Text

IR 05000259-12-005, 05000260-12-005, 05000296-12-005; 10/01/2012 - 12/31/2012; Browns Ferry Nuclear Plant, Units 1, 2 and 3; Event Follow-up and Identification and Resolution of Problems
	ML13039A321
Person / Time
Site:	Browns Ferry
Issue date:	02/08/2013
From:	Eugene Guthrie; Division Reactor Projects II
To:	James Shea; Tennessee Valley Authority
References
	IR-12-005
	Download: ML13039A321 (49)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II

245 PEACHTREE CENTER AVENUE NE, SUITE 1200

ATLANTA, GEORGIA 30303-1257

February 8, 2013

Mr. Joseph W. Shea

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 05000259/2012005, 05000260/2012005, AND 05000296/2012005

Dear Mr. Shea:

On December 31, 2012, 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 January 11, 2013, with Mr.

Steve Bono, General Manager Site Operations, and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations, orders, and with the conditions of your

license. The inspectors reviewed selected procedures and records, observed activities, and

interviewed personnel.

Two licensee-identified violations of very low safety significance (Green) were identified during

the inspection. The NRC is treating the violations as a non-cited violations (NCV) consistent

with Section 2.3.2 of the Enforcement Policy. If you contest this non-cited violation, 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.

J. Shea 2

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its

enclosure, and your response (if any), will be available electronically for public inspection in the

NRC Public Document Room or from the Publicly Available Records (PARS) component of the

NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at

http://www.nrc.gov/reading-rm/adams.html.

Sincerely,

/RA/

Eugene F. Guthrie, Chief

Special Project, Browns Ferry

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 05000259/2012005,

05000260/2012005, and 05000296/2012005

cc w/encl. (See page 3)

_________________________ X SUNSI REVIEW COMPLETE G FORM 665 ATTACHED

OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRS

SIGNATURE Via email Via Telecon Via Telecon Via Telecon Via Telecon Via email Via email

NAME DDumbacher CStancil PNiebaum LPressley TStephen DHardage LSuggs

DATE 2/8/2013 2/8/2013 2/8/2013 2/8/2013 2/8/2013 02/07/2013 1/31/2013

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

OFFICE RII:DRS NSIR RII:DRS RII:DRS RII:DRS RII:DRP RII:DRP

SIGNATURE Via email Via email Via email Via email Via Telecon CRK /RA/ EFG /RA/

NAME ASengupta JLaughlin GLaska KSchaaf RBaldwin CKontz EGuthrie

DATE 2/8/2013 02/07/2013 02/07/2013 02/07/2013 2/8/2013 02/08/2013 02/08/2013

E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO

J. Shea 3

cc w/encl:

K. J. Polson

Site Vice President

Browns Ferry Nuclear Plant

Tennessee Valley Authority

Electronic Mail Distribution

S. M. Bono

Plant Manager

Browns Ferry Nuclear Plant

Tennessee Valley Authority

Electronic Mail Distribution

James E. Emens

Manager, Licensing

Browns Ferry Nuclear Plant

Tennessee Valley Authority

Electronic Mail Distribution

E. W. Cobey

Manager, Corporate Licensing

Browns Ferry Nuclear Plant

Tennessee Valley Authority

Electronic Mail Distribution

T. A. Hess

Program Manager

Corporate Licensing

Tennessee Valley Authority

Electronic Mail Distribution

Edward J. Vigluicci

Associate General Counsel, Nuclear

Tennessee Valley Authority

Electronic Mail Distribution

Chairman

Limestone County Commission

310 West Washington Street

Athens, AL 35611

State Health Officer

Alabama Dept. of Public Health

P.O. Box 303017

Montgomery, AL 36130-3017

J. Shea 4

Letter to Joseph W. Shea from Eugene Guthrie dated February 8, 2013

SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION

REPORT 05000259/2012005, 05000260/2012005, AND 05000296/2012005

Distribution w/encl:

C. Evans, RII

L. Douglas, RII

L. Regner, NRR

OE Mail

RIDSNRRDIRS

PUBLIC

RidsNrrPMBrownsFerry Resource

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.: 05000259/2012005, 05000260/2012005, 05000296/2012005

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: October 1, 2012, through December 31, 2012

Inspectors: D. Dumbacher, Senior Resident Inspector

C. Stancil, Resident Inspector

P. Niebaum, Resident Inspector

L. Pressley, Resident Inspector

T. Stephen, Resident Inspector

D. Hardage, Resident Inspector

L. Suggs, Senior Construction Project Inspector (4OA2.5)

A. Sengupta, Reactor Inspector (1R08)

J. Laughlin, Emergency Preparedness Inspector (1EP4)

C. Kontz, Senior Project Engineer (4OA5.4)

G. Laska, Senior Operations Engineer (1R11.3)

K. Schaaf, Operations Engineer (1R11.3)

R. Baldwin, Senior Operations Engineer (1R11.4)

Approved by: Eugene F. Guthrie, Chief

Reactor Projects Special Branch

Division of Reactor Projects

Enclosure

SUMMARY

IR 05000259/2012005, 05000260/2012005, 05000296/2012005; 10/01/2012-12/31/2012;

Browns Ferry Nuclear Plant, Units 1, 2 and 3; Event Follow-up and Identification and Resolution

of Problems

The report covered a three month period of inspection by the resident inspectors, three regional

inspectors, and one headquarters inspector. Two licensee-identified violations 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). The NRCs program for overseeing the safe

operation of commercial nuclear power reactors is described in NUREG-1649, Reactor

Oversight Process Revision 4, dated December 2006.

A. NRC Identified and Self-Revealing Findings

None

B. Licensee Identified Violations

Two violations of very low safety significance, which were identified by the licensee,

have been reviewed by the inspectors. Corrective actions taken or planned by the

licensee have been entered into the licensees corrective action program. This violations

and the corrective action program tracking numbers are described in Section 4OA7 of

this report.

Enclosure

REPORT DETAILS

Summary of Plant Status

Unit 1 operated at full Rated Thermal Power (RTP) except for one planned downpower to

support the refueling outage (U1R9). On October 20, 2012 the unit was shutdown for a

scheduled refueling outage that lasted 45 days. The unit was restarted on December 4, 2012

and returned to full power on December 7, 2012. The unit remained at near full power the

remainder of the quarter.

Unit 2 operated at full RTP except for 3 planned downpowers and 1 unplanned SCRAM. On

October 26, 2012, a planned downpower to 70 percent power was made for one day to

complete Control Rod Sequence Exchange and SCRAM Time Testing. On November 23,

2012, a planned downpower to 95 percent power was made for Control Rod Sequence

Exchange. On December 12, 2012, a planned downpower to 45 percent power was performed

to enable maintenance on 2B recirculation pump Variable Frequency Drive, steam leak repairs

to the 73-3 line, and repairs to 2B cond. booster pump. The plant returned to 100 percent

power on December 14, 2012. On December 22, 2012, Unit 2 reactor automatically scrammed

due to a post maintenance test failure associated with 3D emergency diesel and a wrong-train

human performance error, respectively causing a loss of the 2B reactor protection subsystem

and the 2A reactor protection subsystem. Unit 2 was restarted on December 25, 2012, and

synchronized to the electrical grid on December 26, 2012. The unit remained at near full power

the remainder of the quarter.

Unit 3 operated at full RTP power except for 3 planned downpowers. On October 15, 2012, a

planned downpower to 60 percent power for one day to repair a steam leak on the 3A

Feedwater drain line. On November 19, 2012, a planned downpower to 95 percent power for

one day to repair a steam leak on 3C1/3C2 High Pressure Heaters. On December 14, 2012, a

planned downpower to 60 percent power for control rod sequence exchange and turbine control

valve testing. The unit remained at full power the remainder of the quarter.

Enclosure

4

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

.1 Evaluate Readiness to Cope with External Flooding

a. Inspection Scope

The Inspectors reviewed licensee flood protection barriers and procedures for coping

with external flooding. The inspection covered the FSAR and related flood analysis

documents to identify those areas that can be affected by external flooding and seasonal

susceptibilities such as floods caused by hurricanes, heavy rains and flash floods. The

review covered design flood level documentation and corrective actions for safety

related areas. The inspectors conducted a walkdown of the Unit Common intake

structure Residual Heat Removal Service Water (RHRSW) pump rooms. Specific focus

addressed: sealing of equipment below the flood line, such as electrical conduits; sealing

of equipment floor plugs, holes or penetrations in floors and walls between flood areas;

and adequacy of watertight doors between flood areas. This activity constitutes two

External Flood Protection samples.

Common Intake Structure RHRSW pump room hatches and vents as part of

Temporary Instruction (TI) -187, Independent Flooding Walkdowns

Licensee walkdown packages associated with RHRSW pump room and Diesel

Generator building CO2 room watertight doors as part of Temporary Instruction (TI) -

187, Flooding Walkdowns

b. Findings

No findings were identified.

.2 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

Prior to and during the onset of cold weather conditions, the inspectors reviewed the

licensees implementation of 0-GOI-200-1, Freeze Protection Inspection, including

applicable checklists: Attachment 1, Freeze Protection Annual Checklist; Attachment 2,

Freeze Protection Operational Checklist; and as applicable, Attachments 3 through 12,

Freeze Protection Daily Log Sheets for individual watch stations. The inspectors also

reviewed the list of open FZ-coded Work Orders and Problem Evaluation Reports

(PERs) to verify that the licensee was identifying and correcting potential problems

relating to cold weather operations. In addition, the inspectors reviewed procedure

requirements and walked down selected areas of the plant, which included the main

control rooms, Residual Heat Removal Service Water (RHRSW) and Emergency

Equipment Cooling Water (EECW) pump rooms, and all units Emergency Diesel

Generator (EDG) buildings, to verify that affected systems and components were

Enclosure

5

properly configured and protected as specified by the procedure. The inspectors

discussed cold weather conditions with Operations personnel to assess plant equipment

conditions and personnel sensitivity to upcoming cold weather conditions. This

constitutes one Readiness for Seasonal Extreme Weather sample.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial Walkdown

a. Inspection Scope

The inspectors conducted three 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 three Equipment Alignment inspection samples.

Unit 1 Shutdown Cooling System

Unit 1 Auxiliary Decay Heat Removal System

Unit 3 Residual Heat Removal (RHR) System - Division II

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Fire Protection Tours

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

Enclosure

6

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 Fire Protection

inspection samples.

Unit 1 Reactor Building, EL 593 1B Electrical Board Room (Fire Area 4)

Unit 2 Reactor Building, EL 593 2B Electrical Board Room (Fire Area 8)

Unit 3 Reactor Building, EL 519 through 639 (Fire Zone 3-1)

Unit 3 Reactor Building, EL 621 3A Electric Board Room (Fire Area 13)

Unit 3 Reactor Building, EL 621 3A 480 Shutdown Board Room, (Fire Area 14)

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

a. Inspection Scope

Non-Destructive Examination Activities and Welding Activities: From October 29, 2012,

through November 1, 2012, the inspector conducted a review of the implementation of

the licensees In-service Inspection (ISI) Program for monitoring degradation of the

reactor coolant system, emergency feedwater systems, risk-significant piping and

components, and containment systems. The inspectors reviewed the implementation of

the licensees Risk Informed ISI program for monitoring degradation of the reactor

coolant system (RCS) boundary and risk significant piping boundaries. The inspectors

activities consisted of an on-site review of NDE and welding activities to evaluate

compliance with the applicable edition of the ASME Boiler and Pressure Vessel Code

(BPVC),Section XI (Code of record: 2001 Edition through 2003 Addendum) and that

indications and defects (if present) were appropriately evaluated, and dispositioned in

accordance with the requirements of the ASME Code,Section XI acceptance standards

or NRC approved alternative requirement.

