair Turbine Bldg Supply fan 3A

WO 09-714761-000 Troubleshoot/repair Turbine Bldg Supply fan 1A

Section 1R04: Equipment Alignment

1-OI-75, Core Spray System, Rev. 28 1-OI-75/ATT-2, Attachment 2 Core Spray System Panel Lineup Checklist, Rev. 21

1-OI-75/ATT-3, Attachment 3 Core Spray System Electrical Lineup Checklist, Rev. 21

EN 49071, Diesel Generator A Fuel Pressure Sense Line Broken

LCO Tracking 1-075-TS-2013-0027

PER 731716, Diesel Generator A Fuel Pressure Sense Line Broken Attachment

WO 06-718768-001 Standby Gas Treatment (SBGT) building supply fan "A" fan belt

WO 112869614 Alarm received when placing SBGT train "A" in service

WO 113290380

BFN-1-FM-084-0020C Drywell or Suppression Chamber Exhaust to SBGT

WO 113748747 Access point for internal inspection of SBGT South Exhaust header duct

WO 114352746 Replace damper motor actuator on SBGT train "B"

WO 114671511 High vibrations on SBGT Train "A"

Section 1R05: Fire Protection

Fire Protection Report Vol. 1, Fire Hazards Analysis, Section 2, Fire Areas 23 and 24, Rev. 14 Fire Protection Report Vol. 2, Appendix V,Section IV, Pre-Fire Plans for Browns Ferry Nuclear Plant - Diesel Generator Building Unit 3, Pre-Plan Nos.

DG3-565 and

DG3-583, Rev. 51

Fire Protection Impairment Permit (FPIP) 09-1920, Appendix R Safe Shutdown Manual Actions FPIP 12-3341, U1 App R

LCO 67-FPR-2012-15

FPIP 12-3342, U1 App R

LCO 23-FPR-2012-14

FPIP 12-3369, 120 VAC Lighting Panels not physically separated FPIP 12-3438, Unit 1/2 A and B EDG Inoperable,

PER 521793 FPIP 12-3459, Unanalyzed Condition Impacting Emergency Diesel Generator Loading

FPIP 12-3560, Unit 3 Scaffolding Support, Various

FPIP 12-3596, App R Cable PP698-IB

FPIP 12-3597, App R Cables ES134-I and ES136-I FPIP 12-3612, 0-AHU-31-88

EL 593' AHU 1A FPIP 12-3696, Ability to open 3-FCV-23-40 RHRHX 3C Outlet Vlv from MCR and Backup Pnl

FPIP 12-3731, 2-FCV-74-67 OOS

FPIP 13-3897, 0-Fan-30-73, 480V Aux BD Rm B Exh Fan DG TDB

FPIP 13-3900, Black Foam Insulation on Pipes in Unit 3 Exclusion Zone

FPIP 13-3931, Cooldown Rate Limits FPIP 13-3943, DG D 125 VDC Battery Charger A Tagged Out FPIP 13-3949, 2B 480V Shutdown Board Deenergized for Maintenance

FPIP 13-3950, App R LCO 0-2253-FPR-2013-77

Fire Protection Report Volume 1, Fire Hazards Analysis, Section 2, Fire Zone 3-2, Rev. 14

Fire Protection Report Volume 2, Appendix Q,Section IV, Pre-Fire Plans for Browns Ferry Nuclear Plant - Reactor Building Unit 3, Pre-Plan Nos.

RX3-519,

RX3-519NE,

RX3-519SE, and

RX3-565, Rev. 51

NPG-SPP 18.4.7 Control of Transient Combustibles, Rev 3

Section 1R07: Heat Sink Performance

NRC Generic Letter (GL) 89-13 and Supplement 1

NPG-SPP-09.7 Corrosion Control Program, Rev 0003

NPG-SPP-09.14

GL 89-13 Implementation, Rev 0001

TVA

GL 89-13 Heat Exchanger Inspection training

TVA Commitment R20

950504 865 Implementation of Alternative Actions to institute

GL 89-13

Electric Power Research Institute (EPRI)

1025271 Heat Exchanger Visual Inspection - Partial Tube Blockage Considerations EPRI 7552 Heat Exchanger Performance Monitoring Guidelines

EPRI

1003320 Supplemental Guidance for Testing and Monitoring Service Water Heat Exchangers

PER 717087

GL 89-13 Alternative response

PER 717975 NRC review of Raw water inspection on

WO 08-717490

Attachment

CRP-ENG-F-12-019 TVA self assessment for Heat Exchanger Programs

ASME

OM-S/G 2007 Part 21 Inservice Performance Testing of Heat Exchangers in Light-Water Reactor Power Plants Work Order (WO)

113635862 Preventative Maintenance performance of 0-TI-63 Residual Heat Removal (RHR) Service Water Flow Blockage Monitoring for the 2A and 2C RHR Heat Exchangers

WO 08-717490 RHR heat exchanger 2A inspection results

WO 114186484 RHR heat exchanger 2A inspection

WO 10671206 RHR heat exchanger 2B inspection results

WO 10671215 RHR heat exchanger 2C inspection results

WO 112074180 RHR heat exchanger 2D inspection results

WO 112529744 RHR heat exchanger 2A and 2C flow blockage monitoring

WO 113635862 RHR heat exchanger 2A and 2C flow blockage monitoring

WO 112095955 RHR heat exchanger 2B and 2D flow blockage monitoring

WO 113410999 RHR heat exchanger 2B and 2D flow blockage monitoring

WO 114091771 Weekly operate each RHRSW pump to provide 30 minutes of chemical treatment to heat exchangers

Section 1R11: Licensed Operator Requalification

NPG-SPP 17.8.1 Rev. 7

Section 1R12: Maintenance Effectiveness

0-TI-346 Maintenance Rule Performance Indicator Monitoring, Trending, and

1-XA-55 panel 5A Annunciator, revision 8 2-XA-55 panel 5A Annunciator, revision 9

Maintenance Rule Expert Panel meeting minutes from 5/23/2013

PER 732555 Fouling of 2A RHR Heat Exchanger tubes by Asiatic clam shells/relics Reporting - 10CFR50.65, Attachment 25, Rev 45 Unit 2 Residual Heat Removal (RHR) system health report dated 5/23/2013

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

0-AOI-100-7, Severe Weather, R34 0-SR-3.8.1.1 (Temporary Diesel Generator Implementation) Rev 15 performance dated 5/20/13

Browns Ferry Action Tracking Number 0-031-TS-2013-0238 for A Control Bay Chiller Browns Ferry License Amendment 280, 307, and 266 to Renewed Facility Operating License Daily Plant Status Report, dated May 29, 2013

