ML12227A711
| ML12227A711 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/14/2012 |
| From: | Eugene Guthrie Division Reactor Projects II |
| To: | James Shea Tennessee Valley Authority |
| References | |
| IR-12-003, IR-12-502 | |
| Download: ML12227A711 (72) | |
See also: IR 05000259/2012003
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
August 14, 2012
Mr. Joseph W. Shea
Vice President, Nuclear Licensing
Tennessee Valley Authority
Chattanooga, TN 37402-2801
SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION
REPORT 05000259/2012003, 05000260/2012003, 05000296/2012003,
05000259/2012502, 05000260/2012502, AND 05000296/2012502
Dear Mr. Shea:
On June 30, 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 July 10, August 10 and 14th, 2012, with Mr.
Keith Polson 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.
One NRC identified and 3 self revealing findings of very low safety significance (Green) were
identified during this inspection. Three of these findings were determined to involve violations of
NRC requirements. Further, a licensee-identified violation which was determined to be of very
low safety significance is listed in this report. The NRC is treating the violations as non-cited
violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy. If you contest these
non-cited violations, you should provide a response within 30 days of the date of this inspection
report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document
Control Desk, Washington DC 20555-0001, with copies to: (1) the Regional Administrator,
Region II; (2) the Director, Office of Enforcement, United States Nuclear Regulatory
Commission, Washington, DC 20555-0001; and (3) the NRC Resident Inspector at the Browns
Ferry Nuclear Plant.
In addition, if you disagree with any cross-cutting aspect assignment in the report, you should
provide a response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the
Browns Ferry Nuclear Plant.
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 (the Public Electronic Reading Room).
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 05000259/2012003,
05000260/2012003, 05000296/2012003
cc w/encl. (See page 3)
_________________________ X SUNSI REVIEW COMPLETE
OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS
SIGNATURE Via email Via email Via email Via email BRB /RA for/ BRB /RA for/ BRB /RA for/
NAME DDumbacher CStancil PNiebaum LPressley MSpeck RHamilton CDykes
DATE 08/14/2012 08/14/2012 08/14/2012 08/14/2012 08/14/2012 08/14/2012 08/14/2012
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRS RII:DRS RII:DRP RII:DRP
SIGNATURE Via email Via email Via email EFG /RA/
NAME RKellner MCoursey CKontz EGuthrie
DATE 07/26/2012 08/14/2012 08/14/2012 08/14/2012
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
J. Shea 3
cc w/encl: James L. McNees, CHP
K. J. Polson Director
Site Vice President Office of Radiation Control
Browns Ferry Nuclear Plant Alabama Dept. of Public Health
Tennessee Valley Authority P. O. Box 303017
Electronic Mail Distribution Montgomery, AL 36130-3017
C.J. Gannon
General 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
Manager, Corporate Nuclear Licensing -
Tennessee Valley Authority
Electronic Mail Distribution
Edward J. Vigluicci
Assistant General Counsel
Tennessee Valley Authority
Electronic Mail Distribution
T. A. Hess
Tennessee Valley Authority
Electronic Mail Distribution
Chairman
Limestone County Commission
310 West Washington Street
Athens, AL 35611
Donald E. Williamson
State Health Officer
Alabama Dept. of Public Health
RSA Tower - Administration
Suite 1552
P.O. Box 30317
Montgomery, AL 36130-3017
J. Shea 4
Letter to Joseph W. Shea from Eugene Guthrie dated August 14, 2012
SUBJECT: BROWNS FERRY NUCLEAR PLANT - NRC INTEGRATED INSPECTION
REPORT 05000259/2012003, 05000260/2012003, 05000296/2012003,
05000259/2012502, 05000260/2012502, AND 05000296/2012502
Distribution w/encl:
C. Evans, RII
L. Douglas, RII
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.: 05000259/2012003, 05000260/2012003, 05000296/2012003,
05000259/2012502, 05000260/2012502, 05000296/2012502
Licensee: Tennessee Valley Authority (TVA)
Facility: Browns Ferry Nuclear Plant, Units 1, 2, and 3
Location: Corner of Shaw and Nuclear Plant Roads
Athens, AL 35611
Dates: April 1, 2012, through June 30, 2012
Inspectors: D. Dumbacher, Senior Resident Inspector
C. Stancil, Senior Resident Inspector
P. Niebaum, Resident Inspector
L. Pressley, Resident Inspector
M. Speck, Senior Emergency Preparedness Inspector (1EP2, 1EP3,
1EP5, 4OA1)
R. Hamilton, Senior Health Physicist (2RS1, 2RS2, 2RS6, 4OA1)
C. Dykes, Health Physicist (2RS7)
R. Kellner, Health Physicist (2RS8)
M. Coursey, Reactor Inspector (1R08)
Approved by: Eugene F. Guthrie, Chief
Reactor Projects Special Branch
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000259/2012003, 05000260/2012003, 05000296/2012003, 05000259/2012502,
05000260/2012502, 05000296/2012502; 04/01/2012 -06/30/2012; Browns Ferry Nuclear Plant,
Units 1, 2 and 3; Operability Evaluations, Radioactive Solid Waste Processing and Radioactive
Material Handling, Storage, and Transportation, and Event Follow-Up.
The report covered a three month period of inspection by resident and regional inspectors. Four
findings were identified. The significance of most findings is identified by their color (Green,
White, Yellow, and Red) using Inspection Manual Chapter (IMC) 0609, Significance
Determination Process (SDP); and, the cross-cutting aspects were determined using IMC
0310, Components Within the Cross-Cutting Areas. Findings for which the SDP does not
apply may be Green or be assigned a severity level after NRC management review. The 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.
NRC Identified and Self-Revealing Findings
Cornerstone: Initiating Events
- Green. A self-revealing finding (FIN) was identified for the licensees failure to
perform preventive maintenance on the Unit 3 Main Control Room (MCR)
annunciator power supplies. As a result, a power supply failed which led to a fire in
annunciator panel 3-X-55-5A in the Unit 3 control room. The licensee initiated
actions to extinguish the fire, replace the two affected power supplies and develop a
preventive maintenance program to replace the power supplies every ten years.
Additional corrective actions to replace all power supplies that have been installed for
more than four years are pending. This was captured in the licensees corrective
action program as problem event report (PER) 496592.
The performance deficiency was determined to be more than minor because it was
considered sufficiently similar to example 4.f of Inspection Manual Chapter (IMC)
0612, Appendix E, for an issue that resulted in a fire hazard in a safety-related area
of the plant. The finding was associated with the Initiating Events Cornerstone and
required a phase 3 analysis in accordance with IMC 0609 because the finding
increased the likelihood of, and actually caused, a fire in the Unit 3 control room.
The phase 3 analysis determined that without an impact to additional plant
equipment, or a major impact on human action failure rates, the finding was
determined to be Green. The cause of this finding was related to the cross cutting
aspect of Problem Identification in the Corrective Action Program component of the
Problem Identification and Resolution area because the licensee should have
recognized the electrolytic capacitors were installed beyond their recommended
service life and scheduled replacement prior to their failure P.1(a). (Section
4OA3.6)
Enclosure
3
Cornerstone: Mitigating Systems
- Green. An NRC-identified non-cited violation (NCV) of the Technical Specifications
5.4.1.a was identified for the licensees failure to maintain an Emergency Equipment
Cooling Water (EECW) pump flood barrier in accordance with written procedures
which resulted in the inoperability of two other safety related pumps. The licensee
immediately restored the flood protection configuration of the C Residual Heat
Removal Service Water (RHRSW) pump room by properly re-installing the flood
protection cover and permanently stenciled the aluminum plate with the required
procedure for installation. The licensee entered this issue into their corrective action
program as PER 532050.
The finding was more than minor because it was associated with the Mitigating
Systems cornerstone attribute of Protection Against External Events, and adversely
affected the cornerstone objective to ensure the availability, reliability, and capability
of RHRSW pumps to perform their intended safety function during a design basis
flooding event. Specifically, the improper re-installation of an external flood
protection cover resulted in the inoperability of two Residual Heat Removal Service
Water (RHRSW) pumps. The significance of this finding was evaluated in
accordance with the IMC 0609 Attachment 4, Phase 1- Initial Screening and
Characterization of Findings, which required a Phase 3 analysis because the finding
involved the degradation of equipment designed to mitigate a flooding event and it
was risk significant due to external initiating event core damage sequences. The
finding was determined to be Green because of the short exposure time, and the low
likelihood of the flood. The cause of this finding was directly related to the cross
cutting aspect of Supervisory Oversight in the Work Practices component of the
Human Performance area, because of the foremans assumption that workers knew
to restore the flood protection cover to meet procedural requirements without a
formal pre-job brief H.4(c). (Section 1R15)
Cornerstone: Public Radiation Safety
- Green. A self-revealing non-cited violation (NCV) of 10 CFR 71.5, Transportation of
Licensed Material, was identified by inspectors for the licensees failure to comply
with Department of Transportation (DOT) regulations during shipment of radioactive
materials. Specifically, the licensee failed to ensure proper packaging of two DOT 7A
Type A packages as required by Department of Transportation (DOT) regulations in
49 CFR 173.475, Quality Control Requirements Prior To Each Shipment Of Class 7
(Radioactive) Materials. This issue has been entered into the licensees corrective
action program as SR 570902.
The finding was more than minor because it is associated with the Public Radiation
Safety Cornerstone, Plant Facilities/Equipment and Instrumentation attribute,
involving transportation packaging and adversely affected the cornerstone objective,
to ensure adequate protection of public health and safety from exposure to
radioactive materials released into the public domain as a result of routine civilian
nuclear reactor operation. Specifically, the failure to correctly secure the package
Enclosure
4
contents to prevent movement could have resulted in damage or failure of the
container during transportation. The finding was determined to be of very low safety
significance (Green) because it did not involve radiation limits being exceeded, a
package breach, a certificate of compliance issue, a low-level burial ground non-
conformance, or a failure to make emergency notifications. The cause of this finding
was directly related to the cross cutting aspect of Documents, Procedures and
Component Labeling in the Resources component of the Human Performance area
because the licensee did not effectively incorporate package design specifications
into their transportation program to ensure that all internal restraining devices are
correctly installed to secure the CRDM in place to prevent damage to the transport
package. (H.2(c)) (Section 2RS8)
- Green. A self-revealing non-cited violation (NCV) of 10 CFR 71.5, Transportation of
Licensed Material, was identified by inspectors for the licensees failure to comply
with Department of Transportation (DOT) regulations during shipment of radioactive
materials. Specifically, the licensee failed to ensure proper closure of a DOT 7A Type
A package as required by Department of Transportation (DOT) regulations in 49
CFR 173.475, Quality Control Requirements Prior To Each Shipment Of Class 7
(Radioactive) Materials. This issue has been entered into the licensees corrective
action program as SR 571151.
The finding was more than minor because it is associated with the Public Radiation
Safety Cornerstone, Plant Facilities/Equipment and Instrumentation attribute,
involving transportation packaging and adversely affected the cornerstone objective,
to ensure adequate protection of public health and safety from exposure to
radioactive materials released into the public domain as a result of routine civilian
nuclear reactor operation. Specifically, the failure to apply the correct torque to the
package closure bolts could have resulted in incomplete sealing of the container or
failure of the cover bolts during transportation. The finding was determined to be of
very low safety significance (Green) because it did not involve radiation limits being
exceeded, a package breach, a certificate of compliance issue, a low-level burial
ground non-conformance, or a failure to make emergency notifications. The cause
of this finding was directly related to the cross cutting aspect of Documents,
Procedures and Component Labeling in the Resources component of the Human
Performance area because the licensee did not effectively incorporate the vendor
provided container loading and shipping instructions into their work package and
transportation program to ensure correct torque values were used to close the
shipping container. (H.2(c)) (Section 2RS8).
Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at full power for most of the report period except for an unplanned downpower
on June 29, 2012, to 75 percent power to reduce load on the B Phase Main Bank Transformer
due to a lifting oil pressure relief. The unit returned to full power on June 30, 2012.
Unit 2 operated at full power for most of the report period except for one planned and one
unplanned downpower. On April 20, 2012, the unit performed a planned downpower to 66
percent power for rod pattern adjustment, scram time testing and turbine valve testing. The unit
returned to full power on April 22nd. On May 15, 2012, the unit performed an unplanned
downpower to 92 percent power to insert control rod 30-51 for scram outlet valve repair and
returned to full power the same day.
Unit 3 operated at full power for most of the report period except for one planned downpower,
one manual and two automatic scrams, and one unplanned downpower. On April 6, 2012, the
unit was shutdown for a scheduled refueling outage that lasted 49 days. The unit was restarted
on May 19th. On May 22nd, an automatic scram occurred from 19.5 percent power with the
main turbine generator offline due to a 3A Unit Station Service Transformer differential relay trip
caused by incorrect relay setting. On May 24, 2012, during reactor startup and heatup an
unplanned manual scram occurred as a result of a partial control rod insertion caused by a
combination of a signal spike and an inappropriate operator downrange on separate
intermediate power range monitors. The unit restarted the same day. On May 29, 2012, a main
generator current transformer manufactured and installed with reverse polarity caused an
automatic scram from 75 percent power. The unit restarted on June 2nd and returned to full
power on June 5th. On June 6th, the unit performed an unplanned downpower from 96 percent
power to 75 percent power to remove the 3B condensate booster pump with high moisture in its
oil system from service. The unit returned to full power on June 8, 2012.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
.1 Offsite and Alternate Alternating Current (AC) Power Systems Readiness
a. Inspection Scope
Prior to the summer season, inspectors reviewed electrical power design features, onsite
risk and work management procedures, and corporate transmission and power supply
procedures to verify appropriate operational oversight and assurance of continued
availability of offsite and alternate AC power systems. Inspectors verified that
communications protocols existed between the transmission system operator and
Browns Ferry Nuclear Plant for coordination of off-normal and emergency events
affecting the plant, event details, estimates of return-to-service times, and notifications of
grid status changes. Inspectors also verified that procedures included controls to
Enclosure
6
adequately monitor both offsite AC power systems (including post-trip voltages) and
onsite alternate AC power systems for availability and reliability. Furthermore,
inspectors interviewed onsite licensed operators and offsite transmission personnel to
determine their understanding and implementation of the power monitoring and
assessment process. Inspectors reviewed the material condition of offsite AC power
systems and onsite alternate AC power systems to the plant, including switchyard and
transformers. This review included review of outstanding work orders affecting these
systems and a walkdown of the switchyard with operations personnel to ensure the
systems will continue to provide appropriate as designed capabilities. This activity
constituted one Offsite and AC Readiness sample.
b. Findings
No findings were identified.
.2 Readiness for Seasonal Extreme Weather Conditions
a. Inspection Scope
Prior to and during the onset of hot weather conditions, the inspectors reviewed the
licensees implementation of 0-GOI-200-3, Hot Weather Operations. The inspectors
also reviewed the Hot Weather Discrepancy Log; and discussed implementation of
0-GOI-200-3 with responsible Operations personnel and management. Furthermore, the
inspectors conducted walkdowns of potentially affected risk significant equipment
systems located in the Unit 1 and 2 Diesel Generator Building, and the Unit 3 Diesel
Generator Building. The inspectors also performed a walkdown of the Standby Gas
Treatment (SBGT) Building. This activity constituted 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 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.
Enclosure
7
- Unit 1&2 A Emergency Diesel Generator
- Unit 3 Residual Heat Removal System - Division II
- Unit 1 Reactor Core Isolation Cooling (RCIC) System
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 the four
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. Furthermore, the
inspectors reviewed applicable portions of the Fire Protection Report, Volumes 1 and 2,
including the applicable Fire Hazards Analysis, and Pre-Fire Plan drawings, to verify that
the necessary firefighting equipment, such as fire extinguishers, hose stations, ladders,
and communications equipment, was in place. This activity constituted four Fire
Protection inspection samples.
- Unit 2 Reactor Building Elevations 519, 541, and 565 west of column line R11 (FZ 2-
1)
- Unit 3 Reactor Building, EL 593 and residual heat removal (RHR) heat exchanger
rooms, EL 565, and 593 near column R15-S and R21-S (FZ 3-3)
- Unit 1, Control Building, EL 593 (FA 16)
- Unit 1,2, and 3 Turbine Building Deluge Sprinkler Control Stations Affecting Control
Bay (FA 25)
b. Findings
No findings were identified.