The inspectors directly observed or reviewed records of the following NDE mandated by

the ASME Code to evaluate compliance with the ASME Code Section XI and Section V

requirements, and if any indications or defects were detected, to evaluate if they were

dispositioned in accordance with the ASME Code or an NRC-approved alternative

requirement.

Directly observed:

o Work Order # 112507860, Ultrasonic examination (UT) (manual) of Instrument

Nozzle Safe End Welds in Feedwater System

Enclosure

7

Reviewed records:

o Work Order # 114009989, UT of HPCI System, pipe to flange

o Work Order # 1-SI-4.6.G, Visual Examination of RHR System of Weld # 1-

47B452H0158

o Work Order # 1-SI-4.6.g, Magnetic particle Testing (MT) of RHR System of Weld

1-47B452H0158-IA

o Work Order # 114009989, Radiography Examination of HPCI system, Turbine

Steam Supply Valve

During non-destructive surface and volumetric examinations performed since the

previous refuelling outage, the licensee did not identify any recordable indications that

required acceptance for continued service, therefore, no NRC review was required for

this inspection procedure attribute.

The inspectors reviewed documentation for the repair/replacement of the following

pressure boundary welds. The inspectors evaluated if the licensee applied the pre-

service non-destructive examinations and acceptance criteria required by the

construction Code. In addition, the inspectors reviewed the welding procedure

specifications, welder qualifications, welding material certifications, and supporting weld

procedure qualification records to evaluate if the weld procedures were qualified in

accordance with the requirements of the Construction Code and the ASME Code

Section IX.

Welding package for HPCI 2 Turbine Exhaust at Drain (Work Order # 112453386)

Welding package for HPCI System, pipe to flange (Work Order # 114009989)

Identification and Resolution of Problems: The inspectors performed a review of ISI-

related problems, including welding that were identified by the licensee and entered into

the Corrective Action Program (CAP) as Condition Report (CRs). The inspectors

reviewed the CRs to confirm that the licensee had appropriately described the scope of

the problem, description of the evaluation and had identified appropriate corrective

actions. The review also included the review of the licensees use, consideration and

assessment of operating experience events applicable to the plant. The inspectors

performed this review to ensure compliance with 10 CFR Part 50, Appendix B, Criterion

XVI, Corrective Action, requirements. Document reviewed are listed in the Attachment.

b. Findings

No findings were identified.

Enclosure

8

1R11 Licensed Operator Requalification

.1 Resident Inspector Quarterly Review

a. Inspection Scope

On October 17, 2012, the inspectors observed a licensed operator requalification

simulator examination for an operating crew according to a Unit 2 Simulator Exercise

Guide, (SEG), scenario which contained at a minimum the following attributes; main

generator hydrogen leak, loss of offsite power (LOOP), inadvertent pump start, failure to

scram, unisolable reactor core isolation cooling (RCIC) system leak, fuel element failure.

The inspectors specifically evaluated the following attributes related to the operating

crews performance:

Clarity and formality of communication

Ability to take timely action to safely control the unit

Prioritization, interpretation, and verification of alarms

Correct use and implementation of procedures including Abnormal Operating

Instructions (AOIs), and Emergency Operating Instructions (EOIs)

Timely control board operation and manipulation, including high-risk operator actions

Timely oversight and direction provided by the shift supervisor, including ability to

identify and implement appropriate technical specifications actions such as reporting

and emergency plan actions and notifications

Group dynamics involved in crew performance

The inspectors assessed the licensees ability to administer testing and assess the

performance of their licensed operators. The inspectors attended the post-examination

critique performed by the licensee evaluators, and verified that licensee-identified issues

were comparable to issues identified by the inspector. The inspectors also reviewed

simulator physical fidelity (i.e., the degree of similarity between the simulator and the

reference plant control room, such as physical location of panels, equipment,

instruments, controls, labels, and related form and function). This activity constitutes

one Resident Inspector quarterly review of Licensed Operator requalification inspection

sample.

b. Findings

No findings were identified.

Enclosure

9

.2 Control Room Observations

a. Inspection Scope

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. 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.

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.

.3 Biennial Licensed Operator Requalification

a. Inspection Scope

The inspectors reviewed the facility operating history and associated documents in

preparation for this inspection. During the week of October, 8-11, 2012, the inspectors

reviewed documentation, interviewed licensee personnel, and observed the

administration of operating tests associated with the licensees operator requalification

program. Each of the activities performed by the inspectors was done to assess the

effectiveness of the facility licensee in implementing requalification requirements

identified in 10 CFR Part 55, Operators Licenses. The evaluations were also

performed to determine if the licensee effectively implemented operator requalification

guidelines established in NUREG-1021, Operator Licensing Examination Standards for

Power Reactors, and Inspection Procedure 71111.11, Licensed Operator

Requalification Program. The inspectors also evaluated the licensees simulation

facility for adequacy for use in operator licensing examinations using ANSI/ANS-3.5-

1985, American National Standard for Nuclear Power Plant Simulators for use in

Operator Training and Examination. The inspectors observed two shift crews during the

Enclosure

10

performance of the operating tests. Documentation reviewed included written

examinations, Job Performance Measures (JPMs), simulator scenarios, licensee

procedures, on-shift records, simulator modification request records, simulator

performance test records, operator feedback records, licensed operator qualification

records, remediation plans, watchstanding records, and medical records. The records

were inspected using the criteria listed in Inspection Procedure 71111.11. Documents

reviewed are listed in the Attachment.

The inspectors selected PER 245312 Reactivity Management [Control] Plan (RCP)

requires improvement for a detailed review. PER 245312 states: Review of a

completed RCP for the Unit 3 down power and shutdown and other RCPs indicated a

lack of rigor in the execution of some of the RCP steps. Incomplete guidance on a

number of the RCP steps, and some knowledge discrepancies exist. The Quality

Assurance group recommended that additional training be developed for operators and

to improve implementing reactivity management plans. The training department

developed training on reactivity management plans that was delivered to the operations,

group. The reactor engineering group was invited to attend these training sessions to

add additional technical knowledge. Inspectors reviewed the training presentations

developed for the additional training. It appears this training was effective.

b. Findings

No findings were identified.

.4 Annual Review of Licensee Requalification Examination Results

a. Inspection Scope

On December 12, 2012, the licensee completed the annual requalification operating

examinations required to be administered to all licensed operators in accordance with 10

CFR 55.59(a)(2). The inspectors performed an in-office review of the overall pass/fail

results of the individual operating examinations and the crew simulator operating

examinations in accordance with Inspection Procedure (IP) 71111.11, Licensed

Operator Requalification Program. These results were compared to the thresholds

established in Inspection Manual Chapter (IMC) 0609, Significance Determination

Process, Appendix I, Operator Requalification Human Performance Significance

Determination Process.

b. Findings

No findings were identified.

Enclosure

11

1R12 Maintenance Effectiveness

.1 Routine

a. Inspection Scope

The inspectors reviewed the two specific structures, systems and components (SSC)

within the scope of the Maintenance Rule (MR) (10 CFR 50.65) 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 licensees 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 MR expert panel meetings to

verify that regulatory and procedural requirements were met. This activity constituted

two Maintenance Effectiveness inspection samples.

Browns Ferry Valve Stem Packing Program

Control Room Emergency Ventilation System

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 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 licensees risk assessments and adequacy of RMA implementations. This activity

constituted four Maintenance Risk Assessment inspection samples.

Enclosure

12

October 19, 2012, Unit 3 RPS 3A Motor Generator Set Failure with 3E Raw Cooling

Water Pump and G Control and F Service Air Compressors Out of Service (OOS)

October 25, 2012, Units 1 and 2, C Shutdown Board, C EDG and RHRSW Pumps

B2 and B3 OOS, 2B CCW Pump OOS, Unit 1 ORAM Yellow

November 1, 2012, Unit 1 Orange planned outage risk associated with local leak rate

test of valve 74-68 (Operation with Potential to Drain the Reactor Vessel, OPDRV)

December 13, 2012, Unit 2 elevated Green risk for Single Loop Operations

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

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 licensees 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 six Operability

Evaluation inspection samples.

Emergency Equipment Cooling Water South Header Combined Leakage of D

Strainer and B3 Pump (PERs 617833 and 617840)

3C Emergency Diesel Generator Fast Start and Load Sharing Relay Configurations

(PER 617890)

Reactor Building South Access Watertight Door Broken Frame Weld (PER 623264)

High Pressure Coolant Injection (HPCI) Steam Line Inboard Isolation Valve Failure

Due to Inadequate Manufacturers Assembly (PER 639155)

Residual Heat Removal Service Water (RHRSW) Pump Room Sump Debris (PER

618735)

Unit Common Standby Gas Treatment Train C Inoperable longer than allowed by

Technical Specification (PERs 590208 and 604350)

b. Findings

No findings were identified.

Enclosure

13

1R18 Plant Modifications

.1 Permanent Plant Modifications

a. Inspection Scope

The inspectors reviewed the Design Change Notice (DCN) and completed work package

(WO 113086731) for DCN 70488, Replace Unit 1 HPCI 73-16 gate valve with a different

design gate valve, including related documents and procedures. The inspectors

reviewed licensee procedures NPG-SPP-9.3, Plant Modifications and Engineering

Change Control, and NPG-SPP-6.9.3, Post-Modification Testing, and observed part of

the licensee=s activities to implement this design change made while the unit was online.

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.

b. Findings

No findings were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors witnessed and reviewed the four post-maintenance tests (PMT) listed

below to verify that procedures and test activities confirmed SSC operability and

functional capability following the described maintenance. The inspectors reviewed the

licensees 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 four Post Maintenance Test inspection samples.

Unit 1: Residual heat removal service water (RHRSW) Pump B1 Hand switch, (0-

HS-023-0015A/3), replacement per WO 111436112

Unit 1: High Pressure Coolant Injection (HPCI) Turbine Steam Supply Valve, (1-

FCV-073-0016), per WO 113657859, MOVATS Testing per ECI-0-000-MOV9; and

WO 113195490, HPCI Comprehensive surveillance per 1-SR-3.5.1.7(COMP)

Enclosure

14

Unit 3: 3C Emergency Diesel Generator Fast Start and Load Sharing Relay

Configurations (PER 617890) per WO 114009014 and Procedure 3-SR-3.8.1.1(3C),

Diesel Generator 3C Monthly Operability Test

Unit 3: 3A Residual Heat Removal (RHR) pump breaker and cubicle preventive

maintenance per WO 113690981 and 3-SR-3.5.1.6 (RHR-I)

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

.1 Unit 1 Scheduled Refueling Outage (U1R9)

a. Inspection Scope

From October 20, 2012, through December 4, 2012, the inspectors examined critical

outage activities to verify that they were conducted in accordance with Technical

Specifications, applicable plant procedures, and the licensees outage risk assessment

and management plans. The inspectors also monitored critical plant parameters, and

observed operator control of plant conditions, during Cold Shutdown (Mode 4), Startup

(Mode 2), and Power Operation (Mode 1). Some of the significant outage activities

specifically reviewed and/or witnessed by the inspectors were as follows:

Outage Risk Assessment

Prior to the U1R9 refueling outage that began on October 20, the inspectors attended

outage risk assessment team meetings and reviewed the Outage Risk Assessment

Report to verify that the licensee had appropriately considered risk, industry experience,

and previous site-specific problems in developing and implementing an outage plan that

assured defense-in-depth of safety functions were maintained. The inspectors also

reviewed the daily U1R9 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 SPP-7.2, 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.