DPR-33,

DPR-52, and

DPR-68 to extend Emergency Diesel Generator Allowed Outage time Electrical one line drawing for Aggreko Temporary Diesel Generators Fire Protection Impairment Permit (FPIP) 13-4008 for Emergency Diesel Generator 3D maintenance period

NPG-SPP-09.11.1, Equipment Out of Service (EOOS) Management, Rev. 5 Operators EOOS Risk Profile dated May 23, 2013

Operators EOOS Risk Profile dated May 29, 2013

Operators Logs dated May 23, 2013

Operators Logs dated May 29, 2013

Attachment

Section 1R15: Operability Evaluations

0-47W451-5 Unit 1 and 2 Mechanical Emergency Equipment Cooling Water

0-47W451-6 Mechanical Emergency Equipment Cooling Water

0-47W451-7 Mechanical Emergency Equipment Cooling Water

2-47E859-1 Unit 2 Flow Diagram Emergency Equipment Cooling Water

ASME

OM-S/G 2007 Part 21 Inservice Performance Testing of Heat Exchangers in Light-Water

BFN-VTD-B580-0030 Vendor Manual for High Pressure Coolant Injection System Browns Ferry Unit 3 RHR 3A and 3C heat exchanger performance testing conducted January

25, 2012 Electric Power Research Institute (EPRI)

1025271 Heat Exchanger Visual Inspection - Partial

EPRI

1003320 Supplemental Guidance for Testing and Monitoring Service Water Heat

EPRI 7552 Heat Exchanger Performance Monitoring Guidelines Flowserve Letter, Subject: TVA Browns Ferry HPCI Pump Shaft Damage, S/N 681-S-0004, dated 18 April 2013 MDQ0023980143 RHR heat exchanger tube plugging analysis Rev 3

MMDP-1, Maintenance Management System, Rev. 25

NEDP-27 Past Operability Evaluations, Rev 1

NEI 96-07, Guidelines for 10

CFR 50.59 Implementation, Rev. 1, Dated November 2000

NPG-SPP-09.14

GL 89-13 Implementation, Rev 0001

NPG-SPP-09.20, Vendor Manual Control, Rev. 2

NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Rev. 13

NPG-SPP-09.4, 10

CFR 50.59 Evaluations of Changes, Tests, and Experiments, Rev. 5

NRC Generic Letter (GL) 89-13 and Supplement 1

OPDP-8 Limiting Conditions for Operations Tracking, Rev 0002

PDO for

PER 726761

PER 713047, Damaged Shaft on Unit 2 HPCI Main Pump

PER 714545, Potential Rework Event

PER 718114, Question Posed to BFN QA

PER 720940, Screening Review Not Performed for HPCI Shaft

PER 721450, Improper Inspection Requirements in U2 HPCI Work Instructions

PER 723045, Potential Rework Event

PER 724582, QA

ID 50.59 issue with Vendor Manual change VR# 9911

PER 726761 D2 Residual Heat Removal (RHR) Service Water pump coupling bolts not

PER 732555 Fouling of 2A RHR Heat Exchanger tubes by Asiatic clam shells/relics Removal (RHR) Service Water Flow Blockage Monitoring for the 2A and 2C RHR Heat Exchangers TVA Commitment R20

950504 865 Implementation of Alternative Actions to institute

GL 89-13

TVA

GL 89-13 Heat Exchanger Inspection training

TVA Problem Evaluation Report (PER)

717975 NRC Review of Raw water Inspection on

TVA Prompt Determination of Operability (PDO) for Problem Evaluation Report (PER)

717975

Vendor Manual

BFN-VTD-B580-0030, Byron Jackson - Flowserve, Rev. 2 Vendor Manual Revision Request 13-B0075, Addition of Flowserve Letter Approving Work to Pump Shaft

WO 08-717490 RHR heat exchanger 2A inspection results

WO 08-717490-000

WO 114186484 RHR heat exchanger 2A inspection

Attachment

WO 114500303 Through wall leak in EECW piping

WO 114609716, Disassemble Pump and Bring Ro tating Assembly to Shop for Repairs Work Order (WO)

113635862 Preventative Maintenance performance of 0-TI-63 Residual Heat

Section 1R18: Plant Modifications

Flowserve Letter, TVA Browns Ferry HPCI Pump Shaft Damage, S/N 681-S-0004, dated April

18, 2013

MMDP-1, Maintenance Management System, Rev. 25

NEI 96-07, Appendix E, User's Guide for

NEI 96-07, Rev. 1 "Guidelines for 10

CFR 50.59 Implementation

NEI 96-07, Guidelines for 10

CFR 50.59 Implementation, Rev. 1

NPG-SPP-06.9.3, Post-Modification Testing, Rev. 4

NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Rev. 13

NPG-SPP-09.4, 10

CFR 50.59 Evaluations of Changes, Tests, and Experiments, Rev. 6

NPG-SPP-09-20, Vendor Manual Control, Rev. 2

PER 718114, Question Posed to BFN QA

PER 720940, Screening Review Not Performed for HPCI Shaft

PER 724582, QA Identified 50.59 Issue With Vendor Manual Change VR#9911 Regulatory Guide 1.187, Guidance for Implementation of 10

CFR 50.59, Changes, Tests and Experiments, dated November 2000 UFSAR Section 6.4.1, High Pressure Coolant Injection System, Amendment 24

VR# 13-B0075, Vendor Manual Revision Request for

BFN-VTD-B580-0030, Rev. 2

WO 114609716, Disassemble Pump and Bri ng Assembly to Shop for Repairs

Section 1R19: Post-Maintenance Testing

1-SR-3.5.1.6 (CS I) Core Spray Flow Rate Loop I 2-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping (ASME Section III, Class 1 and 2), Rev. 32

2-SR-3.1.7.7

2-SR-3.5.1.6(CS I-COMP), Core Spray Loop I Comprehensive Pump Test, Rev. 11

2-SR-3.5.1.7(COMP), HPCI Comprehensive Pump Test, Rev. 21 2-SR-3.5.1.9(RHR I) Residual Heat Removal (RHR) Auto Initiation Test 2-SR-3.5.3.4 Reactor Core Isolation Cooling (RCIC) System Rated Flow at Low Reactor Pressure Vessel (RPV) Pressure N-VT-4, System Pressure Test Visual Examination Procedure, Rev. 26

PER 216996, 2-FCV-074-0073 Failed to Open

PER 718035, Procedural Problems Encountered During Performance of New IST Format

PER 719096, Leaks Observed During U2C17 Vessel Pressure Test, 2-SI-3.3.1.A

PER 719951 RCIC System Rated Flow at Low RPV Pressure Unit 1 Technical Specification 3.5.1.6