Enclosure
8
1R07 Heat Sink Performance
.1 Annual Review
a. Inspection Scope
The inspectors examined activities associated with Unit 3 RHR Heat Exchangers. The
inspectors also reviewed design basis documents, calculations, test procedures,
maintenance procedures and preventive maintenance procedures and results to
evaluate the licensees program for maintaining heat sinks in accordance with the
licensing basis. Specifically inspectors reviewed modifications performed on the Unit 3
RHR Heat Exchanger Flanges. Inspectors reviewed available performance testing
documentation of the 3A and 3C RHR Heat Exchangers.
In addition, the inspectors reviewed the licensees implementation of the GL 89-13
program. Inspectors reviewed associated PERs and corrective actions to verify that the
licensee was identifying issues and correcting them. The inspectors performed
walkdowns of key components of the Unit 3 RHR system to verify material conditions
were acceptable and physical arrangement matched procedures and drawings. This
activity constituted one Annual Heat Sink sample.
b. Findings
No findings were identified.
1R08 Inservice Inspection (ISI) Activities (71111.08G, Unit 3)
a. Inspection Scope
Non-Destructive Examination (NDE) Activities and Welding Activities: From April 16 to
April 20, 2012, the inspectors conducted an on-site review of the implementation of the
licensees Inservice Inspection (ISI) Program for monitoring degradation of the reactor
coolant system, emergency feedwater systems, risk-significant piping and components,
and containment systems in Unit 3. The inspectors activities included a review of non-
destructive examinations (NDEs) to evaluate compliance with the applicable edition of
the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel
Code (BPVC),Section XI (Code of record: 2001 Edition with 2003 Addenda), and to
verify that indications and defects (if present) were appropriately evaluated and
dispositioned in accordance with the requirements of the ASME Code,Section XI,
acceptance standards.
The inspectors directly observed 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 and defects were detected, to evaluate if they were dispositioned in
accordance with the ASME Code or an NRC-approved alternative requirement.
Enclosure
9
- UT Exam of Weld DRHR-03-03, 3-FCV-74-53, Low Pressure Coolant Injection
(LPCI) Loop I Inlet
- UT Exam of Weld DSRHR-03-04, 3-HCV-74-55, 24 in. inlet for Recirculation Loop B
The inspectors reviewed records of the following NDEs mandated by the ASME Code
Section XI to evaluate compliance with the ASME Code Section XI and Section V
requirements and, if any indications and defects were detected, to evaluate if they were
dispositioned in accordance with the ASME Code or an NRC-approved alternative
requirement.
- VT Exam of RPV-WASH-3-50, Reactor Pressure Vessel Stud Washer
- EVT of BFN-3-RPV-068-RA048 Standpipe in Unit 3 Steam Separator
The inspectors reviewed associated documents for the welding activities referenced
below in order to evaluate compliance with procedures and the ASME Code. The
inspectors reviewed the work order, repair and replacement plan, weld data sheets,
welding procedures, procedure qualification records, welder performance qualification
records, and NDE reports.
- Work Order 04-719493-003, 3-FCV-073-016 HPCI Turbine Steam Supply Valve
- Work Order 08-718716-004, Replace Strain Gauges on MS Lines
During non-destructive surface and volumetric examinations performed since the
previous refuelling outage, the licensee did not identify any relevant indications that were
analytically evaluated and accepted for continued service. Therefore, no NRC review
was completed for this inspection procedure attribute.
Identification and Resolution of Problems: The inspectors performed a review of a
sample of ISI-related problems which were identified by the licensee and entered into
the corrective action program as Problem Evaluation Reports (PERs). The inspectors
reviewed the PERs to confirm the licensee had appropriately described the scope of the
problem, and had initiated corrective actions. The review also included the licensees
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. The corrective action
documents reviewed by the inspectors are listed in the report attachment.
b. Findings
No findings were identified.
Enclosure
10
1R11 Licensed Operator Requalification
.1 Resident Inspector Quarterly Review
a. Inspection Scope
On June 11, 2012, the inspectors observed an as-found licensed operator requalification
simulator examination according to Unit 2 Simulator Exercise Guide OPL173.S039. The
scenario involved Partial Loss of Reactor Building Closed Cooling Water, Loss of I & C
Bus B, Anticipated Transient without Scram (ATWS), Lower Water Level (C-5) for Power
Control with Bypass Valves.
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 Abnormal Operating Instructions (AOIs), and
Emergency Operating Instructions (EOIs)
- Timely and appropriate Emergency Action Level declarations per Emergency Plan
Implementing Procedures (EPIP)
- Control board operation and manipulation, including high-risk operator actions
- Command and Control provided by the Unit Supervisor and Shift Manager
The inspectors attended the post-examination critique to assess the effectiveness of the
licensee evaluators and to verify that licensee-identified issues were comparable to
issues identified by the inspector. The inspectors 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 counts for one Observation of Requalification Activity
inspection sample.
b. Findings
No findings were identified.
.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.
Enclosure
11
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 License Operator Requalification inspection sample and one
Control Room Observation inspection sample.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness
.1 Routine
a. Inspection Scope
The inspectors reviewed three 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-03.1, 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
three Maintenance Effectiveness inspection samples.
Enclosure
12
- Unit 1, 2 and 3 Intermediate Range Monitors - System 092
- Unit Common Residual Heat Removal Service Water (RHRSW) Pump Room
Watertight Door Functional Failures
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
.1 Risk Assessment and Management of Risk
a. Inspection Scope
For planned online work and/or emergent work that affected the combinations of risk
significant systems listed below, the inspectors examined five 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 (RMAs) 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 implementation. This activity
constituted five Maintenance Risk Assessment inspection samples.
- Planned refueling outage work on both loops of Unit 3 RHR, 3B Fuel Pool Cooling
pump, Unit 3 500KV off-site power, 3C EDG, 1A Condenser Circulating Water Pump,
1A Control Bay chiller and AHU, B Fire Pump, RCW Booster Pumps 2A and 3A, C3
EECW Pump, and C RHRSW Common Header
- Emergent work on D Emergency Diesel Generator (EDG) for troubleshooting and
corrective maintenance, Unit 2 C Residual Heat Removal (RHR) Heat Exchanger
OOS for piping leak repair, Intake Pumping Station Vent Fan A and B work, and
Common Switchyard Centered LOOP High Risk due to Unit 3 Transformer activities.
EDG, 3B Fuel Pool Cooling (FPC) Pump, 3C and 3D 4kV Shutdown Boards and
Standby Gas Treatment (SBGT) Train C
Unit 1/2 risk associated with RHR Heat Exchanger 2C and RHRSW Pump A3 OOS
and, Common Switchyard Centered LOOP High Risk due to Unit 3 Transformer
activities.
- Planned Unit 2 risk with High Pressure Coolant Injection pump and D EDG OOS
Enclosure
13
b. Findings
No findings were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the six operability/functional evaluations listed below to verify
technical adequacy and ensure that the licensee had adequately assessed TS
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, 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.
- RHRSW Rooms Appendix R Fire Barrier Impacted by Tarpaulin (PER 492957)
- Emergency Equipment Cooling Water (EECW) check valve not fully closed (PER
520497)
- RHRSW Pump Room Watertight Door BFN-0-DOOR-260-C-RHRSW Degraded
(PER 469640)
- Past Operability for C3 Emergency Equipment Cooling Water (EECW) Pump
Foundation Hole Flood Protection Cover Inadequate Installation (PER 532050)
- Units 1,2 and 3 EECW yard drain basins partially blocked, (PER 569282)
- Unit 1 HPCI Turbine Stop Valve, 1-FCV-073-0018, Failed to Trip (PER 539040)
b. Findings
Two findings were identified. One finding is documented as a licensee identified violation
in Section 4OA7.
1) Introduction: The NRC identified a Green non-cited violation (NCV) of Technical
Specification 5.4.1.a for the licensees failure to maintain an Emergency Equipment
Cooling Water (EECW) pump flood barrier in accordance with written procedures which
resulted in the inoperability of two other safety related pumps.
Description:
The safety related Residual Heat Removal Service Water (RHRSW) pumps are housed
in the A, B, C, and D rooms of the intake pumping station. UFSAR Section 12.2.7.1.1
states, in part, that each room is designed to protect the RHRSW pumps from water and
wave forces resulting from a probable maximum flood (PMF) scenario. During
Enclosure
14
maintenance activities, the licensee maintained the design flood protection configuration
through implementation of properly written work instructions.
The C3 Emergency Equipment Cooling Water (EECW) pump is located in the C
RHRSW pump room with two similarly designed C1 and C2 RHRSW pumps. On March
26, 2012, the licensee had removed C3 pump from service for maintenance. The C3
pump and motor had been disassembled and the pump column removed from the intake
sump pit through the pump base plate and foundation leaving an approximate 22 inch
diameter hole. The hole was protected against flooding by a temporary 1/4 inch thick
aluminum cover plate, over a rubber gasket and secured with 8 foundation bolts. The
flood cover was prescribed by work order 112744581 and implemented by maintenance
procedures MCI-0-023-PMP002, Maintenance of EECW and RHRSW Pumps, and MCI-
0-023-PMP003, Emergency Equipment Cooling Water and Residual Heat Removal
Service Water Pump Removal and Installation.
On April 2, 2012, maintenance personnel removed the flood protection cover to facilitate
an inspection. Personnel re-installed the cover with only two bolts and nuts run down to
approximately one inch from being fully secured. On April 5, 2012, inspectors identified
and informed the licensee of the inadequate flood protection barrier. The licensee
immediately re-installed the flood protection cover in accordance with maintenance
procedures. As an added corrective action, the licensee permanently stenciled the
aluminum plate with the required procedure for installation. The licensee determined
that the workers had re-installed the flood protection cover following the inspection
assuming that it was only for foreign material exclusion. The licensee also determined
that the foreman did not direct an adequate pre-job brief and assumed the workers knew
of the procedural flood requirements. Furthermore, the licensee evaluated the
inadequate flood barrier for past operability and concluded that the C RHRSW pump
room would have flooded in the event of a PMF and that the other two RHRSW pumps
in the room, C1 and C2, would be made nonfunctional. The licensee credited the slow
progression of a PMF flood rise (four days and eight hours) to allow time to adequately
install the flood protection cover, and therefore, prevent the loss of the RHRSW pumps.
These actions were contained in licensee abnormal operating instruction 0-AOI-100-3,
Flood Above Elevation 558.
Analysis: The licensees failure to maintain an Emergency Equipment Cooling Water
(EECW) pump flood barrier in accordance with written procedures was a performance
deficiency. The finding was more than minor because it was associated with the
Mitigating Systems cornerstone attribute of Protection Against External Events, and
adversely affected the cornerstone objective to ensure the availability, reliability, and
capability of RHRSW pumps to perform their intended safety function during a design
basis flooding event. Specifically, the improper re-installation of an external flood
protection cover resulted in the inoperability of two RHRSW pumps. The significance of
this finding was evaluated in accordance with the IMC 0609 Attachment 4, Phase 1-
Initial Screening and Characterization of Findings, which required a Phase 3 analysis
because the finding involved the degradation of equipment designed to mitigate a
flooding event and was risk significant due to external initiating event core damage
sequences. A Phase 3 SDP analysis was performed by the regional Senior Reactor
Analyst using a modified NRC plant model. The model had been modified to calculate
Enclosure
15
the impact on the plant from external flooding due to the failure of the RHRSW flood
doors. The plant model was solved for a loss of condenser heat sink, with the initiating
event frequency set to 5E-3 as a conservative estimate for the external flood. Also
assumed was the unavailability of the power conversion system, since the circ water
pumps, and their power supplies would be flooded. Condensate was assumed lost
when the turbine building floods. RHRSW pumps and EECW pumps in the flooded
RHRSW room were failed by model changes for different flood door failure basic events.
This analysis failed only the C room door, which duplicated the impact of an unsecured
flood barrier. For the 4 day exposure time, the result was several orders of magnitude
below the CDF or LERF threshold for a finding of significance. The finding is Green
because of the short exposure time, and the low likelihood of the flood.
The cause of this finding was directly related to the cross cutting aspect of Supervisory
Oversight in the Work Practices component of the Human Performance area, because of
supervisions assumption that workers knew to restore the flood protection cover to meet
procedural requirements without a formal pre-job brief H.4(c).
Enforcement: TS 5.4.1.a. required that written procedures recommended in RG 1.33,
Revision 2, Appendix A, shall be established, implemented, and maintained. Item 9.a of
RG 1.33, Appendix A, stated, in part, that maintenance affecting the performance of
safety-related equipment be properly performed in accordance with written procedures
or documented instructions appropriate to the circumstances. Contrary to the above,
between April 2, and April 5, 2012, the licensee failed to properly perform maintenance
procedures MCI-0-023-PMP002 and MCI-0-023-PMP003, Section 5.0.K. Specifically,
the licensee failed to maintain a flood barrier during maintenance on C3 EECW Pump
which resulted in the inoperability of C1 and C2 RHRSW Pumps. Because this finding is
of very low safety significance (Green) and because it was entered into the licensees
corrective action program as PER 532050, this violation is being treated as a non-cited
violation consistent with the NRC Enforcement Policy. This violation was applicable to
U1, U2 and U3 and is identified as NCV 05000259, 260, 296/2012003-01, Failure to
Maintain Flood Barrier Results in Inoperable Safety Related Pumps.
1R18 Plant Modifications
a. Inspection Scope
The inspectors reviewed the two modifications listed below to verify regulatory
requirements were met, along with procedures, as applicable, such as NPG-SPP-9.3,
Plant Modifications and Engineering Change Control; NPG-SPP-9.5, Temporary
Alterations; and NPG-SPP-6.9.3, Post-Modification Testing. The inspectors also
reviewed the associated 10 CFR 50.59 screenings and evaluations and compared each
against the UFSAR and TS to verify that the modifications did not affect operability or
availability of the affected systems. Furthermore, the inspectors walked down each
modification to ensure that it was installed in accordance with the modification
documents and reviewed post-installation and removal testing to verify that the actual
impact on permanent systems was adequately verified by the tests. This activity
constituted two Plant Modification inspection samples.
Enclosure
16
- Temporary Alteration Control Form (TACF) 1-12-001-073, Removed Thermal
Insulation Attached to BFN-1-FCV-073-0016, HPCI Steam Supply Valve
- Design Change Notice (DCN) 70549, Unit 3 Reactor Water Level Flood-Up
Transmitter and Indication Loop Replacement
b. Findings
No findings were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
The inspectors witnessed and reviewed the six 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 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
six Post Maintenance Test inspection samples.
- Unit 3: Reactor Vessel Head Tensioning and subsequent Pressure Test per MSI-0-
001-VSL001, Reactor Vessel Head Disassembly and Reassembly; 3-SI-3.3.1.A,
ASME Section XI System Leakage Test of the Reactor Pressure Vessel and
Associated Piping; 3-SR-3.4.9.1(1), Reactor Heatup and Cooldown Rate Monitoring;
and 3-SR-3.4.9.1(2), Reactor Vessel Shell Temperature and Reactor Coolant
Pressure Monitoring During In-Service Hydrostatic or Leak Testing
Speed Stop Micro Switches, OI-82, Standby Diesel Generator System and WO
- Unit 1: PMT for Repair of HPCI Stop Valve, WO 113426235
Generator 3C 24-hour Run WO 112472092
- Unit 3: PMT for the 3-FCV-074-0048, RHR Shutdown Cooling Valve wedge
replacement performed under WO 111044044
under WO 113394369
b. Findings
No findings were identified.
Enclosure
17
1R20 Refueling and Other Outage Activities
.1 Unit 3 Scheduled Refueling Outage (U3R15)
a. Inspection Scope
During April 7 to May 26, 2012, the inspectors examined critical outage activities to verify
that they were conducted in accordance with technical specifications, applicable
procedures, and the licensees outage risk assessment and management plans through
the end of the reporting period. Some of the more significant inspection activities
conducted by the inspectors were as follows:
Outage Risk Assessment
Prior to the Unit 3 scheduled 30 day U3C15 refueling outage that began on April 7, 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 U3C15 Refueling Outage Reports, including the
Outage Risk Assessment Management (ORAM) Safety Function Status, and regularly
attended the twice a day outage status meetings. These reviews were compared to the
requirements in licensee procedure NPG-SPP-07.2, Outage Management, and technical
specifications. 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.
Shutdown and Cooldown Process
The inspectors witnessed the shutdown and cooldown of Unit 3 in accordance with
licensee procedures OPDP-1, Conduct of Operations; 3-GOI-100-12A, Unit Shutdown
from Power Operations to Cold Shutdown and Reduction in Power During Power
Operations; and 3-SR-3.4.9.1(1), Reactor Heatup or Cooldown Rate Monitoring.