Enclosure

15

Shutdown and Cooldown Process

The inspectors witnessed the shutdown and cooldown of Unit 1 in accordance with

licensee procedures OPDP-1, Conduct of Operations; 1-GOI-100-12A, Unit Shutdown

from Power Operations to Cold Shutdown and Reduction in Power During Power

Operations; and 1-SR-3.4.9.1(1), Reactor Heatup or Cooldown Rate Monitoring.

Decay Heat Removal

The inspectors reviewed licensee procedures 1-OI-74, Residual Heat Removal System

(RHR); 1-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 walkdowns of system and components to verify

correct system alignment. During planned evolutions that resulted in increased outage

risk conditions 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 licensees 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 orders (i.e., tag order 1-

TO-2012-0003, sections 1-074-0016 and 1-074-0017A for 1D RHR Pump motor and

rotating element replacements

Verified Reactor Coolant System (RCS) inventory controls, specifically the November

1st evolution supporting RHR valve local leak rate testing which had the potential to

drain the reactor vessel (OPDRV), were controlled per 1-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

Verified the licensee was managing fatigue by review of fatigue assessments and

review of certain outage and non-outage workers schedules and work hours

(There were no waiver requests or self declarations.)

Enclosure

16

Reactor Vessel Disassembly and Refueling Activities

The inspectors witnessed selected activities associated with reactor vessel disassembly,

and reactor cavity flood-up and drain down in accordance with 1-GOI-100-3A, Refueling

Operations (Reactor Vessel Disassembly and Floodup). Also, on numerous occasions,

the inspectors witnessed fuel handling operations during the two Unit 1 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.

Torus Closeout

On November 24, 2012, the inspectors reviewed the licensees conduct of 1-GOI-200-2,

Torus Closeout, and performed an independent detailed closeout inspection of the Unit 1

Torus.

Drywell Closeout

On November 27, 2012, the inspectors reviewed the licensees conduct of 1-GOI-200-2,

Drywell Closeout, and performed an independent detailed closeout inspection of the Unit

1 drywell.

Restart Activities

The inspectors specifically conducted the following:

Witnessed heatup and pressurization of Unit 1 reactor pressure vessel in accordance

with 1-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor pressure

Vessel and Associated Piping.

Evaluated licensee actions and response to reactor vessel inner o-ring leakage.

Reviewed Primary Containment Total Leak Rate results

Witnessed Unit 1 approach to criticality and power ascension per 1-GOI-100-1A, Unit

Startup, and 1-GOI-100-12, Power Maneuvering

Reactor Coolant Heatup/Pressurization to Rated Temperature and Pressure per 1-

SR-3.4.9.1, Reactor Heatup and Cooldown Rate Monitoring

Evaluated licensee decision (December 2, 2012,ODMI/PER 651334) to perform plant

startup and operate plant with an existing reactor vessel inner o-ring leak

Enclosure

17

Corrective Action Program

The inspectors reviewed PERs generated during U1R9 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 TS 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 five Surveillance Testing inspection samples, one

in-service, two routine tests, one containment isolation test, and one reactor coolant

system leak detection test.

In-Service Tests:

November 19, 2012, 1-SR-3.1.7.7, Standby Liquid Control System Functional Test

Routine Surveillance Tests:

October 23, 2012, 0-SR-3.8.1.9C, C Emergency Diesel Generator Load Acceptance

Test

November 28, 2012, 1-SR-3.5.1.8, HPCI Main and Booster Pump Set Developed

Head and Flow Rate Test at 150 psig Reactor Pressure

Containment Isolation Valve Tests:

November 1, 2012, Local Leak Rate Test of Unit 1 Residual Heat Removal valves

74-67 and 74-68

Reactor Coolant System Leak Detection Tests:

December 6, 2012, 3-SI-4.2.E-1(B), Drywell Equipment Drain Sump Flow Integrator

Calibration

b. Findings

No findings were identified.

Enclosure

18

Cornerstone: Emergency Preparedness

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The NSIR headquarters staff performed an in-office review of the latest revisions of

various Emergency Plan Implementing Procedures (EPIPs) and the Emergency Plan

located under ADAMS accession numbers ML12296A649, ML12307A285, and

ML12199A022 as listed in the Attachment.

The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in

the revisions resulted in no reduction in the effectiveness of the Plan, and that the

revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to

10 CFR Part 50. The NRC review was not documented in a safety evaluation report and

did not constitute approval of licensee-generated changes; therefore, these revisions are

subject to future inspection. Documents reviewed are listed in the Attachment. This

inspection activity satisfied one inspection sample for the emergency action level and

emergency plan changes on an annual basis.

b. Findings

No findings were identified.

4. OTHER ACTIVITIES

4OA1 Performance Indicator (PI) Verification

.1 Cornerstone: Initiating Events

a. Inspection Scope

The inspectors reviewed the licensees procedures and methods for compiling and

reporting the PIs listed below, including procedure NPG-SPP-02.2. The inspectors

examined the licensees PI data for the specific PIs listed below for the fourth quarter of

2011 through the third quarter of 2012. The inspectors compared the licensees raw

data against graphical representations and specific values reported to the NRC for the

third quarter of 2012 to verify that the data was correctly reflected in the report.

Furthermore, the inspectors validated this data against relevant licensee records (e.g.,

PERs, Daily Operator Logs, Plan of the Day, LERs, 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, to ensure that

industry reporting guidelines were appropriately applied. This activity constituted nine

Performance Indicator Verification inspection samples; three unplanned scrams, three

Unplanned Scrams with Complications, and three Unplanned Power Changes.

Enclosure

19

Unit 1 Unplanned Scrams

Unit 2 Unplanned Scrams

Unit 3 Unplanned Scrams

Unit 1 Unplanned Scrams with Complications

Unit 2 Unplanned Scrams with Complications

Unit 3 Unplanned Scrams with Complications

Unit 1 Unplanned Power Changes

Unit 2 Unplanned Power Changes

Unit 3 Unplanned Power Changes

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

licensees 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

a. Inspection Scope

The inspectors reviewed the specific corrective actions associated with failed GE HFA

Relays on Unit 2 (PER 618132) and Unit 3 (PER 571080).

b. Assessment and Observations

The inspectors had the following observations:

The inspectors reviewed the impact, causal analysis, and corrective actions for five AC

HFA relay failures that resulted in half-scrams from 2003 to 2012. The licensees

apparent cause analysis for the failures concluded the relays were an older style and

had thus failed due to insulation breakdowns. After the inspectors questioned this

conclusion, the licensee determined that the glass enclosure for the relays which

retained heat generated had not been factored into their life expectancy. Additionally,

the inspectors questioned the licensees application of previous guidance related to

replacement of HFA relays. Licensee follow-up identified that additional inspections

were needed to ensure installed HFA relays were of a type capable of operating in the

enclosure device.

Enclosure

20

Browns Ferry conducted walkdowns of approximately 2275 HFA relays on the three

units to determine whether relays required replacement. Forty Four relays had to be

replaced.

c. Findings

No findings were identified.

.3 Semiannual Review to Identify Trends

a. Inspection Scope

As required by Inspection Procedure 71152, the inspectors performed a review of the

licensees 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 July 2012 through December 2012, although some searches expanded beyond

these dates. Additionally, the inspectors review also included the Integrated Trend

Reports (ITR) from the third and fourth quarters of fiscal year 2012. The licensee reports

covered the period of April 1, 2012 to September 30, 2012. Furthermore, the inspectors

verified that adverse or negative trends identified in the licensees PERs, periodic reports

and trending efforts were entered into the CAP. Inspectors interviewed the appropriate

licensee staff and also reviewed procedures, NPG-SPP-02.8, Integrated Trend Review

and NPG-SPP-02.7, PER Trending.

The purpose of the licensees 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, INPO, Nuclear Safety Review Board (NSRB), QA, etc., and determine

why they were not identified by line organizations.

b. Findings and Observations

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

discussed below:

Inspectors noted that licensee-identified third and fourth quarter Corrective Action

Program (CAP) and Human Performance issues continued from the first and second

quarter 2012 trend reports. These issues were also identified as Fleet and Site top

priorities. The majority of the key actions to resolve the gaps for these issues were still

in progress. Trending information provided in the fourth quarter fiscal year 2012 report

showed evidence of improvement in the CAP and Human Performance metrics. The

sites Human Performance error rate metrics were indicating the best of the last two

years.

Enclosure

21

In addition to reviewing the sites progress on the above issues, the inspectors

conducted an independent review of the licensees CAP to identify potential adverse

trends.

.4 Focused Annual Sample Review - Operator Workarounds

a. Inspection Scope

The inspectors conducted a review of existing Operator Workarounds (OWA) to verify

that the licensee was identifying OWAs at an appropriate threshold, entering them into

the corrective action program, establishing adequate compensatory measures,

prioritizing resolution of the problem, and implementing appropriate corrective actions in

a timely manner commensurate with its safety significance. The inspectors examined all

active OWAs listed in the Limiting Condition of Operation Tracking (LCOTR) Log, and

reviewed them against the guidance in BFN-ODM-4.16, Operator Workarounds/

Burdens/Challenges. The inspectors also discussed these OWAs in detail with on shift

operators to assess their familiarity with the degraded conditions and knowledge of

required compensatory actions. Furthermore, the inspector walked down selected

OWAs, and verified the ongoing performance, and/or feasibility of, the required actions.

Lastly, for selected OWAs, the inspector reviewed the applicable PER, including the

associated functional evaluation and corrective action plans (both interim and long term).

b. Findings and Observations

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

which were discussed with the licensee:

Inspectors determined that, in general, Browns Ferry adequately tracks and trends all

operator workarounds, burdens and challenges. This includes estimating, tracking and

compiling the aggregate impact of the workarounds, burdens and challenges.

Inspectors identified multiple occasions where operations staff had routinely updated the

progress of corrective actions within the confines of the software program designed to

track OWAs (eSOMS). The licensee did self-identify that workarounds from equipment

failures have adversely affected the time available for operators to perform their normal

duties.

Inspectors reviewed trending information on workarounds, burdens and challenges

which was reported weekly within the Browns Ferry Plan of the Day. Unit 1 and unit

common systems both had 2 OWAs and were well above the goal for the majority of the

year. Unit 3, however, spent the majority of the year with a high number of operator

burdens. Common unit burdens also finished the year well above the stations goals.

Inspectors noted that workarounds, burdens, and challenges were existing beyond one

operating cycle on a unit. There were examples of multiple burdens and challenges that

were not driven to conclusion prior to plant restart and were allowed to remain

outstanding following refueling outages. Inspectors determined that the site often fails to

adequately address and drive to resolution lower level operator issues when the

opportunity is available to do so.

Enclosure

22

The licensee entered all the above issues into the CAP as SR 665557.