Unit 2 Technical Specification 3.5.1.9

Unit 2 Technical Specification 3.5.3.4

WO 09-713319-000, Replace INBD & OUTBD Mechanical Seals on 2-PMP-073-0054 (HPCI Main Pump)

WO 111421708, Replace

BFN-2-HS-075-0005A CS Pump 2A,

PER 216996 Action

WO 111421831, Replace

BFN-2-HS-075-0014A CS Pump 2C,

PER 216996 Action

WO 111421886, Replace

BFN-2-HS-075-0023A CS SYS I OTBD INJ VLV,

PER 216996 Action Attachment

WO 112735902, Disassemble and inspect

BFN-2-FCV-073-0019 HPCI Governor Valve - OE PER#393549

WO 113595404,

BFN-2-MVOP-075-0022, Support MMG in Valve Inspection

WO 113759995, ASME Section XI System Leakage Test of Rx Pressure Vessel

WO 113807035, REPLACE HPCI RUPTURE DISCS

BFN-2-RPD -073-0729 & 730

WO 113856706, Unit 2 Standby Liquid Control (SLC) System Pump Functional Test,

WO 113858555, HPCI Comprehensive Pump Test

WO 113870368, Core Spray Loop I Comprehensive Pump Test

WO 114291117, Install HPCI Turbine to Main pump coupling after overspeed trip testing.

WO 114313143, Replace EGR Actuator

BFN-2-SM-073-0019, Send removed EGR to manufacture for refurbishment

WO 114324344, Implement

DCN 69896A,

PIC 70701 to Disassemble

BFN-2-FCV-073-0044, Replace Wedge Ref. DCA69896-007

WO 114399491, MMG TO Implement

DCN 70578 to replace Flowserve - Anchor / Darling Double Disc Gate Valve with Crane

WO 114419329, MMG Disassembly and Inspect the Disc to Disc Skirt Connection

WO 114525711, EOC - Inspect & Refurb 2-ISV-73-23 HPCI Turbine Exhaust Check Valve

Section 1R20: Refueling and Other Outage Activities

2-GOI-100-12, Power Maneuvering

2-GOI-100-1A, Unit Startup

2-GOI-200-2 Primary Containment Initial Entry and Closeout Rev 43 Fatigue Assessments for 12 TVA employees and contractors Hours worked for 11 TVA employees and contractors

NPG-SPP 03.21 Fatigue Management and Work Hour Limits, Rev.10

Section 1R22: Surveillance Testing

0-OI-82 Standby Diesel Generator System, Rev 141

0-TI-362, Inservice Testing Program, Rev. 37

0-TI-364, ASME Section XI System Pressure Tests, Rev. 16

1-SR-3.5.1.6(RHR 1) Quarterly RHR System rated flow test - Loop I, Rev 24

1-SR-3.5.1.6(RHR I) Quarterly Residual Heat Removal (RHR) system rated flow test - Loop I,

2-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping (ASME Section III, Class 1 and 2), Rev. 32

2-SR-3.5.1.1(HPCI), Maintenance of Filled High Pressure Coolant Injection (HPCI) Discharge Piping, Rev 8 2-SR-3.6.1.3.8(4), Instrument Line Excess Flow Check Valve Operability Test, Rev. 7 2-SR-3.6.1.3.8(6), Instrument Line Excess Flow Check Valve Operability Test, Rev. 2 3-SR-3.5.1.7, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated Rx Pressure DWG 0-47W600-20, Mechanical Instruments and Controls, Rev. 11

DWG 2-47E600-58, Mechanical Instruments and Controls, Rev. 12

DWG 2-47E610-3-1, Mechanical Control Diagram, Reactor Feedwater System, Rev. 54

PER 567561, Excess Flow Check Valves Currently Absent from IST Scope

SR 722415

SR 722559

Attachment Unit 2 Technical Specification Bases, 3.6 Containment Systems, B 3.6.1.3 Primary Containment Isolation Valves (PCIVs),

Unit 2 Technical Specifications, 3.6 Containment Systems, 3.6.1.3 Primary Containment Isolation Valves (PCIVs), Amendment No. 253 Unit 2 Technical Specifications, 5.5.6 Inservice Testing Program, Amendment No. 253

WO 113829841, Disassembly Check Valve 2-ECKV-003-0231A, Perform Bench Test

WO 114057643, 1-SR-3.5.1.6(RHR I) Quarterly Residual Heat Removal (RHR) system

WO 114060974

WO 114085048, 3-SR-3.8.1.1(3A) Diesel Generator "3A" Monthly Operability Test

WO 114336097, Instrument Line Flow Check Valve Operability Test

WO 114636903, Instrument Line Excess Fl ow Check Valve Operability Test

WO 114789289 for surveillance 2-SR-3.4.4.1 Manual Calculation of Unidentified, Identified, and Total Leakage

Section 2RS1: Radiological Hazard Assessment and Exposure Controls

Procedures, Guidance Documents, and Manuals

RCI-1.2, Radiation, Contamination and Airborne Surveys, Rev. 19

RCI-17, Control of High Radiation Areas and Very High Radiation Areas, Rev. 73

RCI-23, Hot Spot Tracking Program, Rev. 12

RCI-24, Control of Vacuum Cleaners and Portable HEPA Units within the RCA, Rev. 27

RCI-43, Radioactive Material Control, Rev. 4

RCI-47, Diving Operations in the Radiologically Controlled Area, Rev. 1

Records and Data

NSTS Inventory List

NSTS Transaction Confirmation Email for Unit 1

NSTS Transaction Confirmation Email for Unit 3

Browns Ferry Sealed Source List Completed 0-TI-540, Storage of Material in the Spent Fuel Storage Pool and Transfer Canal, Dated 7/20/2012

Radiation Protection Plan for Monitoring and Controlling Hazards Associated with Potential Fuel Leakers Browns Ferry Nuclear Plant - Quality Assurance- Fleet Assessment of Radiation Protection Control of Contamination, Control of Materials, and Control of Radiological Instrumentation Report-

QA-BF-11-005 White Paper: Unit 3 Drywell Neutron Measurements, June 1981

White Paper:

Technical Evaluation Results and HPIC Recommendations for RadCon Instrumentation Response Checks, 2/16/2000 White Paper: Report on Beta Factors for Portable Instrumentation, 11/30/1987 Historical Work Order Report for Work on Radwaste Processing Systems to include Supporting Systems (Systems 33,77 and 78 Survey: M-20130316-37, U-2 Rx Bldg 664' Rx Cavity Initial Survey Survey M-20130319-2, U2R17 Under RPV Head Survey Survey M-20130315-26, U2R17 Sub Pile Room FME Screen Replacement Survey M20130304-17, M0064 Update of 550' Elevation Contamination Levels Survey M-20130315-31, M0066 LHRA Down post Survey of 584' Elevation Sampling of Airborne Radiation Surveys for 3/17/2013

Attachment Sampling of Airborne Radiation Surveys for 4/3/2013

Sampling of Airborne Radiation Surveys for 4/4/2013

Sampling of Airborne Radiation Surveys for 4/9/2013

RWP 13270004

CAP Documents

PER 546870

PER 639446

PER 665174

PER 672053

PER 537742

PER 552800

PER 605414 PER 645715

Section 2RS2: Occupational

ALARA Planning and Controls Procedures, Guidance Documents and Manuals

NPG-SPP-05.2, ALARA Program, Rev 3

NPG-SPP-05.2.1, Operational ALARA Planning and Controls, Rev 2

NPG-SPP-05.2.2, Establishing Collective Radiation Exposure Annual Business Plan Goals,

Rev 0

NPG-SPP-05.2.3, Outage Exposure Estimating and Tracking, Rev 0

RCI 9.1, Radiation Work Permits, Rev 72

RCI-15.2, Temporary Shielding, Rev 26

RCI-27, Source Term Reduction and Control Tennessee Valley Authority Browns Ferry Nuclear Power Plant Long-Term Collective Radiation Exposure Reduction Plan 2013-2017

Records and Data

ALARA Plan 12-0015, U3R15 Refuel Outage Under vessel Maintenance

ALARA Plan 12-0041, U3R15 Outage Refuel Floor Maintenance Activities

ALARA Plan 12-0025 U2R17 Outage Refuel Floor Maintenance Activities

ALARA Plan 13-0010, U2R17 Refuel Outage Under vessel Maintenance

ALARA Plan 13-0023, U2R17 JOG, EOC, and Other Valve Work Activities ALARA Plan 13-0025, U2R17

DRW-Disassemble and Refurbish 2-FCV-069-0001 Valve ALARA Plan 13-0027, U2R17 Outage Refuel Floor Maintenance Activities

ALARA Plan Post-Job Review 12-0015, U3R15 Outage - Under vessel Work Activities

ALARA Plan Post-Job Review 12-0041, U3R15 Outage Refuel Floor Maintenance Activities

ALARA Plan Post-Job Review 13-0010, U2R17 Outage - Under vessel Work Activities ALARA Post-Job Review 12-0025, Replacement of 3A and 3C RHR motors Station ALARA Committee Meeting Minutes, 8/12, 9/12, 10/12, 12/12, 1/13, 2/13, 3/13

Temporary Shielding Request 13-0004, U2R17 Outage Activities

Attachment

RWPS

13270081, Rev 1, Unit-2R17 Outage RX Bldg

13270224, Rev 0, Unit-2R17 Outage Turbine

280258, Rev 0, Unit-2R17 Outage Drywell

280259, Rev 0, Unit-2R17 Outage Drywell

280274, Rev 0, Unit-2R17 Outage Drywell

290004, Rev 0, Unit-2R17 Outage Refuel Floor Maintenance Activities Radiological Surveys

M-20121116-5, M0251 Unit 2

RXB 639'General Area

M-20130315-31, M0066 Unit 2 Drywell 584'

M-20130316-9, M0066 Unit 2 Drywell 584'

M-20130316-22, M0066 Unit 2 Drywell 584'

M-20130320-64, M0066 Unit 2 Drywell 584'

M-20130327-10, M0398 CRD Assembly

M-20130331-40, M0075 Unit 2

RXB 621 General Area

M-20130331-10, M0251 Unit 2

RXB 639' General Area

M-20130401-40, M0456 Unit 2

RXB 565 SW Corner

M-20130406-33, M0098 Unit 2

RXB 565' Steam Tunnel

M-20130407-26, M0433 Service Building 565 Small Tool Decon

M-20130408-4, M0066 Unit 2 Drywell 584'

M-20130410-11, M0251 Unit 2

RXB 639'General Area

CAP Documents

BFN-RP-S-11-001, Snap Shot Self-Assessment, October 3rd-14th, 2011

PER 333029

PER 340519

PER 342266

PER 346571

PER 351622

PER 429768

PER 438039

PER 526373

PER 633006

Section 2RS3: In-Plant Airborne

Radioactivity Control And Mitigation

Procedures

and Guidance Documents

FP-0-000-INS027, Self Contained Breathing Apparatus (SCBA), Rev. 15

FPT-158.049 Breathing Air System Qualificaiton Manual, Rev. 0 Lesson Plan:

HPT-063.002, Dated 10/16/2012

RCI-3.1, Respiratory Protection Program Implementation, Rev. 36

RCI-1.2, Radiation, Contamination and Airborne Surveys, Rev.19

RCI-11.2, Radiation Protection Airborne Instrument Maintenance, Rev. 5

Records and Data Reviewed Qualification Matrix for Respiratory Protection as of 1/29/2013 Breathing Air/Gas certification 1/11/13 Certificate for Gidran Compressor

115722 SN:

505110969A

Attachment Breathing Air/Gas certification 1/11/13 Certificate for Gidran Compressor

115723 SN:

0505110968A Work Order

113713732, Monthly inspection of SCBA kits Work Order

113713734, Monthly Inspection and Maintain SCBA Kits per Requirements in FP-0-

000-INS027

RCI-35 DOP Testing of Vacuum Cleaners and Portable HEPA Filters, Rev. 3 Performed February thru May 2012 and October and November 2012. MSA MMR Certified CARE Technicians certificates, August 3, 2011 and August 3, 2013

Corrective Action Program (CAP) Documents Self Assessment Report

BFN-RP-S-12-002, Radiological Hazard Assessment and Exposure Controls, October 17-21, 2011.