The inspectors reviewed licensee procedures 3-OI-74, Residual Heat Removal System
(RHR); 3-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 an increased
outage risk condition of Yellow 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
Enclosure
18
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 technical specifications, 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 orders)
- Verified Reactor Coolant System (RCS) inventory controls, especially during
evolutions involving operations with the potential to drain the reactor vessel
(OPDRV)
- Verified electrical systems availability and alignment
- Monitored important control room plant parameters (e.g., RCS pressure, level, flow,
and temperature) and technical specifications 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 including areas normally
inaccessible during power operations, refueling floor, torus and drywell.
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 3-GOI-100-3A, Refueling
Operations (Reactor Vessel Disassembly and Floodup). Also, on numerous occasions,
the inspectors witnessed fuel handling operations during the two Unit 3 reactor core fuel
shuffles performed in accordance with technical specifications and applicable operating
procedures. Inspectors also observed control rod unlatching and relatching for control
rod drive mechanism change-outs. In addition, the inspectors verified specific fuel
movements as delineated by the Fuel Assembly Transfer Sheets (FATF). Furthermore,
the inspectors also witnessed and performed a 100 percent core verification examination
of the video verification of the final completed reactor core.
Drywell Closeout
On May 17, 2012, the inspectors reviewed the licensees conduct of 3-GOI-200-2,
Section 5.3 Drywell Closeout, and performed an independent detailed closeout
inspection of the Unit 3 drywell.
Enclosure
19
Torus Closeout
On May 12, 2012, the inspectors reviewed the licensees conduct of procedure 3-GOI-
200-2, Section 5.4 Torus Closeout, and performed an independent detailed closeout
inspection of the Unit 3 torus (suppression pool and chamber). In addition inspectors
reviewed the Foreign Material Exclusion (FME) log for any discrepancies.
Restart Activities
The inspectors specifically conducted the following:
- Witnessed Unit 2 reactor pressure vessel head tensioning in accordance with MSI-0-
001-VSL001, Reactor Vessel Disassembly and Reassembly
- Witnessed heatup and pressurization of Unit 3 reactor pressure vessel in accordance
with 3-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor pressure
Vessel and Associated Piping, and reviewed reactor coolant heatup/pressurization
data per 3-SR-3.4.9.1(1), Reactor Heatup and Cooldown Rate Monitoring, and 3-SR-
3.4.9.1(2), Reactor Vessel Shell Temperature & Reactor Coolant Pressure
Monitoring During In-Service Leak Testing
- Reviewed Reactor Coolant Heatup/Pressurization to Rated Temperature and
Pressure per 3-SR-3.4.9.1(1), Reactor Heatup and Cooldown Rate Monitoring
- Reviewed and verified completion of selected items of 0-TI-270, Refueling Test
Program, Attachment 2, Startup Review Checklist
- Reviewed 2-SR-3.6.1.1.1(OPT-A) Primary Containment Total Leak Rate - Option A,
Revision 11
- Witnessed Unit 3 approach to criticality and power ascension per 3-GOI-100-1A, Unit
Startup, 3-SR-3.3.1.1.5, SRM and IRM Overlap Verification, and 3-GOI-100-12,
Power Maneuvering
Corrective Action Program
The inspectors reviewed PERs generated during refueling outage U3C15 and
periodically attended Corrective Action Review Board (CARB) and PER Screening
Committee (PSC) 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. This constitutes one Refueling Outage activity inspection sample.
b. Findings
No findings were identified.
Enclosure
20
1R22 Surveillance Testing
a. Inspection Scope
The inspectors witnessed portions of, and/or reviewed completed test data for the
following seven surveillance tests of risk-significant and/or safety-related systems to
verify that the tests met technical specification surveillance requirements, UFSAR
commitments, and in-service testing and licensee procedure requirements. The
inspectors review confirmed whether the testing effectively demonstrated that the SSCs
were operationally capable of performing their intended safety functions and fulfilled the
intent of the associated surveillance requirement. This activity constituted seven
Surveillance Testing inspection samples: one inservice test, three routine, two
containment isolation valve and one reactor coolant system leak detection test. .
In-Service Tests:
- 2-SI-4.4.A.1, Standby Liquid Control Pump Functional Test
Routine Surveillance Tests:
- 3-SR-3.8.1.9(3D OL), Diesel Generator 3D Emergency Load Acceptance Test with
Unit 3 Operating
- 3-SR-3.5.1.8, HPCI Main and Booster Pump Set Developed Head and Flow Rate
Test at 150 psig Reactor Pressure, Rev. 13 performed on May 16, 2012
- 3-SI-4.7.A.2.g-3/74g, Unit 3 Primary Containment Local Leak Rate Test (LLRT) RHR
Shutdown Cooling Suction: Penetration X-12
Containment Isolation Valve Tests:
- 3-SR-3.6.1.3.10(B), Primary Containment Local Leak Rate Test (LLRT) Main Steam
Line B: Penetration X-7B
- 3-SI-4.7.A.2.a-f, Primary Containment Integrated Leak Rate test (CILRT), Rev. 10
Reactor Coolant System Leak Detection Tests:
- 2-SI-4.2.E-1(B), Drywell Equipment Drain Sump Flow Integrator Calibration
b. Findings
No findings were identified.
Enclosure
21
Cornerstone: Emergency Preparedness
1EP2 Alert and Notification System Evaluation
a. Inspection Scope
The inspectors evaluated the adequacy of the licensees methods for testing the alert
and notification system in accordance with NRC Inspection Procedure 71114,
Attachment 02, Alert and Notification System (ANS) Evaluation. The applicable planning
standard, 10 CFR Part 50.47(b)(5) and its related 10 CFR Part 50, Appendix E, Section
IV.D requirements were used as reference criteria. The criteria contained in NUREG-
0654, Criteria for Preparation and Evaluation of Radiological Emergency Response
Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, were also
used as a reference.
The inspectors reviewed various documents which are listed in the Attachment. This
inspection activity satisfied one inspection sample for the alert and notification system on
a biennial basis.
b. Findings
No findings were identified.
1EP3 Emergency Preparedness Organization Staffing and Augmentation System
a. Inspection Scope
The inspectors reviewed the licensees Emergency Response Organization (ERO)
augmentation staffing requirements and process for notifying the ERO to ensure the
readiness of key staff for responding to an event and timely facility activation. The
qualification records of key position ERO personnel were reviewed to ensure all ERO
qualifications were current. A sample of problems identified from augmentation drills or
system tests performed since the last inspection was reviewed to assess the
effectiveness of corrective actions.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 03, Emergency Preparedness Organization Staffing and Augmentation
System. The applicable planning standard, 10 CFR 50.47(b)(2), and its related 10 CFR
50, Appendix E requirements were used as reference criteria.
The inspectors reviewed various documents which are listed in the Attachment. This
inspection activity satisfied one inspection sample for the ERO staffing and
augmentation system on a biennial basis.
b. Findings
No findings were identified.
Enclosure
22
1EP5 Maintenance of Emergency Preparedness
a. Inspection Scope
The inspectors reviewed the corrective actions identified through the Emergency
Preparedness program to determine the significance of the issues, the completeness
and effectiveness of corrective actions, and to determine if issues were recurring. The
licensees post-event after action reports, self-assessments, and audits were reviewed to
assess the licensees ability to be self-critical, thus avoiding complacency and
degradation of their emergency preparedness program. The inspectors toured facilities
and reviewed equipment and facility maintenance records to assess licensees
adequacy in maintaining them. In addition, the inspectors reviewed licensee procedures
and training for the evaluation of changes to the emergency plans.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 05, Maintenance of Emergency Preparedness. The applicable 10 CFR
50.47(b) planning standards and related 10 CFR 50, Appendix E requirements were
used as reference criteria.
The inspectors reviewed various documents which are listed in the Attachment. This
inspection activity satisfied one inspection sample for the Maintenance of Emergency
preparedness on a biennial basis.
b. Findings
No findings were identified.
1EP6 Drill Evaluation
a. Inspection Scope
During the report period, the inspectors observed an Emergency Preparedness (EP) drill
that contributed to the licensees Drill/Exercise Performance (DEP) and Emergency
Response Organization (ERO) performance indicator (PI) measures on June 13, 2012,
to identify any weaknesses and deficiencies in classification, notification, dose
assessment and protective action recommendation (PAR) development activities. The
inspectors observed emergency response operations in the simulated control room and
certain Emergency Response Facilities to verify that event classification and notifications
were done in accordance with EPIP-1, Emergency Classification Procedure and other
applicable Emergency Plan Implementing Procedures. The inspectors also attended the
post-drill critique to compare any inspector-observed weakness with those identified by
the licensee in order to verify whether the licensee was properly identifying weaknesses.
This inspection activity satisfied one inspection sample for the Drill Evaluation of
b. Findings
No findings were identified.
Enclosure
23
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety (OS)
2RS1 Radiological Hazard Assessment and Exposure Control
a. Inspection Scope
Radiological Hazard Assessment: The inspectors reviewed a number of radiological
surveys, including those performed for airborne areas, of locations throughout the facility
including the Unit 3 (U3) drywell, Unit 1 (U1), Unit 2 (U2), and U3 reactor buildings, the
turbine building, and the independent spent fuel storage installation (ISFSI). The
inspectors also walked down many of the same areas and select radioactive material
storage locations with a survey instrument, evaluating material condition, postings, and
radiological controls. Of specific interest was the Condensate Storage Tank area which
due to a liquid radwaste processing problem created an actual radiation area outside the
building, near on-going work. The inspectors observed jobs in radiologically risk-
significant areas including high radiation areas and areas with, or with the potential for,
airborne activity. The inspectors evaluated the surveys in relation to the identified
hazards for sufficient detail and frequency.
Instructions to Workers: During plant walk downs, the inspectors observed labeling and
radiological controls on containers of radioactive material. The inspectors also reviewed
radiation work permits (RWP) used for accessing high radiation areas and airborne
areas, verifying that appropriate work control instructions and electronic dosimeter (ED)
setpoints had been provided and to assess the communication of radiological control
requirements to workers. The inspectors reviewed selected ED dose and dose rate
alarms, to verify workers properly responded to the alarms and that the licensees review
of the events was appropriate. The inspectors observed pre-job RWP briefings and
health physics technician coverage of workers. The inspectors reviewed the various
methods being used to notify workers of changing or changed radiological conditions.
Contamination and Radioactive Material Control: The inspectors observed the release
of potentially contaminated items from the radiologically controlled area (RCA) and from
contaminated areas such as the drywell. The inspectors also reviewed the procedural
requirements for, and equipment used to perform, the radiation surveys for release of
personnel and material. During plant walk downs, the inspectors evaluated radioactive
material storage areas and containers, including satellite RCAs and the low level
radwaste facility, assessing material condition, posting/labeling, and control of
materials/areas. In addition, the inspectors reviewed the sealed source inventory and
verified labeling, storage conditions, and leak testing of selected sources. The
inspectors verified if Category 1 and 2 sealed sources had been appropriately reported
to the National Source Tracking System and physically verified the presence and
controls of these sources. The sources were verified to be physically present and in
proper working order.
Enclosure
24
Radiological Hazards Control and Work Coverage: The inspectors evaluated licensee
performance in controlling worker access to radiologically significant areas and
monitoring jobs in-progress associated with the Unit 3 refueling outage. Established
radiological controls were evaluated for selected tasks including diver area setup for
torus underwater coatings inspection and desludging activities, equipment staging for
control rod drive work, reactor water cleanup sludge sampling, and work to support the
extended power uprate for Unit 3. The inspectors evaluated the effectiveness of
radiation exposure controls, including air sampling, barrier integrity, engineering controls,
and postings through a review of both internal and external exposure results. The
inspector followed up on two minor airborne radioactivity events.
During walk downs with a radiation survey meter, the inspectors independently verified if
ambient radiological conditions were consistent with licensee performed surveys, RWPs,
and pre-job briefings; observed the adequacy of radiological controls; and observed
controls for radioactive materials stored in the spent fuel pool. ED alarm set points and
worker stay times were evaluated against area radiation survey results for drywell and
refueling floor activities.
Risk-Significant High Radiation Area and Very High Radiation Area Controls: The
inspectors discussed the controls and procedures for locked-high radiation areas
(LHRAs) and very high radiation areas (VHRAs) with health physics supervisors and the
radiation protection manager. During plant walk downs, the inspectors verified the
posting/locking of LHRA/VHRA areas.
Radiation Worker Performance and Radiation Protection Technician Proficiency The
inspectors observed radiation worker performance through direct observation, via
remote camera monitoring, and via telemetry. These jobs were performed in high
radiation, airborne, and/or contaminated areas. The inspectors also observed health
physics technicians providing field coverage of jobs and providing remote coverage.
Problem Identification & Resolution: Licensee Corrective Action Program (CAP)
documents associated with radiation monitoring and exposure control were reviewed
and assessed. This included review of selected Problem Evaluation Reports (PERs)
related to radworker and health physics technician performance. The inspectors
evaluated the licensees ability to identify, characterize, prioritize, and resolve the
identified issues in accordance with procedure NPG-SPP-3.1, Corrective Action
Program, Rev. 2. The inspectors also evaluated the scope of the licensees internal
audit program and reviewed recent assessment results. Licensee CAP documents
reviewed are listed in Section 2RS1 of the Attachment.
Radiation protection activities were evaluated against the requirements of Updated Final
Safety Analysis Report (UFSAR) Section 12; Technical Specification Sections 5.4 and
5.7; 10 Code of Federal Regulations (CFR) Parts 19 and 20; and approved licensee
procedures. Radiological control activities for ISFSI areas were evaluated against 10
CFR Part 20, 10 CFR Part 72, and TS details. Records reviewed are listed in Section
2RS1 of the Attachment.
Enclosure
25
The inspectors completed 1 sample, as described in Inspection Procedure (IP)
b. Findings
No findings were identified.
2RS6 Radioactive Gaseous and Liquid Effluent Treatment
a. Inspection Scope
Program Reviews: The inspectors reviewed the 2010 and 2011 Annual Radiological
Effluent Release Report documents for consistency with the requirements in the Offsite
Dose Calculation Manual (ODCM) and Technical Specifications. Unexpected results
were followed up to determine the cause. Radioactive effluent monitor operability issues
were discussed with plant staff. The inspectors reviewed the ODCM changes made
since the last inspection against the guidance in NUREG-1301 and RG 1.109, RG 1.21,
and RG 4.1.
Walk-Downs and Observations: The inspectors walked-down selected components of
the gaseous and liquid discharge systems to ascertain material condition, configuration
and alignment. To the extent practical, the inspectors observed the material condition of
abandoned in place liquid waste processing equipment for indications of degradation or
leakage that could constitute a possible release pathway to the environment. The
inspectors also observed the collection and analysis of gaseous effluent samples (noble
gas, iodine, particulates) from the plant stack. The inspectors walked-down portions of
the Standby Gas Treatment System, to ascertain material condition, configuration, and
alignment. In addition, the inspectors reviewed the most recent HEPA and charcoal
filtration surveillance testing results for each train of the standby gas treatment system.
Sampling and Analyses: In addition to observing collection of gaseous effluent samples
from the plant stack, the inspectors observed a chemistry technician verifying plant stack
flow rates. The results of the chemistry count rooms inter-laboratory comparison
program were reviewed and discussed with cognizant licensee personnel.
Dose Calculations: The inspectors reviewed several gas release permits, and monthly
gaseous/liquid effluent dose calculation summaries. The magnitudes of the releases
were determined to be a small fraction of the applicable limits. The inspectors reviewed
the contributions to public dose from the abnormal releases. The sites 10 CFR 61
analysis was reviewed for expected nuclide distribution from the aspects of quantifying
effluents, the treatment of hard to detect nuclides, determining appropriate calibration
nuclides for instruments and whole body counting libraries. The inspectors also
reviewed the licensees most recent Land Use Census results and changes in the
ODCM since the last inspection.
Ground Water Protection: The licensees implementation of the Industry Ground Water
Protection Initiative was reviewed for changes since the last inspection as well.
Groundwater sampling results obtained since the last inspection were reviewed.
Enclosure
26
Licensee response, evaluation, and follow-up to spills and leaks since the last inspection
were reviewed in detail.
Problem Identification and Resolution: Selected corrective action program documents
associated with the effluent monitoring and control program, including problem
evaluation reports (PERs) and audits, were reviewed and assessed. The inspectors
verified that problems were being identified at an appropriate threshold and resolved in
accordance with procedure NPG-SPP-3.1, Corrective Action Program, Rev. 2 and
Rev. 3.