.5 Focused Annual Sample Review - Environmental Qualification (EQ)

a. Inspection Scope

Inspectors conducted a review of the licensees EQ program to verify that the licensee

was identifying and resolving problems associated with EQ equipment at an appropriate

threshold, entering them into the corrective action program, establishing adequate

compensatory measures, prioritizing resolution of the problems, and implementing

appropriate corrective actions in a timely manner commensurate with its safety

significance. During the week of December 17, the inspectors interviewed engineering

personnel and reviewed a sample of the licensees electronic EQ binders, EQ program

assessments, health reports, environmental qualification information releases (EQIRs)

and EQ related problem evaluation reports (PERs) to ensure general EQ program and

10 CFR 50.49 adherence. Inspectors also conducted walkdowns of a sample of

accessible EQ equipment to ensure configuration control was being maintained and that

equipment was installed in accordance with the tested configuration.

b. Findings

A violation of very low safety significance, which was identified by the licensee, has been

reviewed by the inspectors. Corrective actions taken or planned by the licensee have

been entered into the licensees corrective action program. This violation and the

corrective action program tracking number are described in Section 4OA7 of this report.

4OA3 Event Follow-up

.1 (Closed) Licensee Event Reports (LERs) 05000259, 260, and 296/2012-008-00 and -01,

Standby Gas Treatment System Train C Inoperable Longer Than Allowed by Technical

Specifications

a. Inspection Scope

On August 2, 2012, the licensee placed the common Standby Gas Treatment (SGT)

Train C relative humidity heater into service which resulted in an annunciator alarm

indicating power was lost to the SGT Train C filter bank heater element. Licensee

troubleshooting identified that the motor control center (MCC) bucket containing the

associated breaker was misaligned due to a missing retaining device. The licensee

concluded that the SGT Train C had been inoperable since preventive maintenance

performed in September, 2011, as the result of an inadequate maintenance instruction

which allowed installation of a breaker bucket with a single retaining device. The

inspectors reviewed the initial LER issued on October 1, 2012, the LER revision issued

on December 14, 2012, and associated Problem Evaluation Report (PER) 604350,

which included the cause determination and corrective action plans. These licensee

evaluations concluded that the relative humidity heater was not required for the SGT

Train C to perform accident required functions.

Enclosure

23

b. Findings

No findings were identified. This LER is closed.

.2 (Closed) Licensee Event Report (LER) 05000296/2012-004-00, Manual Reactor Scram

During Startup Due to Multiple Control Rod Insertion

a. Inspection Scope

The inspectors reviewed the LER for potential performance deficiencies and/or violations

of regulatory requirements. The LER was associated with the Unit 3 manual reactor scram that occurred during a reactor startup on May 24, 2012. The inspectors reviewed

the root cause report associated with this event and discussed the issue with appropriate

members of plant staff. The cause of the scram was attributed to Unit Operator error

combined with IRM signal spikes associated with manipulation of the scram reset switch

and a degraded IRM High Voltage coaxial cable connector on the 3A IRM. This

condition was documented in the licensees corrective action program as PER 558437.

Additional documents reviewed are listed in the Attachment. This LER is closed.

b. Findings

No findings were identified. This LER is closed

.3 (Closed) Licensee Event Reports (LER) 05000296/2012-006-00; 05000296/2012-006-

01, Main Steam Relief Valves Lift Settings Outside Technical Specification Required

Setpoint

a. Inspection Scope

The inspectors reviewed LER 05000296/2012-006-00 and 05000296/2012-006-01 dated

July 24, 2012, and August 31, 2012, and the applicable PER 558488. On May 25, 2012,

two of thirteen Browns Ferry Nuclear Plant Unit 3 main steam relief valves, during

testing, had mechanically actuated at pressures outside the allowed +/- percent

tolerance per Technical Specification 3.4.3 setpoint. One relief valve lifted high at + 3.98

percent and the other low at negative 3.1 percent. This Technical Specification Limiting

Condition for Operation required 12 of the S/RVs to be capable to mechanically open to

relieve excess pressure when the lift setpoint is exceeded (safety function). The

licensees analysis concluded that the variations in lift setting pressures did not prohibit

the ability of the MSRVs to perform the function to open in order to provide over

pressure protection. Twelve S/RVs were available to relieve excess pressure if the

setpoint had been exceeded. However, contrary to the technical specifications

surveillance requirement, only 11 operable main steam relief valves passed the licensee

lift test procedure. The root cause was determined by the Tennessee Valley Authority to

be that the valve design does not make allowances for corrosion bonding. Browns Ferry

captured the corrective actions in PER 558488.

Enclosure

24

b. Findings

A violation of very low safety significance, which was identified by the licensee, has been

reviewed by the inspectors. Corrective actions taken or planned by the licensee have

been entered into the licensees corrective action program. This violation and the

corrective action program tracking number are described in Section 4OA7 of this report.

This LER is closed.

4OA5 Other Activities

.1 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.

.2 (Discussed) NRC Temporary Instruction (TI) 2515/187, Inspection of Near-Term Task

Force Recommendation 2.3 Flooding Walkdowns

a. Inspection Scope

Inspectors conducted independent walkdowns to verify that the licensee completed the

actions associated with the flood protection feature specified in paragraph 03.02.a.2 of

this TI. Inspectors are performing walkdowns at all sites in response to a letter from the

NRC to licensees, entitled Request for Information Pursuant to Title 10 of the Code of

Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the

Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, dated

March 12, 2012 (ADAMS Accession No. ML12053A340).

Enclosure 4 of the letter requested licensees to perform external flooding walkdowns

using an NRC-endorsed walkdown methodology (ADAMS Accession No.

ML12056A050). Nuclear Energy Industry (NEI) document 12-07 titled, Guidelines for

Performing Verification Walkdowns of Plant Protection Features, (ADAMS Accession

No. ML12173A215) provided the NRC-endorsed methodology for assessing external

flood protection and mitigation capabilities to verify that plant features, credited in the

CLB for protection and mitigation from external flood events, and are available,

functional, and properly maintained.

Enclosure

25

b. Findings

Findings or violations associated with the flooding, if any, will be documented in the 1st

quarter integrated inspection report of 2013.

.3 (Discussed) Temporary Instruction 2515/188 - Inspection of Near-Term Task Force

Recommendation 2.3 Seismic Walkdowns

a. Inspection Scope

The inspectors accompanied the licensee on their seismic walkdowns of Unit 1 HPCI,

Unit 2 D Shutdown Board Room and Unit 3 Core Spray Loop I on August 8, 2012, and

verified that the licensee confirmed that the following seismic features associated with

Unit 1 HPCI, Unit 2 D 4kV shutdown board, 2B 250VDC RMOV board, 2B 480V RMOV

board and Unit 3 A and C Core Spray Pumps were free of potential adverse seismic

conditions:

Anchorage was free of bent, broken, missing or loose hardware

Anchorage was free of corrosion that is more than mild surface oxidation

Anchorage was free of visible cracks in the concrete near the anchors

Anchorage configuration was consistent with plant documentation.

SSCs will not be damaged from impact by nearby equipment or structures.

Overhead equipment, distribution systems, ceiling tiles and lighting, and masonry

block walls are secure and not likely to collapse onto the equipment.

Attached lines have adequate flexibility to avoid damage.

The area appears to be free of potentially adverse seismic interactions that could

cause flooding or spray in the area.

The area appears to be free of potentially adverse seismic interactions that could

cause a fire in the area.

The area appears to be free of potentially adverse seismic interactions associated

with housekeeping practices, storage of portable equipment, and temporary

installations (e.g., scaffolding, lead shielding).

Observations made during the walkdown that could not be determined to be acceptable

were entered into the licensees corrective action program for evaluation.

Additionally, inspectors verified that items that could allow the spent fuel pool to drain

down rapidly were added to the SWEL and these items were walked down by the

licensee. Documentation reviewed are listed in the Attachment.

b. Findings

No findings were identified.

Enclosure

26

.4 Follow-up On Alternative Dispute Resolution Confirmatory Orders (IP 92702)

a. Inspection Scope

During the inspection period the inspectors performed a follow-up review of TVAs

implementation of Confirmatory Order for Office of Investigation Report Nos. 2-2006-025

& 2-2009-003, item number 1. This item is closed.

1. By no later than ninety (90) calendar days after the issuance of this Confirmatory

Order, TVA shall implement a process to review proposed licensee adverse

employment actions at TVAs nuclear plant sites before actions are taken to

determine whether the proposed action comports with employee protection

regulations, and whether the proposed actions could negatively impact the SCWE.

During the inspection period the inspectors performed a follow-up review of TVAs

implementation of Confirmatory Order for Office of Investigation Report Nos. 2-2006-025

& 2-2009-003, item numbers 4, 6, and 10. These items are not closed.

4. Through the end of calendar year 2013 and on approximately a quarterly basis, TVA

shall continue to analyze SCWE trends and develop planned actions, as appropriate

6. Through calendar year 2013, TVA shall conduct Town Hall-type meetings at least

annually at its nuclear power plants and corporate office with TVA and contractor

employees which address topics of interest, including a discussion on TVAs policy

regarding fostering a SCWE.

10. TVAs annual online computer-based training course initiative, which discusses the

components of a nuclear safety culture, what is meant by a SCWE, and the avenues

available to raise concerns, shall be maintained through calendar year 2013.

b. Findings and Observations

No findings were identified.

4OA6 Meetings, Including Exit

.1 Exit Meeting Summary

On January 11, 2013, the resident inspectors presented the quarterly inspection results

to Mr. S. Bono, General Plant Manager, Site Operations, and other members of the

licensees staff, who acknowledged the findings. All proprietary information reviewed by

the inspectors as part of routine inspection activities were properly controlled, and

subsequently returned to the licensee or disposed of appropriately.

4OA7 Licensee-Identified Violations

The following violations of very low safety significance (Green) were identified by the

licensee and are violations of NRC requirements which meet the criteria of the NRC

Enforcement Policy, for being dispositioned as a Non-Cited Violation.

Enclosure

27

The licensee-identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI for

the licensees failure to assure that conditions adverse to quality, such as

deficiencies, and nonconformances are promptly identified and corrected.

Specifically, the licensee failed to take timely corrective actions to address an

extensive backlog of EQ information releases which resulted in not meeting their

environmental qualification program and the 10CFR 50.49 auditability requirements.

Contrary to this requirement, since January of 2010, the licensee failed to take

prompt and appropriate corrective actions to evaluate and correct an extensive

backlog of EQIRs, which resulted in 81 of the licensees 99 required Environmental

Qualification equipment files not being updated to reflect the as-installed

specifications and configuration of EQ equipment. The licensee entered this issue

into their corrective action program as PERs 238931 and 624137. The finding was

determined to be of very low safety significance (Green) using Attachment 4 to IMC 0609, Significance Determination Process, because the incomplete corrective

actions did not result in an actual loss of safety function.

Unit 3 Technical Specification 3.4.3, Safety/Relief Valves, required that twelve of

thirteen main steam safety relief valves (MSRVs) lift at a setpoint within plus or

minus three percent of a specified value. Contrary to this, during TS required

surveillance testing following the Unit 3 Cycle 9 refueling outage, the licensee

discovered that the lift setpoints of two MSRVs exceeded the plus or minus three

percent TS allowed pressure band. This TS violation was entered into the licensees

CAP as PER 558488. The finding was determined to be of very low safety

significance because the as-found lift setpoint conditions of the Unit 3 MSRVs were

evaluated and determined to meet the design basis criteria for the most limiting

reactor pressure vessel over-pressurization events.

Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

S. Bono, General Manager Site Ops

J. Boyer, Assistant Director of Engineering

E. Cobey, Licensing

J. Davenport, Licensing

G. Dudley, Site Welding/Repair & Replacement

M. Ellet, Maintenance Rule Coordinator

J. Emens, Nuclear Site Licensing Manager

F. Froscello, ISI Program

W. Hayes, Reactor Engineering Manager

M. Henderson, Vessel Internals Program

H. Higgins, LOR Supervisor (Acting)

L. Hughes, Operations Manger

M. Hunter, Mechanical Maintenance Manager

D. Kettering, Electrical Systems Engineering Manager

T. McCaney, Operations

B. McCreary, Senior Program Manager, Employee Concerns

J. McCormack, Ventilation Systems Engineer

F. Nielson, IWE/IWL Programs

M. Oliver, Site Licensing

K. Polson, Site Vice President

M. Rasmussen, W.C. Manager

T. Scott, PI Manager

R. Stowe, Equipment Reliability Manager

J. Shea, Vice President Nuclear Licensing

P. Summers, DSL

C. Vaughn, Operations Training Manager

M. Webb, Site Licensing

M. Wilson, Site Training Direct

Attachment

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Closed

05000259, 260, and 296/2012-008-00 LER Standby Gas Treatment System Train C

Inoperable Longer Than Allowed by

Technical Specifications (Section 4OA3.1)

05000259, 260, and 296/2012-008-01 LER Standby Gas Treatment System Train C

Inoperable Longer Than Allowed by

Technical Specifications (Section 4OA3.1)

05000296/2012-004-00 LER Manual Reactor Scram During Startup Due

to Multiple Control Rod Insertion (Section

4OA3.2)

05000296/2012-006-00 LER Main Steam Relief Valves Lift Settings

Outside Technical Specification Required

Setpoint (Section 4OA3.3)

05000296/2012-006-01 LER Main Steam Relief Valves Lift Settings

Outside Technical Specification Required

Setpoint (Section 4OA3.3)

05000259, 260, 296- 00 ORD 12/29/2009 Confirmatory Order Action 1

(Section 4OA5.4)

Discussed

2515/187 TI Inspection of Near-Term Task Force

Recommendation 2.3 Flooding Walkdowns

(Section 4OA5.2)

2515/188 TI Inspection of Near-Term Task Force

Recommendation 2.3 Seismic Walkdowns

(Section 4OA5.3)

05000259, 260, 296- 00 ORD 12/29/2009 Confirmatory Order Action 4

(Section 4OA5.4)

05000259, 260, 296- 00 ORD 12/29/2009 Confirmatory Order Action 6

(Section 4OA5.4)

05000259, 260, 296- 00 ORD 12/29/2009 Confirmatory Order Action 10

(Section 4OA5.4)

Attachment

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection - Severe Weather Readiness and External

Flooding

NPG-SPP-10.14, Freeze Protection, Rev. 0

47W225-16, Diesel Generator Building Units 1-3, Environmental Data El 583.5, Rev. 4

47W225-17, Diesel Generator Building Units 1&2, Environmental Data El 565.5, Rev. 4

47W225-18, Diesel Generator Building Unit 3, Environmental Data El 565.5, Rev. 4

47W225-19, Diesel Generator Building Unit 3, Environmental Data El 583.5, Rev. 4

PCR 12003545, Remove Abandoned Equipment from Freeze Protection GOI

PER 661731, Freeze Protection GOI Refers to Breaker With No Landed Field Wiring

PER 661742, Freeze Protection GOI Attachment 1 Documented Removed Piping

PER 661745, Freeze Protection GOI Attachment 1 References Abandoned Equipment

PER 661747, Operator Incorrectly Initialed D EDG room space heater

CTP-FWD-100, Flood Protection Walkdowns NEI 12-07, Rev. 0

NEI 12-07, Guidelines for Performing Verification Walkdowns of Plant Flood Protection

Features, Rev. 0-A

WO 113618794, Perform Flood Protection Walkdowns IAW CTP-FWD-100

DWG 31N203, Concrete Pumping Station Outline - Sheet 1, Rev. 8

0-AOI-100-3, Flood Above Elevation 558, Rev. 35

PER 637130, Flood Walkdowns - Preventive Maintenance Hatches and Manways

Section 1R04: Equipment Alignment

PIP 95-71 Reactor Level and Pressure Instrumentation

1-47E811-1 Flow Diagram Residual Heat Removal System

0-OI-72 Auxiliary Decay Heat Removal (ADHR) System Operations

0-OI-72/ATT-1ADHR System Valve Lineup Checklist

0-OI-72/ATT-2 ADHR System Panel Lineup Checklist

0-OI-72/ATT-3 ADHR System Electrical Lineup Checklist

0-OI-72/ATT-4 ADHR Instrument Inspection Checklist

0-15E900-1 Electrical Instrument Details

0-47E610-72-1 Control Diagram ADHR System Sheet 1

0-47E610-72-2 Control Diagram ADHR System Sheet 2

0-47E873-1 Flow Diagram ADHR Sheet 1

0-47E873-2 Flow Diagram ADHR Sheet 2

3-47E811-1, Flow Diagram Residual Heat Removal System, Rev. 67

3-OI-74/ATT-1, Valve Lineup Checklist Unit 3, Rev. 87

3-OI-74/ATT-2, Panel Lineup Checklist Unit 3, Rev. 87

3-OI-74/ATT-3, Electrical Lineup Checklist Unit 3, Rev. 88

SRs: 652649, 652431

Section 1R05: Fire Protection

Fire Protection Report, Volume 1, Fire Protection Plan, Units1/2/3, Rev. 14

Fire Protection Report, Volume 2, Sections IV, Pre-Plan No. RX3-519 Torus Area and HPCI

Room, Rev. 48

Fire Protection Report, Volume 2, Sections IV, Pre-Plan No. RX3-519NW, Rev. 48

Fire Protection Report, Volume 2, Sections IV, Pre-Plan No. RX3-519SW, Rev. 48

Fire Protection Report, Volume 2, Sections IV, Pre-Plan No. RX3-565, Rev. 48

Attachment

4

Fire Protection Report, Volume 2, Sections IV, Pre-Plan No. RX3-593, Rev. 48

Fire Protection Report, Volume 2, Sections IV, Pre-Plan No. RX3-621, Rev. 48

Fire Protection Report, Volume 2, Sections IV, Pre-Plan No. RX3-639, Rev. 48

Fire Protection Report, Volume 1, Fire Hazards Analysis, Units 1/2/3, Rev. 14

Section 1R08: Inservice Inspection Activities (71111.08G)

Procedures

54-ISI-363-007(AREVA), Remote Underwater In-Vessel Visual Inspection of Reactor Pressure

Vessel Internals, Components, and Associated Repairs in Boiling Water Reactors, Rev. 7

MMDP-10, Controlling Welding, Brazing, And Soldering Processes, Rev. 11

MMDP-8, Controlling Welding, Brazing, And Soldering (WBS) Materials, Rev. 4

MMDP-9, Qualification, Certifications of Personnel Performing Welding Processes, Rev. 6

N-MT-6, TC 11-09, Administration of NDE Procedures for Magnetic Particle Examination, Rev. 6

NPG-SPP-03.1, Corrective Action Program, Rev. 5

NPG-SPP-03.1.4, Corrective Action Program Screening and Oversight, Rev. 9

NPG-SPP-03.1.7, PER Analysis, Actions, Closures and Approvals, Rev. 8

NPG-SPP-09.7, Corrosion Control Program, Rev. 2

N-RT-1, Radiographic Examination of Nuclear Power Plant Components, Rev. 28

N-UT-4, Generic Procedure for the Ultrasonic Examination of Austenitic Pipe Welds, Rev. 11

N-UT-76, Generic Procedure for the Ultrasonic Examination of Ferritic Pipe Welds, Rev. 7

N-VT-1, Visual Examination Procedure for ASME Section XI Preservice and Inservice, Rev. 44

O-TI-140, Monitoring Program for Flow Accelerated Corrosion, Rev. 4

O-TI-365, Unit 1 Reactor Pressure Vessel Internals Inspection (RPVII), Rev. 3

Corrective Action Documents

PER 541322

PER 536424

PER 562226

PER 533410

PER 526313

PER 631474

PER 465158

PER 295682

PER 461544

PER 278797

PER 545569

PER 533699

SR 638639

SR 642831

Other

001, Indication Notification Report of Steam Dryer, Rev. 0

1-SI-4.6G, Inservice Inspection and Risk-Informed Inservice Inspection Program Unit 1, Rev. 26

Areva Certificate of Personnel Qualification (EVT-1/BWRVIP) (Brown), ID# B7717

Areva Certificate of Personnel Qualification (EVT-1/BWRVIP) (Telschow), ID# T5817

Areva Certificate of Personnel Qualification (Vision) (Telschow), ID# T5817

Densitometer, Serial Number 027605

Dwg#1-47B452H0158, Mechanical RHR System Pipe Support, Rev. 0

Attachment

5

Dwg#1-47E812-1-ISI, ASME Section XI HPCI- Code Class Boundary, Rev. 11

Dwg#1-FAC-001-036 (CSI), Unit 1 FAC Location Sketch, Main Steam Lines from Manifold HDR

to the HP Turbine and Bypass Valve Loop, Rev. 0

Dwg#1-FAC-006-044 (CSI), Unit 1 FAC Location Sketch, Misc. 8 Drain Header A, B, & C to

Condenser 1A, 1B, & 1C, Rev. 0

Dwg#1-FAC-006-052 (CSI), Unit 1 FAC Location Sketch, OPER Vent Lines from Feedwater

Heater A4/B4 & C4 to Condensers A/B & C,, Rev. 0

Dwg#1-ISI-0091-C, HPCI Weld Locations, Rev. 0

Dwg#1-ISI-0363-C, RHR Shutdown Support Locations, Rev. 0

Dwg#HPCI-1-018-4, HPCI System Weld, Rev. 67

PQR GT-11-0-1

PQR GT-11-SPEC-1

Source Certificate for Ir192, Holder# 79637B

Structural Integrity Certificate of Personnel Qualification (UT) (May), ID# 1908

Structural Integrity Certificate of Personnel Qualification (Vision) (May), ID# 1908

Structural Integrity Certificate of Personnel Qualification (Visual) (May), ID# 1908

TVA Certificate of Personnel Qualification (MT) ((Priestley), ID# 1UPWAOJ7H

TVA Certificate of Personnel Qualification (RT) (Fox), ID# TDM143XY3

TVA Certificate of Personnel Qualification (RT) (Melford Sr.), ID# PCJ49PAS1

TVA Certificate of Personnel Qualification (UT) (Case), ID# 9XUIL0MVC

TVA Certificate of Personnel Qualification (UT) (Welch), ID# RGV1VT3

TVA Certificate of Personnel Qualification (Vision) (Case), ID# 9XUIL0MVC

TVA Certificate of Personnel Qualification (Vision) (Fox), ID# TDM143XY3

TVA Certificate of Personnel Qualification (Vision) (Ledford), ID# 905506

TVA Certificate of Personnel Qualification (Vision) (Melford Sr.), ID# PCJ49PAS1

TVA Certificate of Personnel Qualification (Vision) (Priestley), ID# RGV1VT3

TVA Certificate of Personnel Qualification (Vision) (Welch), ID# RGV1VT3

TVA Certificate of Personnel Qualification (VT) (Case), ID# 9XUIL0MVC

TVA Certificate of Personnel Qualification (VT) (Priestley), ID# 1UPWAOJ7H

TVA Certificate of Personnel Qualification (VT) (Welch), ID# RGV1VT3

TVA Certificate of Personnel Qualification (Welding) (Dwens), ID# RX4MTJMA0

TVA Certificate of Personnel Qualification (Welding) (Garne), ID# 89GUHH9BE

TVA Certificate of Personnel Qualification (Welding) (McCrelen), ID# 3L6CVC05B

TVA Certificate of Personnel Qualification (Welding) (Pierce), ID# FW1CJ5V48

TVA Certificate of Personnel Qualification (Welding) (Potts), ID# 27E0204B0

TVA Certificate of Personnel Qualification (Welding) (Potts), ID# HCQLUFOQD

TVA Certificate of Personnel Qualification (Welding) (Terry), ID# BFK8QFZ8U

TVA Certificate of Personnel Qualification (Welding) (Tomkins), ID# Z09MA8T2D

TVA Certificate of Personnel Qualification (Welding) (Tucker), ID# MWM1KT9SP

TVA Certificate of Personnel Qualification (Welding) (Whitley), ID# HEF006ZH1

URS Certificate of Personnel Qualification (UT) (Butler), ID# 23863

URS Certificate of Personnel Qualification (UT) (Fish), ID# 11401

URS Certificate of Personnel Qualification (UT) (Fish), ID# 61771

URS Certificate of Personnel Qualification (Vision) (Butler), ID# 23863

URS Certificate of Personnel Qualification (Vision) (Fish), ID# 11401

URS Certificate of Personnel Qualification (Vision) (Fish), ID# 61771

URS Certificate of Personnel Qualification (VT) (Butler), ID# 23863

URS Certificate of Personnel Qualification (VT) (Fish), ID# 11401

Attachment

6

URS Certificate of Personnel Qualification (VT) (Fish), ID# 61771

WPS, DWPS GT-11-0-1-N, Rev. 2

Section 1R11: Licensed Operator Requalification

Simulator Exercise Guide, (SEG), Rev. 2

Records:

License Reactivation Packages (2 Records Reviewed)

LORP Training Attendance records

Medical Files (16 Records Reviewed)

Remedial Training Records (Various)

Remedial Training Examinations (2 Records Reviewed)

Various condition reports over the last two years related to licensed operator on shift

performance

Various closed condition reports that were simulator related

Written Examinations:

2011 RO week 1

2011 SRO week 1

2011 RO week 3

Annual Examination Scenarios:

LOR-EXAM- 26, REV.3

LOR-EXAM- 27, REV.3

LOR-EXAM- 51A, REV. 3

LOR-EXAM- 19, REV. 3

LOR-EXAM- 41, REV. 3

LOR-EXAM- 50a, REV. 3

LOR Practice scenarios:

OPL177.060 Rev 9

OPL177.084 Rev 4

OPL177.093 Rev 1

JPMs:

JPM-70ap- Secure Drywell Sprays

JPM 177TC- Secondary Containment Radiation Alert

JPM 204 U2 -Secure System II from suppression pool cooling

JPM 231ap (U1) -Inhibit ADS

JPM 222 r2- Perform Control Room Transfer of 4KV Unit Board 2B Power Supplies

JPM 234 -Operator 4 Manual Actions 0-SSI-21

JPM254-1-EOI Appendix-7C

JPM 263ap-Spreading Room Smoke Removal

JPM 265ap-Unit 2 Recirc Pump Recovery with manual scram

JPM 266-USST 1B Transformer Tap Changer Auto Checks

Attachment

7

Procedures:

NPG-SPP-17.4.1 Exam Security and Exam Database Management Rev. 05, (07-31-2012)

NPG-SPP-17.8.1 Licensed Operator Requalification Examination Development and

Implementation, Rev. 07, (05-31-2012)

NPG-SPP-17.8.2 Job Performance Measures Development, Administration, and

Evaluation, Rev. 02, (04-04-2012)

NPG-SPP-17.8.3 Simulator Exercise Guide Development and Revision, Rev. 02, (03-30-2012)

NPG-SPP-17.8.4 Conduct of Simulator Operations, Rev. 0, (12-27-2011)

TRN-12 Simulator Regulatory Requirements, Rev. 11, (11-02-2011)

Simulator Static and Normal Tests:

100% Steady State Test, Revision 11

82% Steady State Test, Revision 11

46% Steady State Test, Revision 11

Unit 3 Simulator Normal Testing of GOIs Revision 11

Simulator Transient Tests:

Transient Test #1, Manual Scram, Revision 11

Transient Test # 4, Simultaneous Trip of All Recirculation Pumps, Revision 11

Transient Test # 6, Turbine Trip < 30% Power, Revision 11

Transient Test # 7, Maximum Rate Power Ramp, Revision 11

Simulator Malfunction Tests:

RP06-Auto Scram Channel Failure, Revision 11

TC08-Control Valve Position Unit Failure, Revision 11

TC10-EHC Pressure Transducer Failure, Revision 11

ED27-Loss of Power to an ECCS 250V RMOV Board Breaker Failure, Revision 11

PERs:

PER 595296 Operation Missed Technical Specification Call

PER 566196 Develop Case Study for Drain Down Event

PER 558521 Shift Manning in the U3Control Room Inadequate for Start up

PER 245312 Reactivity Management Plan (RCP) requires improvement

Standards:

ANSI/ANS-3.5-1985, American National Standard Nuclear Power Plant Simulators for Use

In Operator Training and Examination

ANSI/ANS-3.4-1983, Medical Certification and Monitoring of Personnel Requiring Operator

Licenses for Nuclear Power Plants

Other Documents:

Self Assessment BFN-TRN-12-010 Operations Training Department IP71111.11B Inspection

Preparations, (July 7-September 10, 2012)

Snapshot self-Assessment Report BFN-OPS-S-004 February 10-14, 2012

Reviewed three LERs for Unit 3 and 1 for Unit 1

Attachment

8

Section 1R12: Maintenance Effectiveness

0-SR-3.7.3.4, Control Bay Habitability Zone Pressurization Test

Air Conditioning (a)(1) plan

Crevs (a)(1) plan

MCI-0-000-PCK001, Generic Maintenance Instructions for Valve Packing, Rev.29

NETP-117, Valve Stem Packing Enhancement Program, Rev. 0

NPG-SPP-03.4, Maintenance Rule Performance Indicator Monitoring, Trending and

Reporting10CFR50.65

PER 236646, A Fleet Valve Stem Packing Program Should be Implemented

PER 244202, BFN-3-VTV-10-502 Blown Packing

PER 252382, MR (a)(1) Plan due to Trend in Plant Shutdown Events Induced by Valve Packing

HU Events

PER 329005, CREVS in (a)(1) status

PER 423569, System 31 Maintenance Rule Performance Criteria Exceeded

PER 473637, 1-FCV-68-79 Drywell Packing Leak

PER 533052, MSIV LLRT Failure due to Valve Packing Blowby

PER 565652, System 31 (a)(1) Plan

PER 567503, (a)(1) Plan Interim Performance Criteria Exceeded

PER 614107, Evaluate need for additional System 031 PMs

PER 652791, Excess Tripping of AC Units

TVA NPG Quick Human Error Analysis Tool, PER 244202, 8/12/2010

TVA Nuclear Power Group BFN Engineering Support Morning Status, Degraded

Conditions/Non-Conforming

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

BFN Plan of the Day, 10/25-26/2012

U1R9 Risk Plan, Inventory Control During An OPDRV Activity (TI-106)

Operator Aid (Drawing) for Unit 1 RHR tie to 1B Recirculation Pump Discharge piping

NRC Staff Position on Dispositioning Boiling-Water Reactor Licensee Noncompliance with

Technical Specification Requirements During Operations with a Potential for Draining the

Reactor Vessel

Enforcement Guidance Memorandum (EGM) 11-003, Enforcement Guidance Memorandum on

Dispositioning Boiling-Water Reactor Licensee Noncompliance with Technical Specification

Containment Requirements during Operations with a Potential for Draining the Reactor Vessel

Procedure 2-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity,

Rev 14

NPG-SPP-09.11.2, Equipment Out of Service (EOOS) Management, Rev. 5

NPG-SPP-07.2.11, Shutdown Risk Management, Rev. 2, dated 10/4/2011

ORAM Outage Safety Assessment for November 1, 2012 Orange risk

NRC Regulatory Issues Summary 2012-11,

NPG Daily Outage Report, U1R9, 10/25-26/2012

WO 114056943

2-SR-3.4.2.1, Jet Pump Mismatch and Operability, Rev. 34

2-SR-3.4.1(SLO) Reactor Recirculation System Single Loop Operation, Rev. 09

U2 RCP 121130-000, Reactivity Maneuver Plan U2 Single Loop Operation (SLO)

Attachment

9

EOOS Operators Risk Worksheet, 12/13/2012

NPG-SPP-09.11.1, Equipment Out of Service (EOOS) Management, Rev. 05

Section 1R15: Operability Evaluations

0-AOI-100-3, Flood Above Elevation 558, Rev. 35

0-OI-67, Emergency Equipment Cooling Water System, Rev. 96

Calculation MDQ0067910008, Flow Requirements of EECW Fed Components, Rev. 16

Calculation MDQ0023870149, RHRSW Pump Compartment Sump and Sump Capacity,

Rev. 10

Design Criteria BFN-50-7067, Emergency Equipment Cooling Water System

FSAR Section 1.2.72, Probable Maximum Flood, BFN-21

FSAR Section 2.4, Hydrology, Water Quality, and Aquatic Biology, BFN-19

FSAR Section 10.10, Emergency Equipment Cooling Water (EECW) System, BFN-22

FSAR Section 12.2.7.1.2, Principle Structures and Foundations, Personnel Access Doors,

BFN-24

FSAR Appendix 2.4A, Browns Ferry Nuclear Plant Maximum Possible Flood, BFN-24

PER 617833, B3 RHRSW/EECW Pump Has Shaft Seal Leak

PER 617840, D EECW South Header Strainer Leaking

SR 622301, Potential Non-Conservative Assumptions in Calculation for Leakage into RHRSW

Pump Rooms

Technical Specification and Basis 3.7.2 Emergency Equipment Cooling Water (EECW) System

and Ultimate Heat Sink (UHS), Amendment 235 and Rev. 0 respectively

0-TI-403, Appendix A, Determination of Common Cause Failure for Emergency Diesel

Generators, Rev. 0, dated 10/04/12

Design Criteria BFN-50-7082, Standby Diesel Generator

Drawing 3-45E767-5, Wiring Diagram Diesel Generators Schematic Diagram, Rev. 26

Engine Systems Inc. Bill of Material, TVA-Browns Ferry/EDG Governor Upgrade, ESI IWO

8001206 & 8002031, Rev. 5

FSAR Section 8.5, Standby AC Power Supply and Distribution, BFN-24

OPDP-1, Conduct of Operations, Rev. 24

PER 617890, 3EB EDG Relay Failed During Dynamic Testing

PER 619824, Reconfigured Relays with Incorrect Part Numbers Installed in 3EB Governor

Upgrade Modification

PER 619972, Failure to Document Critical Thinking per OPDP-1

PER 621030, WO Needed to Investigate and Resolve Configuration of Relays Installed in 3EC

EDG Governor

PER 621079, Timeliness of PDO for 3C Diesel Generator

SR 628201, PER 619824 Requires Re-Screening

SR 628216, Improper Classification of PER 617890

SR 628627, Repeat Failure of a CC2 Component

Technical Specifications Task Force (TSTF) - 531, Revision of Specification 3.8.1, Required

Actions B.3.1 and B.3.2, Rev. 0

Unit 3 Technical Specification and Basis 3.8.1 AC Sources - Operating, Amendment 266