PER 540458 PER540518

PER 541114

PER 542892

PER 542800

PER 558515

PER 501980

PER 627928

PER 649509

PER 558510

Section 2RS4: Occupational Dose Assessment

Procedures

and Guidance Documents

NPG-SPP-05.1, Radiological Controls, Rev. 3

RCDP-7, Bioassay and Internal Dose Program, Rev. 4

RCI-1.1, Radiation Operations Program Implementation, Rev.153

RCI-2.1, External Dosimetry Program Implementation, Rev. 61

RCI-8.1, Internal Dosimetry Program Implementation, Rev. 49

RCTP-105, Personnel In-processing and Dosimetry Administrative Processes, Rev. 4

RCTP-106, Special Dosimetry Operations, Rev. 2

Records and Data Reviewed

NVLAP Certifications for Landaur OSLs and Panasonic TLDs

Corrective Action Program (CAP) Documents See Section 2RS1

Section 2RS5: Radiation Monitoring Instrumentation

Procedures, Guidance Documents and Manuals Browns Ferry Nuclear Plant Units 1, 2 and 3, Offsite Dose Calculation Manual, Rev. 21

Chemistry Instructions:

CI-901, Post Accident Sampling Procedure for Stack Effluents - Mid/High Range Particulate and Iodine Filter Removal from WRGERMS, Rev. 0014

CI-1107, Procedure for Obtaining Monthly QA Data for Germanium Detectors, Rev. 0004

Attachment Component Calibration Instructions:

CCI-0-RM-90-146, Area Radiation Monitors Calibration, Rev. 0025

CCI-0-RM-90-100, Calibration of the Canberra ARGOS 5AB Personnel Contamination Monitor, Rev. 0001 Emergency Plan Implementing Procedure,

EPIP-1, Emergency Classification Procedure, Rev.

0049

Memorandum L91

000217 800, Technical Evaluation Results and HPIC Recommendations for RADCON Instrumentation Response Checks, 2/6/2000

Memorandum L91

871130 800, Report on Beta Factors for Portable Instrumentation, 11/30/1987 NPG Standard Programs and Processes procedures:

NPG-SPP-05.0, Radiological and Chemistry Control, Rev. 0002

NPG-SPP-03.1, Corrective Action Program, Rev. 0005 Radiological Control Instructions:

RCI-11.1, Radiation Protection Instrument Program Implementation, Rev. 0073

RCI-11.2, Radiation Protection Airborne Instrument Maintenance, Rev. 0005

RCI-11.3, Radiation Protection Radiation/Contamination Instrument Maintenance, Rev.

28

RCI-11.6, Calibration of the Canberra ARGOS 5AB Personnel Contamination Monitor, Rev.

0000

RCI-11.7, Calibration of the Eberline PCM 2 Personnel Contamination Monitor, Rev. 0001

RCI-11.8, Calibration of the Eberline

PCM-1B Personnel Contamination Monitor, Rev. 0000

RCI-11.9, Calibration of the Eberline

PM-7 Portal Monitor, Rev. 0002

RCTP-102, Calibration of Canberra

GEM-5 Personnel Monitors, Rev. 0000

RCTP-109, Whole Body Counting and Calibration, Rev. 0000 Surveillance Instruction, 0-SI-4.2.K.1, Airborne Effluents - Main Stack Monitoring System Calibration, Rev. 0034 Surveillance Instruction, 2-SI-4.2.K.2, Reactor Building Vent Exhaust Radiation Monitor Calibration and Functional Test 2-RM-90-250, Rev. 0002

Surveillance Procedure, 1-SR-3.3.3.1.4(5A), Containment High Range Radiation Monitoring Channel Calibration and Functional Test (1-RM-90-272A) Division I, Rev. 0006

Surveillance Procedure, 1-SR-3.3.3.1.4(5B), Containment High Range Radiation Monitoring Channel Calibration and Functional Test (1-RM-90-273A) Division II, Rev. 0007 Technical Instruction, 0-TI-15, Radioactive Gaseous Effluent Engineering Calculations and Measurements, Rev. 0016 Technical Instruction, 0-TI-45, Liquid Process Radiation Monitors, Rev. 0017

Records and Data Reviewed

ARGOS Calibration Package, TVA Tag # 00RE90-100, S/N 0901-100, 4/30/12 ARGOS Calibration Package, TVA Tag # 00RE90-203, S/N 1008-203, 8/24/12 and11/12/12 ARGOS Calibration Package, TVA Tag # 00RE90-206, S/N 1008-206, 8/28/12

Browns Ferry System 90 - Radiation Monitoring System Health Reports, Units 1, 2 and 3 for the period 2/1/2012 - 1/31/2013

Browns Ferry Nuclear Plant PC Based Whole Body Counter (PCWBC) Software Requirements Specification, Rev. 1 Calibration Package, Gamma Spectroscopy System, Detector #2, 2/21/2001 Calibration Package, Gamma Spectroscopy System, Detector #7, 12/20/2012

Canberra FastScan Whole Body Counting System Calibration Package, 2/6/2013

Attachment Certificate of Calibration, Digital Flow Calibrators: S/N 3275, 6/21/2012 and 8/20/2012; S/N 3075, 6/21/2011 and 8/20/2012; Certificate of Calibration, Model 878-10 Calibrator, S/N 107, 7/24/1984; S/N 131, 8/3/2012

Certificate of Calibration, Co-60 Source, S/N 22688-135, 2/12/1987; Ba-133 Source, S/N 22695-135, 2/12/1987; Ba-133 Source, S/N 22696-135, 2/12/1987; Ba-133 Source, S/N 22697-

135, 2/12/1987; Tc-99 Source, S/N 033/88, 1/7/1988; Tc-99 Source, S/N 034/88, 1/7/1988;

Tc-99 Source, S/N 037/88, 1/7/1988; Tc-99 Source, S/N 041/88, 1/7/1988; Tc-99 Source, S/N I5-478, 11/15/2011; Am-241 Source, S/N 1742/90, 3/13/1990; Cs-137 Source, S/N 1555-44-1, 11/15/2011; Cs-137 Source, S/N 2043-95, 5/10/1995; Cs-137 Source, S/N G1-

979, 6/15/2009; and Mixed Gamma source, S/N 86251A-143, 1/10/2012 Certificate of Calibration, Portable Calibration Unit 846D959G001, S/N 190214805-2, 6/1/2009

Daily Background and Quality Control Check Charts, Gamma Spectroscopy System, Detector #7, for the period 1/1 - 4/11/2013 Form

INST-2, Instrument Source and Response Check Form, multiple instruments, 4/9/13 and

4/10/13 Form

INST-13, Contamination Monitor Performance Log, multiple instruments, 4/8/13 and

4/10/13

Memorandum W46

120906 001, Calibration of the Western Area Radiological Laboratory (WARL) High Level Cs-137 Source Range, 9/6/2012

Memorandum, Subject: Request for Funding to Improve Instrumentation for Radiation Safety, 4/2/2013