Documents reviewed are listed in Section 2RS6 and 2RS7 of the report Attachment.
The inspectors completed one sample as required by inspection procedure 71124.06.
b. Findings
No findings were identified.
2RS7 Radiological Environmental Monitoring Program (REMP)
a. Inspection Scope
REMP Status and Results: The inspectors discussed changes and reviewed the ODCM
and the Annual Radiological Environmental Operating Report documents issued for
calendar year (CY) 2010 and CY 2011. The inspectors also reviewed and evaluated
REMP contract laboratory cross-check program results, and current procedural guidance
for environmental sample collection and processing. Inspectors reviewed the Annual
Radiological Effluent Release Report for CY 2010 & CY 2011 under section 2RS6.
Equipment Walk-down: The inspectors observed sample collection activities of selected
air sampling stations as specified per procedure. The inspectors observed equipment
material condition and verified operability, including verification of flow rates/total sample
volume results, for the weekly airborne particulate filter and iodine cartridge change-outs
at selected atmospheric sampling stations. The material condition and placement of
environmental thermoluminescent dosimeters and water sampling stations were verified
by direct observation at select ODCM locations. Land use census results actions for
missed samples including compensatory measures and availability of replacement
equipment were discussed with environmental technicians and knowledgeable licensee
staff. Inspectors also reviewed calibration and maintenance surveillance records for the
installed environmental air sampling stations.
Procedural guidance, program implementation, quantitative analysis sensitivities, and
environmental monitoring results were reviewed against 10 CFR Part 20; Appendix I to
10CFR Part 50; TS Sections 6.8 Procedures and Programs and 6.9, Reporting
Requirements; ODCM, Rev. 15; RG 4.15, Quality Assurance for Radiological Monitoring
Programs (Normal Operation) - Effluent Streams and the Environment; and the Branch
Technical Position, An Acceptable Radiological Environmental Monitoring Program -
1979. Documents reviewed are listed in Section 2RS7 of the Attachment.
Enclosure
27
Meteorological Monitoring Program: The inspectors walked-down the meteorological
tower and observed local data collection equipment readouts. The physical condition of
the tower and the instruments were observed and equipment operability, and
maintenance history were discussed with responsible licensee staff. The transmission of
locally generated meteorological data to the main control room operators was also
verified. The inspectors reviewed applicable tower instrumentation calibration records
for the meteorological measurements of wind speed, wind direction, and temperature,
and evaluated measurement data recovery for CY 2010 and CY 2011.
Licensee procedures and activities related to meteorological monitoring were evaluated
against: ODCM; FSAR; RG 1.23, Meteorological Monitoring Programs For Nuclear
Power Plants, and ANSI/ANS-2.5-1984, Standard for Determining Meteorological
Information at Nuclear Power Sites. Documents reviewed are listed in Section 2RS7 of
the Attachment.
Problem Identification and Resolution: The inspectors reviewed selected PERs in the
areas of environmental monitoring and meteorological monitoring. The inspectors
evaluated the licensees ability to identify, characterize, prioritize, and resolve the
identified issues in accordance with NPG-SPP 3.1, Corrective Action Program, Rev. 2.
The inspectors also evaluated the scope of the licensees internal audit program and
reviewed recent assessment results. Documents reviewed are listed in Sections 2RS6
& 2RS7 in the Attachment.
The inspectors completed one sample as required by inspection procedure 71124.07.
b. Findings
No findings were identified.
2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and
Transportation
a. Inspection Scope
Waste Processing and Characterization: During inspector walk-downs, accessible
sections of the liquid and solid radioactive waste (radwaste) processing systems were
assessed for material condition and conformance with system design diagrams.
Inspected equipment included floor drain tanks; phase separator tanks; resin and filter
packaging components; and abandoned evaporator equipment. The inspectors
discussed component function, processing system changes, and radwaste program
implementation with licensee staff.
The 2010 and 2011 Annual Radiological Effluent Release Report and radionuclide
characterizations for select waste streams from 2010, and each major waste stream
from 2012 were reviewed and discussed with radwaste staff. For cleanup waste phase
separator resin, reactor water cleanup resin, Thermex resin, and dry active waste (DAW)
the inspectors evaluated analyses for hard-to-detect nuclides, reviewed the use of
scaling factors, and examined quality assurance comparison results between licensee
Enclosure
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waste stream characterizations and outside laboratory data. Waste stream mixing and
concentration averaging methodology for resins and filters was evaluated and discussed
with radwaste staff. The inspectors also reviewed the licensees procedural guidance for
monitoring changes in waste stream isotopic mixtures.
Radwaste processing activities and equipment configuration were reviewed for
compliance with the licensees Process Control Program (PCP) and UFSAR, Chapter 9.
Waste stream characterization analyses were reviewed against regulations detailed in
10 CFR Part 20, 10 CFR Part 61, and guidance provided in the Branch Technical
Position on Waste Classification (1983). Reviewed documents are listed in Section
2RS8 of the Attachment.
Radioactive Material Storage: During walk-downs of radioactive material storage areas
in the radwaste building and outdoor low-level storage yard, the inspectors observed the
physical condition and labeling of storage containers and the posting of Radioactive
Material Areas. The inspectors also reviewed licensee procedural guidance for storage
and monitoring of radioactive material.
Radioactive material and waste storage activities were reviewed against the
requirements of 10 CFR Part 20. Reviewed documents are listed in Section 2RS8 of the
report Attachment.
Transportation: The inspectors directly observed preparation activities for shipment of a
high integrity container (HIC) of resin. The inspectors noted package markings and
placarding, performed independent dose rate measurements, and interviewed shipping
technicians regarding Department of Transportation (DOT) regulations.
Selected shipping records were reviewed for consistency with licensee procedures and
compliance with NRC and DOT regulations. The inspectors reviewed emergency
response information, DOT shipping package classification, waste classification,
radiation survey results, and evaluated whether receiving licensees were authorized to
accept the packages. Licensee procedures for opening and closing Type A shipping
containers were compared to manufacturer requirements. In addition, training records
for selected individuals currently qualified to ship radioactive material were reviewed.
Transportation program implementation was reviewed against regulations detailed in 10
CFR Part 20, 10 CFR Part 71, 49 CFR Parts 172-178, as well as the guidance provided
in NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.
Documents reviewed during the inspection are listed in Section 2RS8 of the Attachment.
Problem Identification and Resolution: The inspectors reviewed PERs in the area of
radwaste/shipping. The inspectors evaluated the licensees ability to identify and resolve
the issues in accordance with procedure NPG-SPP-3.1, Corrective Action Program, Rev.
2 and Rev. 3. The inspectors also evaluated the scope of the licensees internal audit
program and reviewed recent assessment results. Licensee corrective action program
documents reviewed are listed in Section 2RS8 of the Attachment.
Enclosure
29
The inspectors completed one sample as required by inspection procedure 71124.08.
b. Findings
.1 Failure to adequately secure radioactive shipping container contents for transport
Introduction: A self-revealing Green Non-Cited Violation (NCV) of 10 CFR 71.5,
Transportation of Licensed Material, was identified for the licensees failure to ensure
proper packaging of two DOT 7A Type A packages as required by 49 CFR 173.475(e),
Quality Control Requirements Prior To Each Shipment Of Class 7 (Radioactive)
Materials.
Description: On March 22, 2010, the licensee shipped control rod drive mechanisms
(CRDMs) to GE Hitachi Nuclear (GEH) for refurbishment in six Department of
Transportation (DOT) approved Type A boxes. Each box contained four CRDMs. In a
letter dated September 17, 2010, GEH informed the licensee that their receipt inspection
of containers 1343-S and 966-S on April 23, 2010, identified that pig shield containment
lid restraint bars designed to secure the CRDMs and pig shields in place were not
installed and were laying loose in the bottom of the container. The licensee documented
the issue in PER 236118. Licensee investigation determined that the radwaste
packaging inspector failed to follow procedural requirements and verify that the CRDMs
were properly secured within the container to prevent movement during shipping. The
inspectors reviewed the Container Certification, container closure procedure for the
CRDM boxes, licensee radioactive material shipment procedures, and engineering
documents concerning the container meeting DOT 7A requirements. The inspectors
noted that although the container closure procedure did not specifically address internal
packaging and the restraint bars, the container certification states that All contents must
be securely positioned to prevent shifting during normal conditions of transport., and
that site procedural guidance requires verification that the contents of the package have
been secured and satisfies the requirements of 10 CFR 71.87, prior to shipment.
Analysis: The failure to properly secure, or adequately block or brace the material within
a Class 7 (radioactive) materials package to prevent movement during transport prior to
shipment was determined to be a performance deficiency. Specifically, the licensee
failed to follow established site procedures and applicable documents provided by the
package vendor for package inspection and verification to ensure materials are secured
within containers. The finding was more than minor because it is associated with the
Public Radiation Safety Cornerstone, Plant Facilities/Equipment and Instrumentation
attribute, involving transportation packaging and adversely affected the cornerstone
objective to ensure adequate protection of public health and safety from exposure to
radioactive materials released into the public domain as a result of routine civilian
nuclear reactor operation. Specifically, the failure to correctly secure the package
contents to prevent movement could have resulted in damage or failure of the container
during transportation. The significance of the finding was evaluated using IMC 0612,
Appendix D, Public Radiation Safety Significance Determination Process. The issue
was evaluated using the Public Radiation Safety flowchart because it involved
radioactive material control, specifically, transportation. The finding was determined to
be of very low safety significance (Green) because it did not involve radiation limits being
Enclosure
30
exceeded, a package breach, a certificate of compliance issue, a low-level burial ground
non-conformance, or a failure to make emergency notifications.
The cause of this finding was directly related to the cross cutting aspect of Documents,
Procedures and Component Labeling in the Resources component of the Human
Performance area because the licensee did not effectively incorporate package design
specifications into their transportation program to ensure that all internal restraining
devices are correctly installed to secure the CRDM in place to prevent damage to the
transport package. H.2(c)
Enforcement: 10 CFR 71.5, Transportation of Licensed Material, required, in part, that
each licensee who transports licensed material outside the site of usage, as specified in
the NRC license, or where transport is on public highways, or who delivers licensed
material to a carrier for transport, shall comply with the applicable requirements of the
DOT regulations in 49 CFR Parts 107, 171 through 180, and 390 through 397,
appropriate to the mode of transport.
49 CFR 173.475(e), Quality Control Requirements Prior To Each Shipment Of Class 7
(Radioactive) Materials, required, in part, that before each shipment of any Class 7
(radioactive) materials package, the offeror must ensure, by examination or appropriate
tests, that each special instruction for filling, closing, and preparation of the packaging
for shipment has been followed. Licensee procedure RWTP-100, Radioactive
Material/Waste Shipments, contains package inspection and verification requirements
to ensure materials are secured within containers.
Contrary to the above, on March 22, 2010, the licensee failed to comply with the
applicable requirements of DOT regulation 49 CFR 173.475(e) for transport of licensed
material. Specifically, the licensee failed to follow Container Certification guidance, in
that the CRDMs were not properly packaged and secured inside two CRDM shipping
containers as required by licensee procedure RWTP-100. Because this violation was of
very low safety significance and it was entered into the licensees CAP (SR 570902), this
violation is being treated as an NCV, in accordance with Section 2.3.2 of the NRC
Enforcement Policy. (NCV 05000259, 260, 296/2012003-02; Failure to Properly Prepare
a DOT Type A Package for Transport)
.2 Failure to Implement DOT Type A Package Closure Requirements
Introduction: A self-revealing Green Non-Cited Violation (NCV) of 10 CFR 71.5,
Transportation of Licensed Material, was identified for the licensees failure to properly
close a DOT 7A Type A packages as required by DOT 49 CFR 173.475(f) Quality
Control Requirements Prior To Each Shipment Of Class 7 (Radioactive) Materials.
Description: On September 7, 2011, the licensee shipped a DOT approved Type A
shipping container, containing an ISP surveillance capsule, to MP Machinery and
Testing, LLC (MPM) for analysis of the contents. In a letter dated September 9, 2011,
MPM informed the licensee that upon arrival at the MPM facility the closure bolts on the
shipping container were found to be undertorqued at 30 ft-lbs torque, not 390 ft-lbs
torque as specified in the DOT Package Certification provided by MPM. The licensee
Enclosure
31
documented the issue in PER 431446. Licensee investigation determined that the ISP
surveillance capsule shipping container closure bolts did not have the correct torque
applied due to inadequate procedure guidance, unfamiliarity of the workers with the task,
and a lack of procedure use and adherence. Preparation of the surveillance capsule for
shipment occurred over several months, the Technical Instruction was revised during the
period, and the container instructions provided by the vendor were not used during
loading activities. The inspectors reviewed the DOT Package Certification, container
loading and shipping instructions, Technical Instruction for obtaining and packaging the
Reactor Vessel Test Specimens (both revisions), and the work order used to remove
and package the ISP surveillance capsule for shipment. The inspectors noted that
although detailed instructions for loading and closure of the container were provided by
the vendor, the instructions and required container closure torque values were not
included, or referenced, in the Technical Instruction or the work package.
Analysis: The failure to properly close a Class 7 (radioactive) materials package was
determined to be a performance deficiency. Specifically, the licensee failed to follow
established site procedures and applicable vendor documents for closing the package
resulting in inadequate torque of the shipping container closure bolts. The finding was
more than minor because it is associated with the Public Radiation Safety Cornerstone,
Plant Facilities/Equipment and Instrumentation attribute, involving transportation
packaging and adversely affected the cornerstone objective to ensure adequate
protection of public health and safety from exposure to radioactive materials released
into the public domain as a result of routine civilian nuclear reactor operation.
Specifically, the failure to apply the correct torque to the package closure bolts could
have resulted in incomplete sealing of the container or failure of the cover bolts during
transportation. The significance of the finding was evaluated using IMC 0612, Appendix
D, Public Radiation Safety Significance Determination Process. The issue was
evaluated using the Public Radiation Safety flowchart because it involved radioactive
material control, specifically, transportation. The finding was determined to be of very
low safety significance (Green) because it did not involve radiation limits being
exceeded, a package breach, a certificate of compliance issue, a low-level burial ground
non-conformance, or a failure to make emergency notifications
The cause of this finding was directly related to the cross cutting aspect of Documents,
Procedures and Component Labeling in the Resources component of the Human
Performance area because the licensee did not effectively incorporate the vendor
provided container loading and shipping instructions into their work package and
transportation program to ensure correct torque values were used to close the shipping
container. H.2(c)
Enforcement: 10 CFR 71.5, Transportation of Licensed Material, required, in part, that
each licensee who transports licensed material outside the site of usage, as specified in
the NRC license, or where transport is on public highways, or who delivers licensed
material to a carrier for transport, shall comply with the applicable requirements of the
DOT regulations in 49 CFR Parts 107, 171 through 180, and 390 through 397,
appropriate to the mode of transport.
Enclosure
32
49 CFR 173.475(f) Quality Control Requirements Prior To Each Shipment Of Class 7
(Radioactive) Materials, required, in part, that each closure, valve, or other opening of
the containment system through which the radioactive content might escape is properly
closed and sealed.
Contrary to the above, on September 7, 2011, the licensee failed to comply with the
applicable requirements of DOT regulation 49 CFR 173.475(f) for transport of licensed
material. Specifically, the licensee failed to properly close an opening in the containment
system of a Class 7 (radioactive) materials package. Because this violation was of very
low safety significance and it was entered into the licensees CAP (SR 571151), this
violation is being treated as an NCV, in accordance with Section 2.3.2 of the NRC
Enforcement Policy. (NCV 05000259, 260, 296/2012003-03; Failure to Implement DOT
Type A Package Closure Requirements)
4. OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness
4OA1 Performance Indicator (PI) Verification
Cornerstone: Mitigating Systems
.1 Safety System Functional Failures; Mitigating Systems Performance Indicator- Heat
Removal (Reactor Core Isolation Cooling)
a. Inspection Scope
The inspectors reviewed the licensees procedures and methods for compiling and
reporting the following Performance Indicators (PIs), including procedure NPG-SPP-02.2
Performance Indicator Program. The inspectors examined the licensees PI data for the
specific PIs listed below for the second quarter 2011 through first quarter of 2012. The
inspectors reviewed the licensees data and graphical representations as reported to the
NRC to verify that the data was correctly reported. The inspectors also validated this
data against relevant licensee records (e.g., PERs, Daily Operator Logs, Plan of the
Day, Licensee Event Reports, etc.), and assessed any reported problems regarding
implementation of the PI program. Furthermore, the inspectors met with responsible
plant personnel to discuss and go over licensee records to verify that the PI data was
appropriately captured, calculated correctly, and discrepancies resolved. The inspectors
also used the Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment
Performance Indicator Guideline, to ensure that industry reporting guidelines were
appropriately applied. This activity constituted six mitigating systems performance
indicator inspection samples.