WO 114009014, DG 3C Resolution of Relay Configurations

Drawing 0-46W401-10, Architectural Plans EL 519 and 565, Rev. 0

Drawing 0-34N303, Watertight Personnel Access Doors, Rev. 0

Drawing 1-47E852-1, Flow Diagram Floor & Dirty Radwaste Drainage, Rev. 26

PER 623264, Broken Weld on Door Frame BFN-0-DOOR-260-230A

Attachment

10

WO 07-717762-00, Repair Door Seal

WO 114020199, Repair Broken Weld on Door Frame BFN-0-DOOR-260-230A

Design Criteria BFN-50-7073, High Pressure Coolant Injection System, Rev. 22

Flowserve Design, Seismic, and Weak-Link Analysis, RAL-2634, Size 10 Class 900 Carbon

Steel Double Disc Gate Valve, Rev. 2

Flowserve Drawing , Anchor/Darling, BW/IP, Durco and Valtek Valves, 10 - 900 lb. Double

Disc Gate Valve Weld Ends, Carbon Steel, Body Drain Pipe with Cap, Smart Stem and

AdvanSeal with Limitorque SMB-2-80 Actuator, Rev. B

Flowserve Instruction Manual FCD ADENIM0003-00, Anchor Darling Double-Disc Gate Valves

FSAR Section 6.4.1, High Pressure Coolant Injection System, BFN-24

FSAR Section 7.4.3.2, High Pressure Coolant Injection System (HPCI) Control and

Instrumentation, BFN-24

PER 627529, 1-FCV-73-2 As-Found Leak Rate was 600 scfh with an Admin Limit of 30 scfh

PER 639155, Internal Damage Found in 1-FCV-73-2

PER 638761, Internal Inspection of 1-FCV-73-3

SR 642135, Perform Inspection on 2-FCV-73-2 Wedge Pin During U2R17 Outage

Technical Specification and Basis 3.5.1 EECS - Operating, Amendment 269 and Rev. 53

respectively

Vendor Technical Document BFN-A391-0340, Instruction Manual for Anchor Darling 10 - 900

Double Disc Gate Valve, Rev. 5

WO 02-011512-001, Remove/Replace 1-FCV-73-2 with New Design

WO 08-711832-000, 1-FCV-73-3 Repaired Due to Wrong Pinion/Worm Gear

WO 110937504, 1-FCV-73-16 Repaired Due to Pressure Locking

WO 00-003350-000, 2-FCV-73-2 Pin Sheared During MOVATS

WO 112075251, 2-FCV-73-3 Stem Replaced

WO 110811072, 2-FCV-73-16 Reworked due to Seat Leakage

WO 112330027, 3-FCV-73-16 Replaced due to Seat Leakage

0-OI-23, Residual Heat Removal Service Water System, Rev. 92

Calculation MDQ0023890078, Pump Performance Analysis for New RHRSW Compartment

Sump Pumps, Rev. 3

Design Criteria BFN-50-7023, Residual Heat Removal Service Water System

PER 618735, Removed Scrap Wire From RHRSW Pump Room Sumps

PER 623106, Potential Non-Conservative Assumptions in Calculation for Leakage into RHRSW

Pump Room Sump

SR 622301, Potential Non-Conservative Assumptions in Calculation for Leakage into RHRSW

Pump Rooms

FSAR SBGT

R14 981211 106 TVA Alternate Source Term Calculation Input Parameters

R 92 960718 850 TVA Alternate Source Term Calculations

PER 590208 Past Operability Determination for SBGT C relative humidity heater inoperable

PER 604350 Breaker for SBGT C relative humidity heater inoperable

LER 50-259/2012-008 SBGT Train C inoperable longer than Technical Specification allowable

time

WO 113759132 Standby Gas Treatment Train C filter heating element lost power

Attachment

11

Section 1R18: Plant Modifications

NPG-SPP-9.3, Plant Modifications and Engineering Change Control, Rev. 9

NPG-SPP-6.9.3, Post-Modification Testing, Rev. 3

DCN 70488, Replace Flowserve Gate Valve with Crane-Kalsi Sentinel Gate Valve, Rev. B

WO 113086731, Implement DCN 70488 to Replace BFN-1-FCV-0016

WO 113657859, Perform MOVATS to Support WO 113086731 for Valve Replacement

WO 114009989, Perform Pre-fab Welding on New HPCI Valve

UFSAR, Section 6.4.1 High Pressure Coolant Injection System, BFN-24

DWG DCA No. 70488-119, CD05897 RB, Crane Bolted Bonnet Gate Valve, Rev. Orig.

MCI-0-000-GTV001, Generic Maintenance Instructions for Gate Valves, Rev. 28

MCI-0-000-PCK001, Generic Maintenance Instructions for Valve Packing, Rev. 29

Section 1R19: Post-Maintenance Testing

WO 111436112 - Replace 0-HS-023-0015A

WO 113195620 - Abandon RHRSW pp B2 local controls

0-TI-403, Appendix A, Determination of Common Cause Failure for Emergency Diesel

Generators, Rev. 0, dated 10/04/12

DCN 69532 Stage 7 Continuity Checks

Design Criteria BFN-50-7082, Standby Diesel Generator

Drawing 3-45E767-5, Wiring Diagram Diesel Generators Schematic Diagram, Rev. 26

Engine Systems Inc. Bill of Material, TVA-Browns Ferry/EDG Governor Upgrade, ESI IWO

8001206 & 8002031, Rev. 5

FSAR Section 8.5, Standby AC Power Supply and Distribution, BFN-24

PER 617890, 3EB EDG Relay Failed During Dynamic Testing

PER 619824, Reconfigured Relays with Incorrect Part Numbers Installed in 3EB Governor

Upgrade Modification

PER 621030, WO Needed to Investigate and Resolve Configuration of Relays Installed in 3EC

EDG Governor

PER 621079, Timeliness of PDO for 3C Diesel Generator

PER 629871, Evaluation of Standby diesel Generator 3B and 3C Relay Issues following

Implementation of DCN 69532 in Response to NRC Concerns

PMTI-69532-STG007, 3C Emergency Diesel Generator Governor Control Upgrade, Rev. 2

PER 629081, 3C Diesel PMT Not Ready to Work Due to Inadequate Planned PMT

PER 642810, During Performance of PMT IAW WO 114009014, 3C Start Relay BFN-3-RLY-82-

3C2

Technical Specifications Task Force (TSTF) - 531, Revision of Specification 3.8.1, Required

Actions B.3.1 and B.3.2, Rev. 0

Unit 3 Technical Specification and Basis 3.8.1 AC Sources - Operating, Amendment 266

WO 111549226, Attachment 1, Diesel Generator 2301A and EGB-13P Governor Setup &

Tuning Instruction

WO 114009014, DG 3C Resolution of Relay Configurations

WO 114114698, 3-SR-3.8.1.1(3C) Diesel Generator 3C Monthly Operability Test

1-GOI-100-12A, Unit Shutdown from Power Operations to Cold Shutdown and Reduction in

Power During Power Operations, Rev. 19

1-SR-3.4.9.1(1), Reactor Heatup or Cooldown Rate Monitoring, Rev. 8

Drawing 1-47E811-1, Flow Diagram Residual Heat Removal System, Rev. 37

OPDP-1, Conduct of Operations, Rev. 24

SR 633312, Unit 1 Suppression Pool Water Level Has Dropped 2.2 Between 10/26 and 10/30

Attachment

12

SR 634421, NRC Has Identified the RHR Flow Drawing Does Not Show Two Isolation Valves

WO 113622372, 1-SR-3.4.9.1(1)

3-SR-3.5.1.6 (RHR I), Quarterly RHR System Rated Flow Test Loop I, Rev. 42

3-SR-3.5.1.6 (RHR I-COMP), RHR Loop I Comprehensive Pump Test, Rev. 06

SR 653678, 3C RHR pump discharge pressure indicator

EPI-0-000-BKR015, 4KV Wyle/Siemens Horizontal Vacuum Circuit Breaker (Type-3AF) and

Compartment Maintenance

1-SR-3.5.1.8, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at 150

psig Reactor Pressure, Rev. 11

WO 113195499, 1-SR-3.5.1.8

WO 113657859, Perform NOVATS to support WO 113086731 for Valve Replacement

WO 113657859, Attachment to WO Task 10, Rev. 3

ECI-0-000-MOV009, Testing of Motor Operated Valves Using MOVATS Universal Diagnostic

System (UDS) and Viper 20, Rev. 28

UFSAR, Section 6.4.1 High Pressure Coolant Injection System, BFN-24

1-SR-3.5.1.7(COMP), HPCI Comprehensive Pump Test, Rev. 21

WO 113195490, HPCI Comprehensive Pump Test

MSI-1-073-GOV001, High Pressure Coolant Injection (HPCI) Turbine Overspeed Trip Test,

Rev. 9

0-TI-383, Evaluation of Test Results for the ASME OM Code Inservice Testing Program, Rev. 1

ASME OM Code IST Test Results Evaluation UNID: 12-1-IST-073-473, dated 11/27/2012, SR

648090

PER 648108, 73-16 Stroke Times

1-SR-3.3.3.1.4(G), Verification of Remote Position Indicators for HPCI System Valves, Rev. 2

WO 113183659, 1-SR-3.3.3.1.4(G)

PER 649588, Drain Valve Double Lit

WO 114164455, Drain Valve Double Lit

1-SR-3.6.1.3.5(HPCI), HPCI System Motor Operated Valve Operability, Rev. 9

WO 113767670, 1-SR-3.6.1.3.5(HPCI),

Section 1R20: Refueling and Other Outage Activities

1-GOI-100-3A, Refueling Operations (RX Vessel Disassembly and Floodup), Rev. 23

1-GOI-100-3B, Refueling Operations, (Reactor Cavity Letdown and Vessel Re-Assembly),

Rev.21

0-GOI-100-3A, Refueling Operations (In-Vessel Operations), Rev. 55

0-GOI-100-3B, Operations in Spent Fuel Storage Pool Only, Rev. 49

0-GOI-100-3C, Fuel Movement Operations During Refueling, Rev. 68

NPG-SPP-05.8, Special Nuclear Material Control, Rev. 3

Fuel Assembly Transfer Form, BFN Nuclear Plant, Transfer Operation No. BFN-1-129

PER 635794, Fuel Move Orientation Error on 11/01/2012

PER 636511, FME on Bundle JYP303 at position 31-06

PER 637790, Fuel Assembly JYE 347 was found improperly seated during core verification

PER 651334, Reactor Pressure Vessel inner O-ring leak ODMI

Tagout: 1-TO-2012-003, Clearance 1-071-0006, RCIC Pump CST Test Valve

Tagout: 1-TO-2012-003, Clearance 1-073-0021, HPCI Pump Injection Valve

Tagout: 1-TO-2012-003, Clearance 1-074-0018B, RHR System I Drywell Spray Outboard Valve

Attachment

13

Section 1R22: Surveillance Testing

Diesel Generator C Emergency Unit 1 Load Acceptance Test; 0-SR-3.8.1.9(C), Rev 7

Diesel Generator Operating Data Acquisition, 0-TI-298, Rev 14

Worker Orders: 111593354, 113745274, 113681157

0-47E610-77-1, Mechanical Control Diagram Radwaste System, Rev. 59

3-47E852-2, Flow Diagram for Clean Radwaste & Decontamination Drainage, Rev. 36

0-47W600-99, Mechanical Instrument and Controls, Rev. 03

Browns Ferry Unit 3 Technical Requirements Manual (TRM)

