ML13039A321
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) | |
See also: IR 05000259/2012005
Text
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
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
- 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
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)
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.
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 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
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
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
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
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
SNM - special nuclear material
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