- Unit 1 Safety System Functional Failures
- Unit 2 Safety System Functional Failures
- Unit 3 Safety System Functional Failures
Enclosure
33
- Unit 1 Mitigating Systems Performance Index - Reactor Core Isolation Cooling
- Unit 2 Mitigating Systems Performance Index - Reactor Core Isolation Cooling
- Unit 3 Mitigating Systems Performance Index - Reactor Core Isolation Cooling
4OA1 Performance Indicator (PI) Verification
Cornerstone: Barrier Integrity
a. Inspection Scope
The inspectors reviewed the licensees procedures and methods for compiling and
reporting the Performance Indicators (PI) listed below, including procedure SPP-3.4,
Performance Indicator for NRC Reactor Oversight Process for Compiling and Reporting
PIs to the NRC. The inspectors reviewed the raw data for the PITs listed below for the
1st through 4th quarters of 2006. The inspectors compared the licensees raw data
against graphical representations and specific values reported to the NRC in the 4th
quarter 2006 PI report to verify that the data was correctly reflected in the report. The
inspectors also reviewed the past history of PERs for any that might be relevant to
problems with the PI program. Furthermore, the inspectors met with responsible
chemistry and engineering personnel to discuss and go over licensee records to verify
that the PI data was appropriately captured, calculated correctly, and discrepancies
resolved. The inspectors reviewed Nuclear Energy Institute 99-02, Regulatory
Assessment Performance Indicator Guideline, to verify that industry reporting guidelines
were applied.
- RCS Activity for Units 2 and 3
- RCS Leakage for Units 2 and 3
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
a. Inspection Scope
The inspectors sampled licensee submittals relative to the PIs listed below for the period
October 1, 2011, and March 31, 2012. To verify the accuracy of the PI data reported
during that period, PI definitions and guidance contained in NEI 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 6, were used to confirm the
reporting basis for each data element.
- Emergency Response Organization (ERO) Drill/Exercise Performance
- ERO Drill Participation
- Alert and Notification System Reliability
Enclosure
34
For the specified review period, the inspector examined data reported to the NRC,
procedural guidance for reporting PI information, and records used by the licensee to
identify potential PI occurrences. The inspectors verified the accuracy of the PI for ERO
drill and exercise performance through review of a sample of drill and event records.
The inspectors reviewed selected training records to verify the accuracy of the PI for
ERO drill participation for personnel assigned to key positions in the ERO. The
inspectors verified the accuracy of the PI for alert and notification system reliability
through review of a sample of the licensees records of periodic system tests. The
inspectors also interviewed the licensee personnel who were responsible for collecting
and evaluating the PI data. Licensee procedures, records, and other documents
reviewed within this inspection area are listed in the Attachment. This inspection
satisfied three Emergency Preparedness inspection samples for PI verification on an
annual basis.
b. Findings
No findings were identified.
Cornerstone: Occupational Radiation Safety
a Inspection Scope
The inspectors reviewed Performance Indicator (PI) data collected from January 1,
2011, through March 31, 2012, for the Occupational Exposure Control Effectiveness PI.
For the reviewed period, the inspectors assessed CAP records to determine whether
high radiation area, VHRA, or unplanned exposures, resulting in TS or 10 CFR 20 non-
conformances, had occurred during the review period. In addition, the inspectors
reviewed selected personnel contamination event data, internal dose assessment
results, and ED alarms for cumulative doses and/or dose rates exceeding established
set-points. The reviewed data were assessed against guidance contained in Nuclear
Energy Institute 99-02, "Regulatory Assessment Indicator Guideline," Rev. 6. The
reviewed documents relative to these PI reviews are listed in Sections 2RS1 and 4OA1
of the Attachment.
b. Findings
No findings were identified.
Public Radiation Safety (PS) Cornerstone
The inspectors reviewed the Radiological Effluent Technical Specification/Offsite Dose
Calculation Manual Radiological Effluent Occurrences PI results from June 18, 2010
through May 2012. The inspectors reviewed PERs, liquid and gaseous effluent release
permits, effluent dose data, and licensee procedural guidance for classifying and
reporting PI events. Reviewed documents are listed in Sections 2RS6 of the
Attachment.
The inspectors completed 1 of the required samples for IP 71151.
Enclosure
35
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 - Operations with a Potential for Draining the
Reactor Vessel (OPDRVs)
a. Inspection Scope
The inspectors reviewed the licensees response to the NRCs EMG-11-03, 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 (OPDRVs). The inspectors
focused on the changes made to licensee procedure 3-POI-200.5, Operations with
Potential for Draining the Reactor Vessel/Cavity and discussed OPDRVs with
Operations staff. The inspectors reviewed the Main Control Room (MCR) operating logs
to verify OPDRVs were identified by the MCR operating crew and appropriate action
taken were necessary. The inspectors also walked down portions of the alternate
reactor water level control make-up and let-down line line-ups to verify they were
established in accordance with the licensees procedures. Documents reviewed are
listed in the Attachment. This activity constituted one in-depth selected issue.
b. Assessment and Observations
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 January 2012 through June 2012, although some searches expanded beyond
Enclosure
36
these dates. Additionally, the inspectors review also included the Integrated Trend
Reports (ITR) from the first and second quarters of fiscal year 2012. The licensee
reports covered the period of October 1, 2011, to March 31, 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. This activity constituted one
semiannual trend review inspection sample.
b. Findings and Observations
No findings were identified, but the inspectors identified a number of observations as
discussed below.
Inspectors observed licensee-identified issues and trends in both the first and second
quarter ITRs that were identical or similar in nature. Inspectors reviewed the repeat
issues to assess the licensees progress of corrective actions associated with the issues
and trends identified. Some of the more notable site/departmental issues were as
follows:
business function by the station. Improvement is needed with problem identification,
cause evaluations and timely completion of corrective actions. This issue was
documented in PERs 346645 and 471366.
- Human Performance/Standards: Human performance practices resulted in
consequential events, specifically: procedure use and adherence, procedure quality,
accountability, human performance fundamentals, and the observation program.
This issue was documented in PERs 410308 and 491985.
- Procedure Use and Adherence: The first quarter 2012 ITR included this in the
Human Performance area (Issue #2) and developed actions to drive rigorous use of
procedures throughout all organization. The second quarter 2012 ITR included this
with the Procedure/Work Order Quality/Procedure Use and Adherence area (Issue
- 2). This issue was documented in PERs 410308 and 491985.
The second quarter ITR contained fifteen fundamental problem statements that were
developed as a result of the 95003 supplemental inspection. The process is intended to
determine the root organizational and/or cultural causes of these issues. Corrective
actions were under development for these fifteen problem areas at the end of the
reporting period.
Enclosure
37
The inspectors conducted an independent review of the licensees CAP to identify
potential adverse trends. The inspectors identified a potential adverse trend with the
licensees control of transient combustible materials in plant areas. A review of PERs
from January 2012 to June 2012 revealed twelve PERs associated with transient and
excessive combustible materials in plant areas however, a PER that identified this as a
trend was not identified by the licensee staff. The inspectors discussed this issue with
the appropriate licensee staff and PER 577382 was initiated to document this as an
adverse trend.
4OA3 Event Follow-up
.1 Unit 3 Automatic Reactor Scram Following Refueling Outage
a. Inspection Scope
On May 22, 2012, while recovering from a refueling outage with control rod and main
turbine generator off-line testing in progress, Unit 3 automatically scrammed from 19.5
percent power. Unit 3 scrammed due to a loss of offsite power when an inadvertent
actuation of 3A Unit Station Service Transformer (USST) differential relay 387SA
resulted from an incorrect relay setting. Inspectors promptly responded to the control
room and verified that the unit was stable in Mode 3 (Hot Shutdown), and confirmed that
all safety-related mitigating systems had operated properly. Inspectors evaluated safety
equipment and operator performance before and after the event by examining existing
plant parameters, strip charts, plant computer historical data displays, operator logs, and
the critical parameter trend charts used for the post-trip report. Inspectors also
interviewed responsible on-shift operations personnel, examined the implementation of
the applicable annunciator response procedures and abnormal operating instructions,
including 3-AOI-100-1, Reactor Scram, and reviewed the written notification made in
accordance with 10 CFR 50.72. Inspectors discussed the preliminary cause of the
incorrect relay setting with responsible Operations and Engineering personnel and
monitored Plant Oversight Review Committee (PORC) event review and restart
meetings. This review included only initial event follow-up.
b. Findings
No findings were identified.
.2 Unit 3 Manual Reactor Scram Following Refueling Outage
a. Inspection Scope
On May 24, 2012, Unit 3 was manually scrammed from Mode 2 (less than 1% rated
power) when operators ranged down the Intermediate Range Monitor (IRM) 'H'
instrument, instead of up, resulting in half scram on Reactor Protection System (RPS) 'B'
trip system. The half scram was being reset after IRM 'H' was properly ranged. As the
operator adjusted the reset scram switch, a spike on IRM 'A' was received on the RPS
'A' trip system, resulting in a partial rod insertion. When the operator identified multiple
Enclosure
38
rods inserting, the actions of the Reactor Scram Procedure, 3-AOI-l00-1, were followed
and a manual scram was inserted. The inspectors evaluated safety equipment and
operator performance before and after the event by examining existing plant parameters,
strip charts, plant computer historical data displays, operator logs, the alarm typewriter
Sequence of Events printout, and the critical parameter trend charts in the post-trip
report. The inspectors interviewed responsible on-shift Operations personnel, examined
the implementation of annunciator response and abnormal operating procedures,
(including 3-AOI-100-1, Reactor Scram) and reviewed the written notification made in
accordance with 10 CFR 50.72. This review included only initial event follow up.
b. Findings
No findings were identified
.3 Unit 3 Automatic Reactor Scram and Forced Outage
a. Inspection Scope
On May 29, 2012, Unit 3 automatically scrammed from 78 percent power due to a power
to load unbalance (i.e., main generator load reject) automatic trip of the main turbine
generator from an A-B phase trip of the main transformer differential relay 387T. The
licensee identified the cause of the differential relay trip to be a B phase current
transformer manufactured and installed with opposite polarity. Preliminarily, the licensee
revealed that factory acceptance and field testing failed to detect the manufacturing
defect of reverse polarity. Inspectors promptly responded to the control room and
verified that the unit was stable in Mode 3 (Hot Shutdown), and confirmed that all safety-
related mitigating systems had operated properly. Inspectors evaluated safety
equipment and operator performance before and after the event by examining existing
plant parameters, strip charts, plant computer historical data displays, operator logs, and
the critical parameter trend charts used for the post-trip report. Inspectors also
interviewed responsible on-shift operations personnel, examined the implementation of
the applicable annunciator response procedures and abnormal operating instructions,
including 3-AOI-100-1, Reactor Scram, and reviewed the written notification made in
accordance with 10 CFR 50.72. Inspectors discussed the preliminary cause of the failed
acceptance and installation testing with responsible Operations and Engineering
personnel. This review included only initial event follow-up.
Operators commenced restart of Unit 3 (i.e., entered Mode 2) on June 2 and achieved
full power on June 6, 2011. During this short forced outage the inspectors examined the
conduct of critical outage activities pursuant to technical specifications, applicable
procedures, and the licensees risk assessment and maintenance plans. Some of the
more significant outage activities monitored, examined and/or reviewed by the
inspectors were as follows:
- Plant Oversight Review Committee (PORC) event review and restart meetings.
- Reactor startup and power ascension activities per 3-GOI-100-1A, Unit Startup
- Reactor vessel and coolant heatup per 3-SR-3.4.9.1(1), Reactor Heatup and
Cooldown Rate Monitoring
Enclosure
39
- Outage risk assessment and management
- Control and management of forced outage and emergent work activities
Corrective Action Program
The inspectors reviewed PERs generated during the Unit 3 forced outage and attended
management review committee meetings to verify that initiation thresholds, priorities,
mode holds, and significance levels were assigned as required.
b. Findings
No findings were identified
.4 (Closed) Licensee Event Report (LER) 05000296/2011-003-00, Automatic Reactor
Scram Due to a Main Turbine Generator Load Reject.
a. Inspection Scope
On September 28, 2011, Unit 3 automatically scrammed from 100 percent power due to
a power to load unbalance (i.e., main generator load reject) automatic trip of the main
turbine generator (MTG) caused by a broken debris screen. The initial follow-up of this
event by the inspectors was documented in Section 4OA3.10 of IR 05000296/2011004.
The inspectors reviewed the applicable LER that was issued on November 28, 2011,
and its associated PER 440539, which included the root cause analysis (RCA) and
corrective actions. The licensee concluded that the direct cause of the Unit 3 turbine trip
and scram was the isolated-phase bus C debris screen failure.
b. Findings
No findings were identified
.5 (Closed) Licensee Event Report (LER) 05000259,296 /2011-009-02, As-Found
Undervoltage Trip for the Reactor Protection System 1A1 Relay that Did Not Meet
Acceptance Criteria During Several Surveillances
a. Inspection Scope
The inspectors reviewed Revision 2 of LER 05000259/2011-009 dated April 25, 2012,
PER 486780, and the associated operability determination, and corrective action plans.
This revised LER was submitted to provide the results of the licensees completed
investigation and evaluation of a second Reactor Protection System (RPS) relay that did
not meet its acceptance criteria during previous surveillance testing for the same reason.
The original LER 05000259/2011-009-00 dated December 5, 2011, the revised LER
05000259/2011-009-01 dated January 31, 2012, applicable PERs 413140 and 442914,
including root cause analysis, operability determination and corrective action plans, were
reviewed by the inspectors and documented in Sections 4OA3.1 and 4OA7 of NRC IR
Enclosure
40
05000259/2012002. As a result of this prior review, the licensee had identified one
violation of NRC requirements associated with Unit 1 RPS 1A1 relay.
On January 6, 2012, while performing an operability determination for the Unit 3 reactor
protection system (RPS) 3C1 relay undervoltage trips, the licensee determined that the
as-found undervoltage trip setpoint for the Unit 3 relay was less than the required
acceptance criteria during several technical specification surveillances. Seven of the
last thirteen surveillance test results were below the technical specification acceptance
criteria. Therefore, based on performance history, the RPS 3C1 relay was determined to
be inoperable from June 9, 2006, to February 2, 2012, when the relay was replaced.
The licensee determined the previous root cause and corrective actions were applicable
in that the surveillance test program did not require past operability reviews when out of
calibration technical specification conditions were corrected during surveillances.
The inspectors reviewed the second LER revision and verified that the supplemental
information provided in the LER was complete and accurate and that the information
was not of a significant nature to warrant any change to the original LER finding.
This licensee identified violation constitutes an additional example as documented in
NRC IR 05000259/2012002 and is not an individual non-cited violation. Further
corrective actions for this additional example are expected to be taken in conjunction
with corrective actions for the previous violation.
b. Findings
One finding for the original and Revision 1 of the LER was previously identified in
Section 4OA7 of NRC IR 05000259/2012002. No additional findings were identified.
The revised LER is considered closed.
.6 (Closed) Licensee Event Report (LER) 05000296/2012-001-00, Annunciator Panel
Power Supply Fire in Unit 3 Control Room
a. Inspection Scope
On January 26, 2012, Unit 3 main control room operators smelled smoke and observed
a flame coming from the bottom of an annunciator panel 3-XA-55-5A power supply. Fire
Operations personnel arrived on the scene within five minutes. The affected circuit
breaker was opened and fire extinguished within ten minutes. Operations personnel
increased plant monitoring to compensate for indications that lost their alarming
functions when the circuit breaker was opened. The fire damage was limited to the
failed annunciator power supply and the power supply directly above it. The inspectors
reviewed the details surrounding this event, interviewed operations and engineering
personnel involved with this issue and reviewed the licensees apparent cause
determination report. This was captured in the licensees corrective action program as
problem event report (PER) 496592. This LER is closed.
Enclosure
41
b. Findings
Introduction: A self-revealing Green finding (FIN) was identified for the licensees failure
to perform preventive maintenance on the Unit 3 Main Control Room (MCR) annunciator
power supplies. As a result, a power supply failed which led to a fire in annunciator
panel 3-XA-55-5A in the Unit 3 MCR.