1-SR-3.5.1.8, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at 150

psig Reactor Pressure, Rev. 11

WO 113195499, 1-SR-3.5.1.8

WO 112923717, Install HPCI Spool Piece to Support 150 LB Test

Section 1EP4: Emergency Action Level and Emergency Plan Changes

Change Packages

CECC EPIP-8, Dose Assessment Staff Activities During Nuclear Plant Radiological

Emergencies, Revision 37

TVA Radiological Emergency Plan, Revision 97 and 98

Section 4OA2: Identification and Resolution of Problems

PER 618132, Relay 71X-85-45H3 Failure

ACE PER 618132, Unit 2 Half Scram Due to Failed HFA Relay

PER 571080, U3 Received 'A' Channel Half Scram due to failed relay

ACE PER 571080, Unit 3 Half Scram due to Failed HFA Relay

PER 621034, Unit 2 RPS Relay Failure

PER 624280, Lower Tier B Level 571080 on Relays had Incorrect Cause and Extent of

Condition

PER 626992, 1-RLY-064-16A-K78 is normally energized, safety related with Clear Lexan Spool

PER 630742, Discrepancy in NPG Procedures

PER 644806, HFA Relay Premature Failure

PER 654390, As part of Extent of Condition, Replace Relay

PER 659897, NRC Identified Request

DWG 2-730E915, Sheet 7, Reactor Protection System, Rev. 26

DWG 2-730E915RF, Sheet 11, Reactor Protection System, Rev. 14

DWG 2-730E915RF, Sheet 12, Reactor Protection System, Rev. 16

TVA Letter on IE Bulletin 84-02, dated July 10, 1984

NPG-SPP-09.18, Integrated Equipment Reliability Program, Rev. 3

NCO840128002, Commitment Documentation for Unit 1

NCO840128003, Commitment Documentation for Unit 2

NCO840128004, Commitment Documentation for Unit 3

BFN-ODM-4.16, Operator Workarounds/Burdens/Challenges, Rev. 4

OPDP-1, Conduct of Operations, Rev. 24

BFN-OPS-S-12-009, BFN Operations Snapshot Self Assessment, Operator Workarounds,

Burdens, and Challenges, dated 9/12-13/2012.

Browns Ferry Daily Alignment Plan of the Day, various dates from 5/2012 to 1/2013.

PER 478937, Workarounds from Equipment Failures Have Been Adverse to Operator Time

SR 665557

Attachment

14

NPG-SPP-02.7, PER Trending, Rev. 3

NPG-SPP-02.8, Integrated Trend Review, Rev. 3

Integrated Trend Report Q3FY12, dated 8/13/2012

Integrated Trend Report Q4FY12, dated 10/31/2012

PIDP-20, Corrective Action Program Lower Level Metrics, Rev. 2

Site Cap Health Monitor, 12/2012

CAP Gap Plan Status Update, 12/30/2012

PER Closure Review Team Revised Charter, 11/16/2012

PER Closure Review Team Charter, 4/14/2012

PER 637135, U1R9 As-Found LLRTs that exceeded Administrative Limits

PER 621027

PER 625651

PER 635998

PER 609687

PER 618693

PER 622477

PER 533052

PER 533698

PER 533787

CRP-ENG-F-12-011, Self Assessment of EQ Program at Browns Ferry, 6/11/2012

CRP-ENG-05-002, Self Assessment of EQ Program, Assessment Period 6/27 - 8/1/2005

EQ Program Health Report, 1/1 - 6/30/2012

EQ Program Health Report, 7/1 - 12/31/2011

NPG-SPP-03.14, Corrective Action Program Screening and Oversight, Rev. 10

NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Rev. 10

1-47E225-100, Unit 1 Harsh Environment Data, Drawing Series Index, Notes and References,

Rev 003, 3/18/2003

2-47E225-100, Unit 2 Harsh Environment Data, Drawing Series Index, Notes and References,

Rev 002, 7/10/1987

1-47E225-103, Unit 1 Harsh Environment Data, EL 519, Rev 002, 3/18/2003

2-47E225-103, Unit 2 Harsh Environment Data, EL 519, Rev 002, 7/10/1987

EQIR No. BFNEQ11826, Delete Pressure Switch 2-PS-075-0007 from EQDP and QMDS,

1/1/2011

EQIR No. BFNEQ09781, Incorporate the Field Verification and Certificate of Conformance for

the Following Motors into the Subject EQDP and QMDS, 10/16/2012

R06121214989, EQIR No. BFNEQ121014, Subject EQDP BFN0EQ-CABL-007, 12/14/2012

R06121220052, EQIR No. BFNEQ11833, Subject EQDP BFN0EQ-IXT-001, 9/21/2011

R06121220051, EQIR No. BFNEQ09776R1, Subject EQDP BFN0EQ-IXT-001, 11/15/2012

R06121220055, EQIR No. BFNEQ90791, Subject EQDP BFN0EQ-IXT-001, 8/25/2009

R06121220053, EQIR No. BFNEQ12914, Subject EQDP BFN0EQ-IXT-001, 12/20/2012

R06121220054, EQIR No. BFNEQ12975, Subject EQDP BFN0EQ-IXT-001, 12/20/2012

R06121220056, EQIR No. BFNEQ12892, Subject EQDP BFN0EQ-IXT-001, 12/20/2012

R06121210952, EQIR No. BFNEQ09785, Subject EQDP BFN0EQ-CSC-001, 8/17/2009

W87060711010, EQIR No. BFNEQ06437, Subject EQDP BFN0EQ-CABL-007, 7/11/2006

W87051121002, EQIR No. BFNEQ05381, Subject EQDP BFN0EQ-CABL-007, 11/16/2005

R06121212978, EQIR No. BFNEQ121004, Subject EQDP BFN0EQ-CABL-034, 11/28/2012

R06121024805, EQIR No. BFNEQ09809, Subject EQDP BFN0EQ-CABL-034, 8/18/2009

R06121026825, EQIR No. BFNEQ12954, Subject EQDP BFN0EQ-CSC-001, 10/24/2012

Attachment

15

R06121026821, EQIR No. BFNEQ11845, Subject EQDP BFN0EQ-CSC-001, 11/22/2011

R06121026823, EQIR No. BFNEQ12898, Subject EQDP BFN0EQ-IZS-004, 10/24/2012

PER 283220, WO 111615866 Replaced 2-PS-075-0007 with a Model Number Different from

that Previously Installed

PER 238931, EQIRs in Violation of SPP-9.2

PER 624137 NPG-SPP-09.3 Does Not Require Timely DCN Closure

PER 559420 EQ Equipment Installed without Supporting EQ Binder

PER 575444 Drawing discrepancy on 1-47E225-103 and 2-47E225-103

EQ Components Selected for Walkdown

BFN-2-CSC-023-0048 Conduit Seal for 2-FT-023-0048

BFN-2-CSC-074-0064 Conduit Seal for 2-FT-074-0064

BFN-2-FT-023-0048 RHR HTX B SW Flow

BFN-2-FT-074-0064 RHR Loop 2 Flow

BFN-2-HS-074-0098B RHR Pump B Suction Crosstie Valve

BFN-2-HS-074-0099B RHR Pump D Suction Crosstie Valve

BFN 2-MTR-074-0028 RHR Pump 2B Motor

BFN 2-PS-074-0031A RHR Pump B Discharge Pressure

BFN 2-PS-074-0031B RHR Pump B Discharge Pressure

BFN 2-MTR-073-0026 HPCI Suppression Pool Inboard Suction Valve

BFN-2-MTR-073-0027 HPCI Suppression Pool Outboard Suction Valve

BFN-2-MVOP-073-0026 HPCI Suppression Pool Inboard Suction Valve Operator

BFN-2-MVOP-073-0027 HPCI Suppression Pool Outboard Suction Valve Operator

BFN-2-CSC-075-021 Conduit Seal at 2-FT-75-21

BFN-2-CSC-075-0057A Conduit Seal Connector for Limit Switch on FCV-75-57

BFN-2-FSV-075-0057 Solenoid Valve for Drain Pump A Bypass Line

BFN-2-FT-075-0021 System 1 Flow CS Flow SQ RT

BFN-2-MTR-075-0005 Core Spray Pump A Motor and Core Spray Pump C Motor

BFN-1-CSC-023-0048 Conduit Seal for 1-FT-023-0048

BFN-1-CSC-074-0064 Conduit Seal for 1-FT-074-0064

BFN-1-FT-023-0048 RHR HTX B SW Flow

BFN-1-FT-074-0064 RHR Loop 2 Flow

BFN-1-HS-074-0098B RHR Pump B Suction Crosstie Valve

BFN-1-HS-074-0099B RHR Pump D Suction Crosstie Valve

BFN 1-MTR-074-0028 RHR Pump 2B Motor

BFN 1-PS-074-0031A RHR Pump B Discharge Pressure

BFN 1-PS-074-0031B RHR Pump B Discharge Pressure

BFN 1-MTR-073-0026 HPCI Suppression Pool Inboard Suction Valve

BFN-1-MTR-073-0027 HPCI Suppression Pool Outboard Suction Valve

BFN-1-MVOP-073-0026 HPCI Suppression Pool Inboard Suction Valve Operator

BFN-1-MVOP-073-0027 HPCI Suppression Pool Outboard Suction Valve Operator

BFN-1-CSC-075-0021 Core Spray System Flow

BFN-1-CSC-075-0057A Pressure Suppression High Level Control

PERs Generated as a Result of Inspection

SR 659324 Maximo Indicates Incorrect Room Location for Valve Operator and Valve Motor

PER 658970 Secord Party Verification Not Performed for Required SPV Fields in Maximo

PER 658469 Fields in Maximo Not Having Required Second Party Verification

Attachment

16

PER 659324 Incorrect Room Reference in Maximo for 1-MVOP-73-26 and 1-MTR-73-26

PER 660812 Maximo indicates incorrect Room location for valve operator and valve motor.

Section 4OA3: Event Follow-up

FSAR Unit 1

R14 981211 106 TVA Alternate Source Term Calculation Input Parameters

R 92 960718 850 TVA Alternate Source Term Calculations

PER 590208 Past Operability Determination for SBGT C relative humidity heater inoperable

PER 604350 Breaker for SBGT C relative humidity heater inoperable

LER 50-259/2012-008 SBGT Train C inoperable longer than Technical Specification allowable

time

WO 113759132 Standby Gas Treatment Train C filter heating element lost power

Root Cause Analysis (RCA) SBGT Train C inoperable

Section 4OA5: Other Activities

0-AOI-100-3, Flood Above Elevation 558, Rev. 35

CTP-FWD-100, Flood Protection Walkdowns NEI 12-07, Rev. 0

FSAR Appendix 2.4A, Browns Ferry Nuclear Plant Maximum Possible Flood, BFN-24

MPI-0-000-INS001, Inspection of flood Protection Devices, Rev. 12

NEI 12-07, Guidelines for Performing Verification Walkdowns of Plant Flood Protect6ion

Features, Rev. 0-A

NRR Japan Lessons-Learned Project Directorate Letter dated 5/31/2012, Endorsement of

Nuclear Energy Institute (NEI) 12-07, Guidelines for Performing Verification Walkdowns of

Plant Flood Protection Features

PER 469640, Aggregate Impact of RHRSW Pump Room Watertight Door Degradations

PER 647926, NTTF 2.3 Flooding Walkdown Small Available Physical Margin

PER 568642, RHRSW Pump Room Watertight Door Degradations

PER 589442, NRC Concerns of Gaps in the NTTF-2.3 Flood Walkdown Scope

Adverse Employment Action Procedure

TVA Fleet ECP Reports

Attachment

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

Attachment

ML13039A321

Text

Navigation menu

Search