PCM-2 Calibration Package, TVA Tag # 00RE90-477, S/N

PCM-2 477, 2/18/12 and 4/6/12

PCM-2 Calibration Package, TVA Tag # 00RE90-479, S/N

PCM-2 479, 4/7/12

PM-7 Calibration Package, TVA Tag # 00RE90-1001, S/N 477, 4/11/12 and 6/12/12

PM-7 Calibration Package, TVA Tag # 00RE90-1007, S/N 369, 4/12/12 and 6/12/12

Small Article Monitor (SAM-11) Calibration Package, TVA Tag #

841990, 11/19/2011 and

10/18/2012 Small Article Monitor (SAM-11) Calibration Package, TVA Tag #

860215, 10/13/2010 and

3/20/2012 Special Test 87-03, Browns Ferry Nuclear Plant (BFN) - Off-Line Liquid Rad Monitor Isotopic/Transfer Calibration, memo dated 3/23/1987 Special Test 87-05, Browns Ferry Nuclear Plant (BFN) - General Electric (GE) Stack Gas Monitor Isotopic/Transfer Calibration, memo dated 5/1/1987 Special Test 87-14, Browns Ferry Nuclear Plant (BFN) - Liquid Radwaste Radiation Monitor Isotopic/Transfer Calibration, memo dated 5/1/1987

Surveillance, 1-SR-3.3.3.1.4(5A), Contai nment High Range Radiation Monitoring Channel Calibration and Functional Test (1-RM-90-272A) Division I, 4/14/2009

Surveillance, 1-SR-3.3.3.1.4(5B), Contai nment High Range Radiation Monitoring Channel Calibration and Functional Test (1-RM-90-273A) Division II, 3/25/2009 Work Order #

112074903, 2-SI-4.2.K.2, Reactor Building Vent Exhaust Radiation Monitor Calibration and Functional Test 2-RM-90-250, 11/23/2011 Work Order #

113613629, 2-SI-4.2.K.2, Reactor Building Vent Exhaust Radiation Monitor Calibration and Functional Test 2-RM-90-250, 6/20/2012 Work Order #

112702543, 0-SI-4.2.D.1, Liquid Radwaste Monitor Calibration/Functional Test, 9/13/2011 Work Order #

113332176, 0-SI-4.2.D.1, Liquid Radwaste Monitor Calibration/Functional Test, 7/18/2012

Attachment Work Order #

112814333, 0-SI-4.2.K.1, Airborne Effluents - Main Stack Monitoring System Calibration, 8/10/2012 Work Order #

111718372, 0-SI-4.2.K.1, Airborne Effluents - Main Stack Monitoring System Calibration, 12/10/2010

Work Order #

113332185, 0-SI-4.2.K.1FT, Airborne Effluents - Main Stack Monitoring System Functional Test, 7/23/2012

Work Order #

113592624, 0-SI-4.2.K.1FT, Airborne Effluents - Main Stack Monitoring System Functional Test, 10/27/2012 Work Order #

111202218, PM and Calibrate Area Radiation Monitors 1-RE-90-21 through 29

per

CCI-0-RM-90-146, 2/7/2011

Work Order #

111671614, 1-SR-3.3.3.1.4(5A), Containment High Range Radiation Monitoring Channel Calibration and Functional Test (1-RM-90-272A) Division I, 5/8/2011

Work Order #

111430001, 1-SR-3.3.3.1.4(5B), Containment High Range Radiation Monitoring Channel Calibration and Functional Test (1-RM-90-273A) Division II, 3/11/2011 Work Order #

112225792, Perform Calibrations on Argos units 100, 202, 206, 204, 205, and 203 to maintain RadCon periodicity, 11/30/2011 Work Order #

111827099, PCM's and

PM-7's need I&C maintenance to get them in serviceable working order for the upcoming outage [PCM S/N's 006, 388, 479, 477, and 478], 2/24/2011 Whole Body Counter Analysis Radionuclide Libraries, 4/10/2013

Corrective Action Program (CAP) Documents

NPG-SPP-03.1, Corrective Action Program, Rev.0005

Common Factor Analysis for

PER 578067, Continuing Trend of RP Instrument Failures, 10/1/2012 Self Assessment Report

CRP-TPR-F-10-002, Port able Radiation Protection Instrumentation, 6/29/2010

Problem Evaluation Reports (PERs)

PER 379628

PER 382492

PER 404596

PER 495968

PER 499901

PER 509371

PER 514991

PER 526282

PER 561910

Section 4OA1: Performance Indicator Verification

Procedures

Records and Data Reviewed Summation of liquid releases, January 1-February 28, 2013

Liquid Waste Release Permit 130325.092.060.L

2012 -2013 Dose Rate Alarms.xlsx Access Control Alarm Report. Pdf - Generated 3/28/2013

RWPs

Documented in 2RS1

Attachment

PERS Documents

PER 513987

PER 515462

PER 516443

PER 539409

PER 546950

PER 585337

PER 619986

PER 656140

PER 689146

Section 4OA1: Performance Indicator (PI) Verification

PER 235338 HPCI isolation during performance of 2-SR-3.3.6.1.6(3)

PER 237470 Unit 1 HPCI inoperability

PER 238663 RCIC declared inoperable due to 2-SR-3.5.3.3 failure

PER 246526 Unit 3 RCIC system inoperable beyond Allowed Outage Time

PER 283357 RCIC turbine governor end bearing failure

PER 283639 Unit 1 RCIC clogged orifice in turbine lube oil piping

PER 293928 Preventable Functional Failure of Unit 1 RCIC

PER 372706 HPCI instrument damage due to 1-FCV-73-45 failure

PER 374831 Unit 1 HPCI declared inoperable

PER 377833 Unit 1 HPCI failure to trip during testing

PER 427221 Unplanned Limiting Condition for Operation entry due to low RCIC oil level

PER 511607 Unit 1 HPCI steam admission valve

PER 551197 RCIC overspeed test not completed

PER 596706 Unit 2 HPCI isolation during 2-SR-3.3.6.1.3(3DFT)

PER 646066 Unplanned RCIC isolation

PER 649345 Unit 1 HPCI 1-FCF-73-6B stroke time failure

PER 692026 RCIC inoperability not reflected in the schedule

PER 697691 HPCI Logic Functional Test unsat due to 73-81 inoperability Service Request (SR)

731456 Incorrect reporting of Unit 3 RCIC and HPCI ESF demand and run hours Unit 1 Mitigating Systems Performance Index (MSPI) Basis Document, Revs 10, 11, 12, and 13

Unit 1 MSPI Derivation Report for High Pressure Coolant Injection (HPCI) Unreliability Index (URI) dated Jun 2012, Sep 2012, Dec 2012, and Mar 2013 Unit 1 MSPI Derivation Report for HPCI Unavailability Index (UAI); Jun 2012, Sep 2012, Dec