Description: On January 26, 2012, Unit 3 main control room operators smelled smoke
and observed a flame coming from the bottom of an annunciator panel power supply.
Within ten minutes, the Fire Brigade responded to the MCR and the circuit breaker was
opened for the affected power supply which extinguished the fire. Damage was confined
to two power supplies in annunciator panel 3-XA-55-5A. The damaged power supplies
were replaced on January 27, 2012 in accordance with Work Order (WO) 113155456.
Corrective action document PER 496592 identified the direct cause of the annunciator
power supply failure as an overcurrent condition caused by a failed electrolytic capacitor.
This PER referenced EPRI recommendations to change out components with electrolytic
capacitors on a time based frequency. TVAs apparent cause concluded the power
supply (capacitor), installed for thirty four (34) years, experienced an age related failure
due to a lack of preventive maintenance.
Age-related failures of electrolytic capacitors have been documented in the industry.
Electric Power Research Institute (EPRI) document, TR-112175, Capacitor Application
and Maintenance Guide, dated August 1999, stated that capacitor change outs are
performed between 7 and 15 years depending on vendor recommendations and plant
operating experience. Another EPRI document, Power Supply Maintenance and
Application Guide (1003096), dated December 2001, stated that many of the power
supplies that failed had been in service greater than 15 years on average. Since 2008
three PERs have been entered in TVAs CAP that document similar failures of these
annunciator power supplies on both Unit 2 and 3 main control room panels. PER
391479 was initiated in June 2011 to evaluate the equipment reliability classification of
these power supplies. Corrective actions to evaluate the annunciator power supply
preventive maintenance strategy were in progress when the fire occurred.
These power supplies were classified as Quality-Related, Non-Critical, Low Duty-Cycle,
Mild Service Condition in accordance with licensee procedure NPG-SPP-09.18.2,
Equipment Reliability Classification. Licensee procedure TVA-NQA-PLN89-A, Nuclear
Quality Assurance Plan stated that the nuclear maintenance program including
corrective and preventive maintenance shall ensure that quality-related structures,
systems and components are maintained at a level sufficient to perform their intended
functions.
Analysis: The failure to perform preventive maintenance on the Unit 3 annunciator
power supplies prior to their age related failure was a performance deficiency.
Specifically, TVA procedure TVA-NQA-PLN89-A, Nuclear Quality Assurance Plan stated
that the nuclear maintenance program including corrective and preventive maintenance
shall ensure that quality-related structures, systems and components are maintained at
a level sufficient to perform their intended functions. These power supplies were
classified as Quality-Related according to TVA procedure NPG-SPP-09.18.2, Equipment
Enclosure
42
Reliability Classification. As a result of the performance deficiency, a Unit 3 MCR
annunciator power supply was left in service for 34 years, failed due to an aged
electrolytic capacitor and resulted in an over-current related fire. The performance
deficiency was determined to be more than minor because it was considered sufficiently
similar to example 4.f of Inspection Manual Chapter (IMC) 0612, Appendix E, for an
issue that resulted in a fire hazard in a safety-related area of the plant. The finding was
associated with the Initiating Events Cornerstone and initially characterized according to
IMC 0609, Significance Determination Process (SDP), Attachment 4, Phase 1 - Initial
Screening and Characterization of Findings. The results of this analysis required a
phase 3 evaluation in accordance with IMC 0609 because the finding increased the
likelihood of and actually caused a fire in the Unit 3 MCR. The regional Senior Reactor
Analyst performed a Phase 3 analysis for the issue. Pictures were provided to an NRC
contractor who provides expertise in fire damage for the agency. It was determined that
the configuration of the fire would not likely result in damage to anything of significance
because the metal box that the annunciators power supplies are located in, would
prevent propagation of the fire beyond the box. It is also unlikely that enough heat or
smoke could be created to require control room evacuation, which would impact the
human actions that would be performed to shut down the plant. Without an impact to
additional plant equipment, or a major impact on human action failure rates, the finding
was determined to be Green. The cause of this finding was related to the cross cutting
aspect of Problem Identification in the Corrective Action Program component of the
Problem Identification and Resolution area, because the licensee was aware of three
previous failures of these power supplies in July 2009 and should have recognized that
the electrolytic capacitors, installed beyond their recommended service life, required
replacement prior to failure P.1(a).
Enforcement: Enforcement action does not apply because the performance deficiency
did not involve a violation of regulatory requirements since the main control room
annunciator power supplies were not safety-related. Because the finding does not
involve a violation, was entered into the licensees corrective action program as PER
496592, and has very low safety significance, it is identified as FIN 05000296/2012003-
04, Failure to Perform Preventive Maintenance on the Unit 3 Main Control Room
Annunciator Power Supplies.
4OA6 Meetings, Including Exit
.1 Exit Meeting Summary
On April 13, 2012, regional inspectors presented the results of the Occupational
Radiation Safety inspection to Mr. P. Summers, Director Safety and Licensing, and other
members of the licensees staff.
On April 20, 2012, regional inspectors presented the results of the Unit 3 Inservice
Inspection to members of the licensees staff.
On June 22, 2012, regional inspectors presented the results of the Public Radiation
Safety inspection to Mr. K. Polson, Site Vice President, and other members of the
licensees staff, who acknowledged the findings. On July 03, 2012, regional inspectors
Enclosure
43
presented changes to the inspection results via telephone to Mr. S. Bono, General
Manager Site Operations, and other members of the licensees staff, who acknowledged
the changes.
On June 29, 2012, regional inspectors presented the results of the Emergency
Preparedness inspection to Mr. S. Bono, General Manager Site Operations, and other
members of the licensees staff.
On July 10, August 10 and 14th, 2012, the resident inspectors presented the results of
the quarterly integrated onsite inspection to Mr. K. Polson, Site Vice President, 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 violation of very low safety significance (Green) was identified by the
licensee and is a violation of NRC requirements which met the criteria of the NRC
Enforcement Policy, for being dispositioned as a Non-Cited Violation:
- A violation of Technical Specification 5.4.1.a was identified by the licensee for the
failure to establish adequate work instructions to ensure proper installation of the gap
setting between the actuator stem and valve stem of Unit 1 HPCI, (High Pressure
Coolant Injection), turbine stop valve, 1-FCV-073-0018. On April 19, 2012, during
the performance of a quarterly surveillance test the turbine stop valve, 1-FCV-073-
0018, failed to close upon repeated demands. A Phase 3 analysis determined the
significance of the finding was very low safety significance (Green) The regional
Senior Reactor Analyst performed a Phase 3 SDP analysis on the finding. The risk
was dominated by the unavailability of the HPCI during the repair time after
discovery of the Stop Valve issue. The finding was determined to be GREEN in the
SDP, primarily due to the short period of time it was fully non-functional. The
licensee initiated PER 539040 to enter the issue into their corrective action program.
Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
T. Adkins, Manager EP Systems
S. Bono, Plant General Manager Site Operations
C. Boschet, QA Manager
J. Boyer, Acting Assistant Director of Engineering
B. Bruce, Acting Systems Engineering Manager
D. Campbell, SM
S. Clement, Operations Fire Protection
M. Durr, Director of Engineering
M. Ellet, Maintenance Rule Coordinator
J. Emens, Nuclear Site Licensing Manager
A. Feltman, Emergency Preparedness Manager
J. Ferguson, Radiation Protection Support Superintendent
C. Gannon, Plant Manager
H. Higgins, Acting Licensed Operator Requalification Supervisor
D. Hughes, Operations Manager
S. Kelly, Work Control Manager
D. Kettering, Electrical Systems Engineering Manager
J. Kimberlin, FIN Manager
R. King, Design Engineering Manager
W. Lee, Corporate EP Manager
R. Norris, Radiation Protection Manager
S. Norris, Engineering Supervisor
P. Parker, Site Security Manager
J. Parshall, Manager, EP Program Planning and Implementation
K. Polson, Site Vice President
E. Quidley, EDG Project Manager
M. Rasmussen, Operations Superintendent
H. Smith, Fire Protection Supervisor
R. Stowe, Equipment Reliability Manager
P. Summers, Director of Safety and Licensing
J. Underwood, Chemistry Manager
C. Vaughn, Operations Superintendent
S. Walton, Electrical Maintenance Superintendent
M. Wilson, Director of Training
A. Yarbrough, BOP System Engineering Supervisor
Attachment
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened and Closed
05000259,260,296/2012-003-01 NCV Failure to Maintain Flood Barrier Results in
Inoperable Safety Related Pumps (Section 1R15.)
05000259,260,296/2012003-02 NCV Failure to Properly Prepare a DOT Type A Package
for Transport) (Section 2RS8)
05000259,260,296/2012003-03; NCV Failure to Implement DOT Type A Package Closure
Requirements) (Section 2RS8)
05000260,296/2012003-04 FIN Failure to Establish Preventive Maintenance for
Unit 2 and 3 Main Control Room Annunciator
Power Supplies (Section 4OA3.6)
Closed
05000296/2011-003-00 LER Automatic Reactor Scram Due to a Main Turbine
Generator Load Reject (Section 4OA3.4)
05000259,296/2011-009-02 LER As-Found Undervoltage Trip for the Reactor
Protection System 1A1 Relay that Did Not Meet
Acceptance Criteria During Several Surveillances
(Section 4OA3.5)
05000296/2012-001-00 LER Annunciator Panel Power Supply Fire in Unit 3
Control Room (Section 4OA3.6)
Discussed
None
Attachment
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
0-GOI-300-4, Switchyard Manual, Rev. 85
0-OI-30F, Common DG Building Ventilation, Rev. 30
0-OI-30F/ATT-1, Attachment 1 Valve Lineup Checklist, Rev. 28
0-OI-30F/ATT-1A, Attachment 1A Valve Lineup Checklist Unit 3, Rev. 28
0-OI-30F/ATT-2, Attachment 2 Panel Lineup Checklist, Rev. 29
LCEI-CI-C9, Procedure for Walkdown of Structures for Maintenance Rule, Rev. 5
NPG-SPP-10.2, Clearance Procedure to Safely Control Energy, Rev. 3
OPDP-2, Switchyard Access and Switching Order Execution, Rev. 6
PER 390201, Concrete Piers in Switchyard Showing Signs of Degradation
PER 534276, Conflicting information on 161-kv grid status during U3R15 outage
PER 536136, U3 Transformer Project Material Storage Area Poses U2 Concern
PER 538016, Intake has no working ventilation fans
PER 539365, Switchyard Deficiencies
PER 539371, 500kV and 161kV Concrete Pedestals
PER 539580, Transformer Yard Discrepancies
PER 539581, Ground Soft in Transformer Yard
PER 539582, Concrete Pedestal Degraded in Transformer Yard
PER 539583, Transformer Yard 500kV Tower Damaged
PER 546871, Hot Weather procedure
PER 566119, Freeze protection heater still in place
PER 568461, Hot weather procedure
PSO PER 546093, Transformer Yard 500 kV P.O. Structure Damage
TRO-TO-SPP-30-128, Browns Ferry Nuclear Plant Grid Operating Guide, Rev. 13
TVA-SPP-10.010, NERC Standard Compliance Processes Shared by TVA's Nuclear Power and
Energy Delivery Organizations, Rev. 0
UFSAR-8.4, Normal auxiliary Power System, Amendment 23
WO 113419591, Hand switch stuck in slow position
WO110926526, Plant air wash pump
Section 1R04: Equipment Alignment
0-47E861-1, Flow & Control Diagram Diesel Starting Air System Diesel Generator A, Rev. 17
0-OI-82/ATT-1A, Standby Diesel Generator A, Valve Lineup Checklist, Rev. 100
0-OI-82/ATT-2A, Standby Diesel Generator A, Panel Lineup Checklist, Rev. 100
0-OI-82/ATT-3A, Standby Diesel Generator A, Electrical Lineup Checklist, Rev. 100
0-OI-82/ATT-4A, Standby Diesel Generator A, Instrument Inspection Checklist, Rev. 101
1-OI-71, Reactor Core Isolation Cooling System, Rev. 14
1-OI-71/ATT-1, RCIC System, Valve Lineup Checklist, Rev. 13
1-OI-71/ATT-2, RCIC System, Panel Lineup Checklist, Rev. 13
1-OI-71/ATT-3, RCIC System, Electrical Lineup Checklist, Rev. 13
3-OI-74, Residual Heat Removal System, Revision 0104
3-OI-74/ATT-1, Valve Lineup Checklist Unit 3, Revision 0086
3-OI-74/ATT-2, Panel Lineup Checklist, Revision 0086
3-OI-74/ATT-3, Electrical Lineup Checklist, Revision 0087
DWG 1-47E813-1, Flow Diagram RCIC System, Rev. 33
Attachment
4
Technical Requirements Manual Section 3.5.3, Equipment Area Coolers
Technical Requirements Manual Section 3.5.4, Maintenance of Filled Discharge Piping
Updated Final Safety Report Section 4.8, Residual Heat Removal System
Section 1R05: Fire Protection
0-SI-4.11.E.1.B(2), Safety Related Fire Hose Replacement, Rev 08
0-SI-4.11.E.1.B(2)/ATT-1, Attachment 1 Fire Hose Replacement Data Sheet, Rev. 08
0-TI-470, Temporary Wiring And Electrical Equipment (600 Volt Or Less), Rev. 1
Active FPIPs dated 5/1/2012
Active FPIPs List, 06/01/2012
DWG 0-47W216-51, Fire Protection - 10 CFR 50 Appendix R, Fire Area Compartmentation and
Zone Drawings, Rev. 7
DWG 0-47W216-56, Fire Protection - 10 CFR 50 Appendix R, Fire Area Compartmentation and
Zone Drawings, Plan EL 593.0 & 586.0, Rev. 7
Fire Hazard Analysis Fire Zone 3-3
Fire Protection Report Vol. 1, Fire Hazards Analysis, Rev. 11
Fire Protection Report Vol. 2, Rev. 48
Fire Protection Report, Volume 1, Section 2, Fire Hazards Analysis, Rev. 11
Fire Protection Report, Volume 2,Section IV, Pre-Plan No. RX2-519 Torus Area and HPCI
Room
Fire Protection Report, Volume 2,Section IV, Pre-Plan No. RX2-519 NW
Fire Protection Report, Volume 2,Section IV, Pre-Plan No. RX2-519 SW
Fire Protection Report, Volume 2,Section IV, Pre-Plan No. RX2-565
FP-0-000-INS001(A), Inspection of Portable and Wheel Type Fire Extinguisher Stations
(Reactor Building), Rev. 17
FP-0-000-INS001(A)/ATT-2, Attachment 2 Inspection Check/Data Sheet Dry Chemical (12 yrs)
Co2 (5 yrs) Halon (12 yrs) Charging Cylinder (5 yrs), Rev. 17
FP-0-000-INS012, Fire Watch Expectations, Rev. 1
FP-0-000-INS019, Fire Protection Weekly Inspection, Rev. 13
NPG-SPP-09.17, Temporary Equipment Control, Rev. 1
NPG-SPP-18.4.6, Control of Fire Protection Impairments, Rev. 0
PER 545547, Room on 1C Hallway Contain Excessive Combustibles
PER 546065, Multiple Extension Cords Plugged Into One Another on 1C Hallway
PER 546188, Roving Fire Watch Route Sheet
Pre-Fire Plan for Browns Ferry Nuclear Plant - Reactor Building Unit 3, pre-plan No. RX3-593
Pre-Fire Plan for Browns Ferry Nuclear Plant - Reactor Building Unit 3, pre-plan No. RX3-565
TVA Safety Manual Chapter 2, Procedure 1004, Extension Cords and Attachments, Rev. 4
Section 1R07: Annual Heat Sink Performance
0-TI-322, RHR Heat Exchanger Performance Testing, Rev. 0
0-TI-364, ASME Section XI System Pressure Tests, Rev. 6
0-TI-389, Raw Water Fouling and Corrosion Control, Rev. 16
0-TI-522, Program for Implementing NRC Generic Letter 89-13, Rev. 1
0-TI-63, RHRSW Flow Blockage Monitoring, Rev. 25
DCN T38580A, Repair 3A and 3C RHR Heat Exchanger Flange Leaks Using Furmanite Sealing
Compound, Rev. A
DWG 0-47E452-1, Mechanical Residual Heat Removal System, Rev. 15
DWG 3-47W452-10, Mechanical Residual Heat Removal System, Rev. 15
Attachment
5
DWG 69-D-160-03, Tube Sheet Details, Rev. 6
EDC 69311A, Repair of 3B and 3D RHR Heat Exchanger Flange Leaks, Rev. A
EPRI NP-7552, Heat Exchanger Performance Monitoring Guidelines, Dec. 1991
Evaluation of Temporary Sealing Compound used as a replacement gasket, Dated 5/8/2012
MCI-0-000-LKS001, On-Line leak Sealing, Rev. 15
MCI-0-074-HEX001, Maintenance of RHR Heat Exchangers, Rev. 23
NPG-SPP-09.7, Corrosion Control Program, Rev. 2
N-VT-4, System Pressure Test Visual Examination Procedure, Rev. 23
P.S. 4.M.4.3 (R4), General Engineering Specification, G-29B, Online Leak Sealing, Rev. 4
PER 543035, Temporary Furmanite repairs on RHR HX 3A, 3C, and 3D are not being tracked
PM 500103065, Inspect / Clean RHRSW Pump Pit
PM 500108601, PM Performance of 0-TI-63, RHRSW Flow Blockage Monitoring, for
1-HEX-74-900A & C.
PM 500116540, PM Performance of 0-TI-63 for 2-HEX-74-900A and 2-HEX-74-900C
PM 500116541, PM Performance of TI-63 for 2-HEX-74-900B and 2-HEX-74-900D
PM 500126928, Clean BFN-3-HEX -074-0900A Heat Exchanger
PM 500126929, PM Performance of 0-TI-63, RHRSW Flow Blockage Monitoring, for
3-HEX-74-900A & C
PM 500126931, Clean BFN-3-HEX -074-0900B Heat Exchanger
PM 500126932, PM Performance of 0-TI-63 for 3-HEX-74-900B and 3-HEX-74-900D.
PM 500126933, Disassemble, Clean, Inspect BFN-3-HEX -074-0900C
PM 500126935, Disassemble, Clean, Inspect BFN-3-HEX -074-0900D.
PM 500133228, PM Perform TI-63 for 1-HEX-74-0900B and D
WO 08-712116, Repair Leak, 3D RHR Heat Exchanger
WO 112857671, Test RHR Heat Exchanger 3A and 3C
WO 95-20541-000 (3A and 3C)
Section 1R11: Licensed Operator Requalification
2-AOI-57-5B, Loss of Instrument & Control Bus
2-AOI-70-1, Loss of Reactor Building Closed Cooling Water
2-C-5, Level/Power Control
2-EOI-1, Reactor Pressure Vessel Control
Section 1R12: Maintenance Effectiveness
0-AOI-100-3, Flood Above Elevation 558, Rev. 35
0-AOI-100-3, Flood Above Elevation 558, Rev. 35
0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting -
10CFR50.65, Rev. 37
0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
10CFR50.65, Rev. 37
Cause Determination Evaluation 1041, May 31, 2011
Design Criteria BFN-50-7023, Residual Heat Removal Service Water (RHRSW) System
Design Criteria BFN-50-7067, Emergency Equipment Cooling Water (EECW) System
Design Criteria BFN-50-C-7101, Protection from Wind, Tornado Wind, Tornado
Depressurization, Tornado Generated Missiles, and External Flooding
FSAR Section 10.10, Emergency Equipment Cooling Water (EECW) System, BFN-24
FSAR Section 10.9, RHR Service Water System, BFN-24
FSAR Section 10.9, RHR Service Water System, BFN-24
Attachment
6
FSAR Section 12.2.7.1.2, Principle Structures and Foundations, Personnel Access Doors,
BFN-24
FSAR Section 12.2.7.1.2, Principle Structures and Foundations, Personnel Access Doors,
BFN-24
FSAR Section 2.4, Hydrology, Water Quality, and Aquatic Biology, BFN-24
FSAR Section 2.4, Hydrology, Water Quality, and Aquatic Biology, BFN-24
MCI-0-023-PMP002, Maintenance of EECW and RHRSW Pumps, Rev. 52
MCI-0-023-PMP003, Emergency Equipment Cooling Water and Residual Heat Removal Service
Water Pump Removal and Installation, Rev. 12
MCI-0-023-PMP004, EECW and RHRSW Pump Impeller Adjustment, Rev. 05 and 06
MPI-0-260-DRS001, Inspection and Maintenance of Doors
NPG-SPP-03.4, Maintenance Rule Performance Indicator Monitoring, Trending and Reporting -
10CFR50.65, Rev. 0
NPG-SPP-06.10, NPG Fix It Now (FIN) Team Process, Rev. 0
NPG-SPP-07.1, On-Line Work Management, Rev. 05
PER 234151, Unit 2 IRM scram signal
PER 310544, Gaps in A and D RHRSW Pump Room Flood Doors
PER 383975, Reliability of RHRSW Pump Room Door Seals
PER 402414, IRM (a)(1) plan
PER 469640, BFN-0-DOOR-260-C-RHRSW, Aggregate Impact of RHRSW Pump Room Doors
PER 470350, BFN-0-DOOR-260-C-RHRSW May Not Seal
PER 481145, B and D RHRSW Pump Room Watertight Doors Were Found Closed and Locked,
But Not Mechanically Restrained
PER 482838, RHRSW B Pump Room Door Failed Chalk Test
PER 482867, RHRSW D Pump Room Door Failed Chalk Test
PER 524957, Review past 48 months of IRM data for MR failures.
PER 532050, NRC Identified C3 EECW Pump Foundation Hole Flood Protection Cover
Inadequately Installed
PER 546734, Lack of specified torque value for pump coupling bolts
PER 561666, NRC Walkdown Identified RHRSW Door Issues
PER 563567, Site Tolerance of Degraded/Nonconforming Issue
PER 563727, RHRSW Intake Structure Doors Have Not Been Evaluated for MR (a)(1)
PER 566123, Document Former NRC Senior Resident Observation
Plant Level Event Data from Mar. 2010 to Feb. 2012
SR 565020, Inaccurate Past Operability Due to CAP Input
SR 568840, NRC Identified - Failure to Accurately Document NRC Observations in CAP
SR 569912, Inconsistency in Flood Cover Description Between Maintenance Procedures
Technical Specification and Basis 3.7.1 Residual Heat Removal Service Water (RHRSW)
System, Amendment 234
Technical Specification and Basis 3.7.2 Emergency Equipment Cooling Water (EECW) System
and Ultimate Heat Sink (UHS), Amendment 234
U1,2,3 Maintenance Rule Data from Nov. 2009 to Feb. 2012
Units 1,2,3 System 092 (IRMs) Health Reports from 10/1/2011 to 1/31/2012
Unplanned Scram Data from Mar. 2010 to Feb. 2012
WO 111457995, Repair BFN-0-DOOR-260-C-RHRSW
WO 111835839, D RHRSW Upper Dog Catching and Missing Dog
WO 111926930, B RHRSW Dogs Lower Linkage Disconnected
WO 112744581, C3 EECW Pump Vibes in Alert, Troubleshoot and Repair
Attachment
7
WO 112972845, Impeller gap adjustment of A3 EECW pump
WO 113062982, Repair BFN-0-DOOR-260-B-RHRSW
WO 113062984, Repair BFN-0-DOOR-260-D-RHRSW
WO 113228273, Why is A RHRSW Door Locked - Door Doesnt Fully Close
WO 113348314, C RHRSW Lower Left Dragging and Scraping Metal
WO 113446620, Performance of 3-SI-4.5.C.1(2) - EECW Pump Operation
WO 113456059, Raw Cooling Water Leak on 3B CRD Pump
WO 113474206, Performance of 3-SI-4.5.C.1(2) - EECW Pump Operation
WO 113475937, D Diesel Generator came up to 500 rpm
WO 113483626, Troubleshoot BFN-0-RLY-082-D/ALM
WO 113486500, Diesel Generator D Air Pressure Alarm Relay
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
1-OI-73, High Pressure Coolant Injection System, Rev. 22
1-SR-3.3.3.1.4(G), Verification of Remote Position Indicators for HPCI System Valves, Rev. 2
1-SR-3.5.1.7, HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at Rated
Reactor Pressure, Rev. 21
BFN Unit 3 Defense in Depth Assessment May 4, 2012
BFN Unit 3 Defense in Depth Assessment, April 15, 16, 17, 18, 2012
BFN-ODM-4.18, Protected Equipment, Rev. 6
Browns Ferry Nuclear Plant Outage Risk Assessment Report, Unit 3 Cycle R15, Rev. 1
DWG 1-47E812-1, Rev. 34
DWG 68-XC-71, Schutte & Koerting Co. Manufacturing Drawing
EOOS Report, Unit 2, dated May 7, 2012
MCI-0-073-VLV001, HPCI Turbine Stop Valve - FCV 73-18 Disassembly, Inspection, Rework
and Reassembly, Revs. 12, 13
MSI-1-073-GOV001, HPCI Turbine Overspeed Trip Test, Rev. 7
NPG-SPP-7.0, Work Management
NPG-SPP-07.1, On Line Work Management, Rev. 5
NPG-SPP-07.2, Outage Management, Rev. 2
NPG-SPP-07.2.11, Shutdown Risk Management, Rev. 2
NPG-SPP-07.2.11, Shutdown Risk Management, Rev. 2
NPG-SPP-07.3, Work Activity Risk Management Process, Rev. 07
NPG-SPP-07.3, Work Activity Risk Management Process, Rev. 7
NPG-SPP-09.11, Probabilistic Risk Assessment (PRA) Program, Rev. 01
NPG-SPP-09.11.1, Equipment Out of Service (EOOS) Management, Rev. 04
NPG-SPP-7.2.11, Shutdown Risk Management, Rev. 2
ORAM Model Change Form, April 18, 2012
ORAM Sentinel Outage Safety Assessment, April 18, 2012
O-TI-367
Outage Risk Assessment Report, U3 Cycle R15, Rev. 1
PER 539040, HPCI Turbine Stop Valve Failed to Trip
PER 539556, HPCI Turbine Main Pump Vibration
PER 541156, HPCI Oil Tank Level Low
PER 541727, HPCI Gland Exhauster Pump Breaker
PER 547134, Shutdown Risk Management, Filling out DID Checklist Once per 24 Hours
PMT-0-000-MEC001, Leak Checks on Tube Fitting, Threaded, Flanged, Bolted or Welded
Connections, Rev. 7
Attachment
8
SR 541069, Adjust Sensitivity on Incipient Fire Detector
U3 ORAM Safety Function Status Report, dated May 5, 2012
WO 113426235, HPCI Turbine Stop Valve Failed to Trip
WO 113426235, HPCI Turbine Stop Valve PMT Step Text
WO 113429679, Task 10: 1-FCV-073-0018, Rev. 0
WO 113435872, HPCI Main & Booster Pump Head & Flow Rate Test
WO 113440357, HPCI Oil Tank Level Low
WO 113441055, Verification of Remote Position Indicators
WO 113445422, Adjust Sensitivity on Incipient Fire Detector
Section 1R15: Operability Evaluations
0-17W300-9, Mechanical Isometric drawing for EECW drains, Rev. 0
0-GOI-200-1, Freeze Protection Inspection, Rev. 69
0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
10CFR50.65, Rev. 37
1-47E859-1, Flow Diagram Emergency Equipment Cooling Water, Rev. 81
1-47E859-1, Flow Diagram for EECW system Unit 1, Rev. 82
2-47E859-1, Flow Diagram for EECW system Unit 2, Rev. 31
3-47E859-1, Flow Diagram for EECW system Unit 1, Rev. 38
3-SI-4.5.C.1(2), EECW Pump Operation, Rev. 119
BFN-50-7067, General Design Criteria Document for the EECW system, Rev. 18
BFN-50-C-7067, EECW System Design Criteria, Rev. 18
Calculation MDN0026910163, Combustible Load Table, Rev. 42
DCN 69957, Appendix R Pump House Tunnel Fire Barrier, Rev. A
DWG 2-47E600-53A, Mechanical Instruments and Controls, Rev. 3
EPI-0-000-FRZ001, Freeze Protection Program for RHRSW Pump Rooms and Diesel
Generator Building, Rev. 19
Fire Protection Report Volume 1, Fire Hazards Analysis for Fire Area 25, Rev. 11
FSAR Section 10.9, RHR Service Water System, BFN-24
FSAR Section 12.2.7.1.2, Principle Structures and Foundations, Personnel Access Doors,
BFN-24
FSAR Section 2.4, Hydrology, Water Quality, and Aquatic Biology, BFN-24
MPI-0-260-DRS001, Inspection and Maintenance of Doors
NPG-SPP-09.0, Engineering, Rev. 1
NPG-SPP-09.3, Plant Modifications and Engineering Change Control, Rev. 6
Past Operability Form for PER 492957, Tarps on RHRSW Rooms
PER 310544, Gaps in A and D RHRSW Pump Room Flood Doors
PER 372194, FPR Justification on Intake Pumping Station Fire Barriers
PER 469640, BFN-0-DOOR-260-C-RHRSW, Aggregate Impact of RHRSW Pump Room Doors
PER 470350, BFN-0-DOOR-260-C-RHRSW May Not Seal
PER 481145, B and D RHRSW Pump Room Watertight Doors Were Found Closed and Locked,
But Not Mechanically Restrained
PER 492957, Tarps on RHRSW Rooms
PER 500804, Immediate Actions Taken for PER 492957 Not Documented
PER 520497, EECW check valve appears to be seeping and repressurizing pipe
PIC 70445, System 26, PER 372194 Corrective Action - IPS Fire Seals, Rev. 0
Prompt Determination of Operability (PDO) for 0-CKV-067-0502, Rev. 0
Prompt Determination of Operability for PER 569282
Attachment
9
SR 482359, RHRSW B Pump Room Door Failed Chalk Test
SR 482401, RHRSW D Pump Room Door Failed Chalk Test
SR 560210, NRC Walkdown Identified RHRSW Door Issues
SR 563000, Site Tolerance of Degraded/Nonconforming Issue
SR 563507, RHRSW Intake Structure Doors Have Not Been Evaluated for MR (a)(1)
SR 565020, Document Former NRC Senior Resident Observation
WO 111457995, Repair BFN-0-DOOR-260-C-RHRSW
Section 1R18: Plant Modifications
3-ARP-9-3E, Panel 9-3, 3XA-55-3E, Rev. 26
3-GOI-100-3A, Refueling Operations (RX Vessel Disassembly and Floodup), Rev. 56
3-GOI-100-3B, Refueling Operations (RX Cavity Letdown and Vessel Re-Assembly), Rev. 50
3-SIMI-3A, Reactor Feedwater System Index, Rev. 32
ACE PER 427252(330400) Initial Cavity Flood-up Overflow into Ventilation Ducts
LCL-3-L-03-055, Reactor Water level Flood-Up Calibration, Rev. 5
Minor Mod DCN 70549, Reactor Water Level Flood-Up Transmitter and Indication Loop
Replacement, Rev. A
NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments, Rev. 5
NPG-SPP-09.5, Temporary Alterations, Rev. 2
NPG-SPP-9.3, Plant Modifications and Engineering Change Control, Rev. 6
NPG-SPP-9.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments, Rev. 5
ODMI-2012-0004, FCV-73-16 Leakage
PER 427252, Initial Cavity Flood-up Overflow into Ventilation Ducts, (PER 330400)
PER 565572, U1 HPCI Steam Admission Valve Leakage
PER 565577, U1 HPCI Steam Admission Valve Leakage
PER 569927, Opportunity for Operations Turnover Improvement
PER 571068, Potential Grease Degradation
SII-3-L-03-055, 500 Reactor Water Level A Refuel Range LT-3-55 Special Calibration for
Vented Vessel and Fuel Pool Flood-Up, Rev. 2
TACF 1-12-001-073, Thermal Insulation Attached to BFN-1-FCV-073-0016, HPCI Steam Supply
Valve, Rev. 0
TACF 2-12-001-073, Thermal Insulation Attached to BFN-2-FCV-073-0016, HPCI Steam Supply
Valve, Rev. 0
VTD-OT01-0020, OTEK Corp. Ops Manual for HI-Q Programmable Controllers, Rev. 3
WO 112971110, WO Request for DCN 70549 to Implement 3-55 Loop Modification on U3
WO 113275768, Implement TACF 1-12-001-073 to remove insulation from BFN-1-FCV-073-
0016
WO 113322598, Implement TACF 2-12-001-073 to remove insulation from BFN-2-FCV-073-
0016
Section 1R19: Post-Maintenance Testing
0-OI-82, Standby Diesel Generator System, Rev. 129
0-SR-3.8.1.1(D), Diesel Generator D Monthly Operability Test, Rev. 39
0-TI-106, General Leak Rate Test Procedure, Rev. 14, performed on April 9, 2012
0-TI-360, Containment Leak Rate Programs, Rev. 33
0-TI-362, Inservice Testing of Pumps and Valves, Rev. 29
3-45E779-41, Wiring Diagram, 480V Shutdown Auxiliary Power Schematic Diagram, Rev. 19
3-45E779-51, Wiring Diagram, 480V Load Shed Div II Schematic Diagram, Rev. 19
Attachment
10
3-47E801-1-ISI, ASME Section XI, Flow Diagram Main Steam Code Class Boundaries, Rev. 19
3-SI-3.3.1.A, ASME Section XI System Leakage Test of the Reactor Pressure Vessel and