2012, and Mar 2013 Unit 1 MSPI Derivation Report for RCIC UAI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013

Unit 1 MSPI Derivation Report for Reactor Core Isolation Cooling (RCIC) URI; Jun 2012, Sep

2012, Dec 2012, and Mar 2013 Unit 2 MSPI Basis Document, Revs 9, 10, 11, and 12

Unit 2 MSPI Derivation Report for HPCI UAI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013

Unit 2 MSPI Derivation Report for HPCI URI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013

Unit 2 MSPI Derivation Report for RCIC UAI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013

Unit 2 MSPI Derivation Report for RCIC URI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013 Unit 3 MSPI Basis Document, Revs 8, 9, 10, and 11 Unit 3 MSPI Derivation Report for HPCI UAI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013

Attachment Unit 3 MSPI Derivation Report for HPCI URI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013 Unit 3 MSPI Derivation Report for RCIC UAI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013 Unit 3 MSPI Derivation Report for RCIC URI; Jun 2012, Sep 2012, Dec 2012, and Mar 2013

Section 4OA2: Identification and Resolution of Problems

0-TI-414, Component Labeling, Signs, Operator Aids, and Permanent Information Postings, Rev. 6

ACE 676061, RCIC Outage Extension Beyond 50 percent of Scheduled Duration, Rev. 1

BFN 'Trend' PERs Year-to-Date

BFN-OPS-S-13-007, Operator Aid/ Permanent Information Postings Self-Assessment Browns Ferry Nuclear Plant Performance Improvement / HU Plan, dated 4/27/2013

Integrated Trend Report Q1FY13, Analysis of Site Performance for the Period October 1 - December 31, 2012 Integrated Trend Report Q2FY13, Analysis of Site Performance for the Period January 1 -

March 31, 2013

NPG-SPP-02.7, PER Trending, Rev. 4

NPG-SPP-02.8, Integrated Trend Review, Rev. 4 Ops Labeling Request for PIP Audit 2011

PER 344224 (SR 343406), Audit not conducted on OA/PIP program as required by 0-TI-414

PER 670901, No Work Performed on Tagged Equip Due to No Parts Available

PER 671384, RHRSW Pump D3 IST Quarterly Pump Test is a Terrible Procedure

PER 676061, RCIC Outage Extension Beyond 50 percent of Scheduled Duration

PER 695276, Work Delay Causing a Challenge to Safe Operations

PER 702110, Gaps Identified in CIPTE Brief for CRD Exchange

PER 702679, Responsible Supervisor does not Ensure CIPTE Brief for CRD Exchange

PER 707725, Briefing for OPDRV for Repair of 2-FCV-69-1

PER 715551, Fire / Medical Emergency

PER 718099, CIPTE Briefing Stopped

PER 724721, CIPTE Test Director(s) and CIPTE Manager Not Assigned in a Timely Manner

PER 726796, Deficiencies noted during 2013 OA/PIP audit

PER 745955, Fire Ops Shift Staffing Relative to Alarm Response

PER 746111, NRC Resident Question about the Quality of CIPTE Briefs Performance Improvement Peer Team Successes, dated May 16, 2013

Potential Trend & Common Factors Analysis Topics - PSC Guidance Document, dated

4/4/2013

SR 745967, Missing Permanent Information Posting

SR 745969, Missing Permanent Information Posting

SR 746017, Question about Long Standing TACF's

SR 746677, Potential Trend in Equipment Tagged with No Work Performed

SR369515, Issues from 2011 Audit Trending High Impact Team Charter, dated 6/15/2013

U2R17 BFN Procedure Use & Adherence Intervention Plan, dated 3/19/2013

Attachment

Section 4OA3: Event Follow-up

0-TI-346 Maintenance Rule Performance Indicator Monitoring, Trending, and Core Isolation Cooling (RCIC) system

LER 50-259/2013-002-00, Manual Reactor Shutdown Due to Decreasing Condenser Vacuum

LER 50-259/2013-002-00, Manual Reactor Shutdown Due to Decreasing Vacuum, dated May

20, 2013

LER 50-296/2013-002-01, Manual Actuation of RCIC System during Reactor Shutdown

LER 50-296/2013-003-00, Automatic Reactor Shutdown Due to an Actuation of the Reactor Protection System from a Turbine Trip, dated April 26, 2013

NPG-SPP-02.3, Operating Experience Program, Rev. 5

PER 511111, Condenser Air Inleakage on Unit 1

PER 687732, Automatic Scram Due to RPS Actuation

PER 698870, U1 Manual Reactor Scram Due to Degrading Condenser Vacuum

PER 703283,

REC 687732 on U3 Vacuum Loss Scram OE Preventable

PER 709733,

PER 698870 was OE preventable

RCA 516455, Operation Focus / Decision Making, Rev. 1

RCA 687732, U3 Feedwater Long Cycle Return Line Connection Separation from Miscellaneous Drain Header Resulting in Automatic Reactor Scram Due to Loss of Vacuum, Rev. 0

RCA 698870 , U1 Manual Reactor Scram Due to Degrading Condenser Vacuum, Rev. 1 Reporting - 10CFR50.65, Attachment 3, Rev 45 Root Cause Analysis for

PER 710206 Event R

eport #48904 Manual Initiation of Reactor

SR 527942, Heater Drain Line in Ceiling of U1 Turbine BLDG Elevation 557 is Vibrating Excessively

SR 727450, WO to Install External Thermocouples

Tech Spec Bases 3.7.5, Main Turbine Bypass System, Rev. 0 UFSAR 14.5.2.2, Generator Load Reject (TCV Fast Closure) with Turbine Bypass Valve Failure (LRNBP), Amendment 22

WO 113258460, BELZONA / Install Rigging To Stabilize Pipe Separation

WO 113357027 Heater Drain Line in Ceiling of U1 Turbine BLDG Elevation 557 is Vibrating Excessively

WO 114512929, Repair / Replace Rod Hanger

BFN-1-MISC-006

WO 114516358, Steam leak identified on the 4" Misc Dr Header weld-o-let to the 24" 1C2

Heater High Level Dump line

WO 114522755, Replace

BFN-1-DRV-006-0557

WO 114522757, Replace 1-DRV-006-0612, 1C1 Tube Drain Valve

WO 114522759, Replace 1C3 Tube Side Relief Valve BFN-1-RFV-006-0652

WO 114695582, WO to Install External Thermocouples

Section 4OA5: Other Activities

Corrective Action Documents Reviewed

PER 214627, Potential Air Line Leak

PER 216735, Outside Loop Fire Protection Not Able to Maintain Pressure Due to Leaks in

System

PER 284309, Fire Protection System Leak in Parking Lot at Facilities Shop

PER 287775, Fire Protection Leak At facilities Office Area

PER 294614, 1" Control Air Line is Pitted and Degraded

PER 437896, Water Leak at Modification Building Attachment

PER 447870,

SR 445165 Closed to

WO 112805941 without PER for Inspection/Cause

PER 569926, Document Station Drainage Pipe Failure Apparent Cause Evaluation

PER 384210, Water and Dirt found in 'A' Train of SGBT during inspections

Corrective Action Documents Generated

PER 721808, Buried Piping Program Owner and Backup

PER 724720, Excavation Plan and checklist

PER 724731, Follow-up of Open Items in

ACE 384210

PER 730396, Procedure

TI-561

PER 733030, Cathodic Protection Design Study

PER 732294, UPTI Milestones Completion

PER 733056, Maintenance Program

PER 733332, Revise Procedure to address Visual Inspection

PER 733325, UPTI Program Staffing

PER 733343, Automatic Trending of Buried Piping Conditions

PER 733376, Update Procedure for Drawing Locations

PER 733379, Update Procedure for Chase and Vault Monitoring and Protection

PER 734019, Collect Soil samples for Cathodic Protection Design Study

Procedure

0-TI-412, Work Permits, Rev. 29

0-TI-417, Inspection of Service Level I, II, III Protective Coatings, Moisture Seal Barrier, Condensate Storage Tanks, and ECCS Suction Strainers, Rev. 19 0-TI-561, Underground Piping and Tanks Integrity Program (UPTI), Rev. 13

NPG-SPP-03.1.5, Apparent Cause Evaluations, Rev. 8

NPG-SPP-03.1.7, PER Analysis, Actions, Closures and Approvals, Rev. 12

NPG-SPP-05.15, Fleet Ground Water Protection Program, Rev. 3

NPG-SPP-09.7, Corrosion Control Program, Rev. 4

NPG-SPP-09.15, Underground Piping and Tanks Integrity Program (UPTI), Rev. 5

Other Documents

EPRI

TR 1016456, Recommendations for an Effective Program to Control the Degradation of Buried Pipe Nuclear Energy Institute (NEI) 09-14, Guideline for the Management of Buried Piping Integrity, Rev. 3 Permit and Excavation Approval Form under safety manual Procedure, 804

Program Health Report, 1/1/2013-6/30/2013

Program Health Report, 7/1/2012-12/31/2012 DCN#66434, HDPE Pipe replacement for EECW Piping/Station Drainage Report No. R0900212.401, Buried Piping Program Indirect Inspections (Soil Analysis), Rev. 2

Report No. R06130129116, Underground Piping and Tanks Integrity Program Inspection Plan, Rev. 2 Report No. R06130524548, UPTI Program Health Color Document Report No. 1200931.401, Sequoyah Nuclear Plant Buried Piping Cathodic Protection Design Study, Rev. 0 Report No. 0801.107-01, Buried Piping Risk Ranking, Rev. 0

Report No.

CRP-ENG-10-BM03, Self Assessment/ Benchmarking Report Attachment Report No. R06

121220058, Condition Assessment-Underground Piping and Tanks Report No.

BFN-ENG-S-13-014, Readiness for NRC

TI 2515/182 Phase 1 Inspection Report No. 0901180.401.R0, GWT/A-Scan Assessment of Various Common Unit Report No. EPIX 778, Sewer Inspection Report No. R06130530563, Training Center Excavation Report No. R061305557, Summary of RHRSW/EECW Excavation/Inspection Report No. R06130530560, Fire Protection Pipes Near Cooling Tower 7

Report No. R06130530562, Unit 2 ECCW Catch Basis Effluent

Below Ground Piping, Rev. 0

Work Order No.

111424600, Find Leak and Repair Degraded Radwaste Pipe Work Order No.

113416710, 2 Leaks on the Fire Protection Outside Loop Underground Header Work Order No. 10710139000, Excavation of 2 sites in the protected area for the purpose of direct inspection (visual, UT) Work Order No.

112942647, Control Valve Removal Inspection Work Order No.

112124272, Cooling Tower 2 Fire Protection Lines Work Order No.

111846969, Cooling Tower 7 Fire Protection Lines Work Order No.

110692468, Demineralized Water Lines Work Order No.

114394625, Service Air Inspection Work Order No.

114299991, Unit 3 CCW Tunnel Inspection Work Order No. 08721794000, Isolation Valve Replacement in Fire Protection Lines

CFR 72.212, Report of Evaluations, Rev. 5.1 dated 9/4/2012

CFR 72.48 Screening Review, 0-SR-DCS3.1.2.1, Rev. 12

TN 0013

CFR 72.48 Screening Review,

EDC 70586A, Use of HBF IAW Holtec CoC Amend. 5 and FSAR Rev. 7, Rev. 1

NPG-SPP-09.9, "10

CFR 72.48 Evaluations of Changes, Tests and Experiments for Independent Spent Fuel Storage Installation", Rev. 1

LIST OF ACRONYMS

ADAMS - Agencywide Document Access and Management System

ADS - Automatic Depressurization System

ARM - area radiation monitor

CAD - containment air dilution

CAP - corrective action program

CCW - condenser circulating water

CFR - Code of Federal Regulations CoC - certificate of compliance

CRD - control rod drive

CS - core spray

DCN - design change notice

EECW - emergency equipment cooling water

EDG - emergency diesel generator

FE - functional evaluation

FPR - Fire Protection Report

FSAR - Final Safety Analysis Report

IMC - Inspection Manual Chapter

LER - licensee event report

NCV - non-cited violation

NRC - U.S. Nuclear Regulatory Commission

ODCM - Off-Site Dose Calculation Manual

PER - problem evaluation report

PCIV - primary containment isolation valve

PI - performance indicator

RCE - Root Cause Evaluation

RCW - Raw Cooling Water

RG - Regulatory Guide

RHR - residual heat removal

RHRSW - residual heat removal service water

RTP - rated thermal power

RPS - reactor protection system

RWP - radiation work permit

SDP - significance determination process

SBGT - standby gas treatment

SLC - standby liquid control

SNM - special nuclear material

SRV - safety relief valve

SSC - structure, system, or component

TI - Temporary Instruction

TIP - transverse in-core probe

TRM - Technical Requirements Manual

TS - Technical Specification(s)

UFSAR - Updated Final Safety Analysis Report

URI - unresolved item

WO - work order