Associated Piping, Rev. 21
3-SR-3.4.9.1(1), Reactor Heatup and Cooldown Rate Monitoring, Rev. 21
3-SR-3.4.9.1(2), Reactor Vessel Shell Temperature and Reactor Coolant Pressure Monitoring
During In-Service Hydrostatic or Leak Testing, Rev. 15
3-SR-3.6.1.3.10(B) Primary Containment Local Leak Rate Test Main Steam Line B: Penetration
X-7B
3-SR-3.6.1.3.10(B-OUTBD), Primary Containment Local Leak Rate Test Main Steam Line B
Outboard Penetration X-7B, Rev. 06, performed on April 8, 2012
3-SR-3.8.1.1(3C) Diesel Generator 3C Monthly Operability Test, Rev. 42, performed on May
15, 2012
3-SR-3.8.1.7(3C), Diesel Generator 3C 24 Hour Run, Rev. 21, performed on April 24, 2012
ECI-0-000-RLY003, Replacement of Relays, Rev. 21
EII-0-000-TCC106, Troubleshooting, Doc. and Config. Control of Elect. Activities, Rev. 62
MCI-0-000-PCK001, Generic Maintenance Instructions for Valve Packing, Rev. 26
MCI-0-074-VLV002, Residual Heat Removal Motor Operated Valves, FCV-74-47, 48, 53 and 67
Disassembly, Inspection, Rework and Reassembly
MCI-0-082-GOV001, Standby Diesel Engine Governor Removal and Installation, Rev. 9
MCR logs
MMDP-1, Maintenance Management System
MSI-0-001-VSL001, Reactor Vessel Disassembly and Reassembly, Rev. 100
NPG-SPP-06.3, Pre-/Post-Maintenance Testing
PER 143225, High Vibration on Generator end bearing on 3D DG
PER 538810, Restart NOI U3RF15-002: RPV Head Deformation due to Foreign Object
PER 541788, High Vibrations on 3C DG
PER 548753, Extent of Condition for D DG, (3A)
PER 548755, Extent of Condition for D DG, (3B)
PER 548756, Extent of Condition for D DG, (3C)
PER 548757, Extent of Condition for D DG, (3D)
PER 553585, Hydro Procedure Discrepancy
SR 532953, 3-FCV-1-27 failed as-found LLRT
SR 542421, Smooth Indication Noted on the Top Surface of RPV Flange During U3R15
SR 546885, Address 3C DG axial vibration
SR 547405, As-found LLRT rotameter did not meet required accuracy
SR 548237, Four Studs Not Pulled While Tensioning the U3 RPV Head
VTD-W290-0050, Instruction Manual for Woodward EG-B10C Governor Actuator, Rev. 2
WO 112472092, Generator Replacement Testing for 3C EDG
WO 112505164, Perform as-left LLRT for B outboard MSIV, Penetration X-7B
WO 113324169, Reassemble Generator for 3C EDG
WO 113394336, Re-torque Valve Packing on 3-FCV-001-0027 (B Outboard MSIV)
WO 113429130, 3-BKR-231-0003B/3C needs cell switch adjustment
WO 113475937, D D/G Came Up To 500 RPM When Started During 0-SR-3.8.1.1(D)
WO 113480500, D/G D Monthly Operability Test
WO 113480917, Replace D D/G Governor Speed Stop Micro Switches
WO 113483626, Troubleshoot/Repair/Replace BFN-0-RLY-082-D/ALM
WO 113483967, D D/G Dryer Assembly High DP Causing Excessive Blow Down
WO 113484062, D D/G Dryer Assembly High DP Causing Excessive Blow Down
Attachment
11
WO 113484918, Lost Terminating Screw
WO 113484954, Extent of Condition for D DG, (3A)
WO 113484954, Extent of Condition for D DG, (3B)
WO 113484957, Extent of Condition for D DG, (3C)
WO 113484958, Extent of Condition for D DG, (3D)
WO 113486500, Troubleshoot/Repair/Replace DG D Air Pressure Alarm Relay
WO Instructions PMT for 113480917, Rev. 0
Section 1R20: Refueling and Other Outage Activities
0-47E820-1, Flow Diagram Control Rod Drive Hydraulic System, Rev. 32
0-OI-2B, Condensate Storage and Transfer System, Rev. 76
0-GOI-100-3A, Refueling Operations (In-Vessel Operations)
0-GOI-100-3B, Operations in Spent Fuel Pool Only
0-GOI-100-3C, Fuel Movement Operations During Refueling
0-GOI-100-3C, Fuel Movement Operations During Refueling, Attachment 6, Core Verification
3-47E804-1, Flow Diagram Condensate, Rev. 45
3-47E818-1, Flow Diagram Condensate Storage and Supply, Rev. 27
3-47E820-2, Flow Diagram Control Rod Drive Hydraulic System, Rev. 19
3-47E855-1, Flow Diagram Fuel Pool Cooling System, Rev. 24
3-AOI-100-1, Reactor Scram, Scram Reports, Rev. 58
3-GOI-100-12A, Unit Shutdown from Power Operations to Cold Shutdown and Reduction in
Power During Power Operations
3-GOI-100-1A, Unit Startup, Rev. 99
3-GOI-200-2, Primary Containment Initial Entry and Closeout, Rev. 34
3-OI-78, Fuel Pool Cooling and Cleanup System, Rev. 60
3-OI-85, Control Rod Drive System, Rev. 75
3-POI-78, Reactor Water Letdown During Refueling Outages Using Submersible Pump/Filter,
Rev. 06
3-SR-3.1.1.5(A), Control Rod Coupling Integrity Check, Att. 5, Startup Sequence, Rev. 25
3-SR-3.4.9.1(1), Reactor Heatup or Cooldown Rate Monitoring
3-TI-179, CILRT Data Acquisition System Setup, Rev. 8
3-TO-2012-0003; Clearance 3-001-0009B
3-TO-2012-0003; Clearance 3-068-0023A
3-TO-2012-0003; Clearance 3-071-0010
3-TO-2012-0003; Clearance 3-075-0009
3-TO-2012-0003; Clearance 3-075-0013
Browns Ferry Nuclear U3R15 Core Verification for BOC16 dated 4/10/2012
MMDP-11, Erection of Scaffolds / Temporary Wolf Platforms and Ladders, Rev. 3
MMTP-102, Erection of Scaffolds / Temporary Work Platforms and Ladders, Revs. 2 & 7
NPG-SPP-09.17, Temporary Equipment Control, Rev. 1
OPDP-1, Conduct of Operations, Rev. 23
PER 542193, Lock High Radiation Area Key
PER 542874, Unacceptable Housekeeping Practices in U3 RWCU HX Room
PER 543083, Housekeeping Inspection of 3B Reactor Water Cleanup Pump Room
PER 547169, U3 RWCU Equipment Drain Screens
PER 547172, U3 RWCU Pump Room Equipment Drain Screen
PER 549286, 3D Diesel Generator 7-Day Tank Leaking From Inspection Port
PER 554943, Pipe Support 3-47B458-564 - Core Spray
Attachment
12
PER 555573, Unit 3 Reactor Scram
PER 556790, Design Error with U3 3A USST
Scaffold Request # 03-1453-3, RWCU HX Room
Scaffold Request # 10-239-3, RWCU HX Room
SR 556367, GOI Step Not Fully Signed Off and Dated
3-TO-2012-004, sections 3-002-0001 and 3-078-0001 for Unit 3 Alternate Reactor Water Level
Control; 3-TO-2012-0003, Section 3-001-0008, for work on Main Steam Line Drain Inboard
Isolation Valve, 1-FCV-001-055;
3-TO-2012-0003; Clearance 3-001-0009B, for maintenance on 3-FCV-1-56; Clearance 3-068-
0023A, for maintenance of Recirculation Pump 3B; Clearance 3-071-0010, for maintenance on
RCIC Barometric Condenser Condensate Pump Motor; Clearance 3-075-0009, for 3A Core
Spray Motor Replacement; and Clearance 3-075-0013, for 3C Core Spray Motor Replacement.
3-POI-200.5
0-GOI-100-3A, Refueling Operations (In-Vessel Operations), 0-GOI-100-3B, Operations in the
Spent Fuel Pool Only, and 0-GOI-100-3C, Fuel Movement Operations During Refueling.
Attachment 6, of 0-GOI-100-3C.
Section 1R22: Surveillance Testing
0-TI-360, Containment Leak Rate Programs, Rev. 33
0-TI-360, Containment Leak Rate Programs, Rev. 33
0-TI-362, Inservice Testing of Pumps and Valves, Rev. 30
0-TI-362, Inservice Testing of Pumps and Valves, Rev. 30
2-SI-4.2.E-1(B), Drywell Equipment Drain Sump Flow Integrator Calibration, Rev. 22
2-SI-4.4.A.1, Standby Liquid Control Pump Functional Test, Rev. 66
3-47E811-1, Flow Diagram Residual Heat Removal System, Rev. 65
3D EDG LAT RA Recorder Chart A Test 1 and 2 Data, dated 4/03/12
3-SR-3.6.1.1.1(OPT-A), Primary Containment Total Leak Rate - Option A, Rev. 11
3-SR-3.6.1.3.10(B), Primary Containment Local Leak Rate Test Main Steam Line B: Penetration
X-7B, Rev. 07 performed on April 29, 2012
3-SR-3.8.1.9(3D OL), Diesel Generator 3D Emergency Load Acceptance Test with Unit 3
Operating, Rev. 14
3-TI-173, Primary Containment Inspection, Rev. 10 and Rev.11
3-TI-179, CILRT Data Acquisition System Setup, Rev. 08
ANSI/ANS-56.8-1994, Containment System Leakage Testing Requirements
Design Criteria BFN-50-7082, Standby Diesel Generator, Rev. 16
DWG 2-47E852-2, Flow Diagram Clean Radwaste & Decontamination Drainage, Rev. 33
FSAR Section 10.10, Emergency Equipment Cooling Water (EECW) System, BFN-24
FSAR Section 8.5, Standby AC Power Supply and Distribution, BFN-24
Main Control Room Logs
NEDP-14, Containment Leak Rate Programs, Rev. 09
NEDP-27, Past Operability Evaluations, Rev. 0
PER 533052, 3-FCV-1-27 failed as-found LLRT
PER 549232, As Found Integrator Indication Found Out Of Tolerance Low
PER 551019, Torus site glass readings were taken while isolated during CILRT
PER 554996, Evaluate potential HPCI preconditioning
PER 568095, 2-SI-4.4.A.1 SLC TEST, Schrader valve
PER 568705, Issue During SLC Pump Functional Test
PER 569867, HIgh vibration on 2A SLC pump
Attachment
13
PER 569895, HIgh vibration on 2B SLC pump
PER 569965, 4 AUOs Not Present for Surveillance
PER 570625, BFN-2-PMP-063-0006A, 2A SLC PUMP (GE-11-2A) Flowrate high
PER 570710,U2 SLC Storage Tank Decreasing Level Trend
PER 571768, Unit 2 SLC Storage Tank decreasing level trend.
SR 531728, Failure to Check Large Load Start
SR 531819, Failure to Send AUOs Locally for Large Load Start
SR 569401, 2-DRV-063-0530 leaking by its seat. Needed excess force to seat valve
Technical Specifications and Bases 3.3.8.1, Loss of Power (LOP) Instrumentation, Amendment
215
Technical Specifications and Bases 3.7.2, Emergency Equipment Cooling Water (EECW)
System and Ultimate Heat Sink (UHS), Amendment 215
Technical Specifications and Bases 3.8.1, AC Sources - Operating, Amendment 266
U2 Bases B 3.4.5 RCS Leakage Detection Instrumentation, Rev. 0
U2 Tech Spec 3.4.5, RCS Leakage Detection Instrumentation, Amendment 253
UFSAR, 4.10 Nuclear System Leakage Rate Limits, Amendment 22
WO 112511675, As Left - 3-SI-4.7.A.2.g-3/74g - PC LLRT - RHR Shutdown Cooling Suction
WO 112816329, Drywell Equipment Drain Sump Flow Integrator Calibration
WO 113145425, 2-SI-4.4.A.1, SLC Pump Functional Test
WO 113614430, Replace the Schrader valve on the bladder for the 2A SLC Pump
WO 113620697, 2-SI-4.4.A.1, SLC Pump Functional Test
WO 113625610, 2-DRV-063-0530 leaking by its seat, Needed excess force to seat valve
Section 1EP2: Alert and Notification System Evaluation
2012 Browns Ferry Emergency Planning Calendar mailer to members of the public in the 10-
mile EPZ
Documentation of bi-weekly siren tests and maintenance for 4th quarter 2011 and 1st quarter
2012
Documentation of Quarterly siren maintenance for 4th quarter 2011 and 1st quarter 2012
EPDP-10, Facilitation of the Alert and Notification System and Notification Tests, Rev. 4
EPDP-14, Evaluation of Changes to Alert and Notification Systems (ANS), Rev. 0
EPDP-16, Designated Emergency Response Equipment (DERE), Rev. 0
EPDP-17, NPG Emergency Plan Effectiveness Review (10 CFR 50.54(q))
EPDP-8, Emergency Preparedness Quality Related Programs, Rev. 1
EPFS-9, Inspection, Service, and Maintenance of the Prompt Notification System (PNS) at
Browns Ferry, Sequoyah, and Watts Bar Nuclear Plants, Rev. 6 and 7
Federal Signal 508 Electro-Mechanical Siren Installation and Operating Instructions, Rev. 12/11
Siren Annual Maintenance records: 2011 and 1st quarter 2012
SR 572389; admin requirements not met in implementing new ANS system
Section 1EP3: Emergency Preparedness Organization Staffing and Augmentation
System
2010, 2011, 2012 quarterly drill reports
2010, 2011, 2012 Unannounced pager test results
2012 Unannounced staffing drill report
239363 OSC Status Board Writer #1 failed to respond to Weekly Pager Test
243962 Operations Representative failed to respond to Weekly Pager Test
246558 Plant Assessment Team Leader failed to respond to Weekly Pager Test
Attachment
14
246569 OSC Status Board Writer #1 failed to respond to Weekly Pager Test
248540 OSC I/C Supervisor failed to respond to Weekly Pager Test
258558 Radiation Protection Manager failed to respond to Weekly Pager Test
266020 OSC I/C Engineer failed to respond to Weekly Pager Test
294582 OSC Mechanical Engineer failed to respond to Weekly Pager Test
327650 Site Vice President failed to respond to Weekly Pager Test
328191 OSC Director failed to respond to Weekly Pager Test
362821 Confused communication on the need to send B5b blackout fire pump to BFN
408093 Assistant OSC Director failed to respond to Weekly Pager Test
423217 CECC Plant Assessment Team member preparation for actual emergencies
475726 2011 Graded Exercise Corrective Actions
541288 QA SSA1203 - EP qualifications not in Qualification Matrix
542221 SAMG Decision Maker training requirements do not exclude Shift Managers as Site
Emergency Director
569374 Simulator issues during the BFN Off Year Exercise
CECC EPIP-3, Operations Duty Specialist Procedure for Alert, Rev. 41
CECC EPIP-4, Operations Duty Specialist Procedure for Site Area Emergency, Rev. 42
Emergency Response Organization Teams listing dated 6/22/2012
EPDP-3, Emergency Plan Exercises and Preparedness Drills, Rev. 5
EPIP-6, Activation and Operation of the Technical Support Center (TSC), Rev. 34
EPIP-7, Activation and Operation of the Operations Support Center (OSC), Rev. 29
EPT500A, 2012 EP Staff Orientation Course Description
TRN 30, Radiological Emergency Preparedness Training, Rev. 19
Various EP staff and ERO member training records
Section 1EP5: Maintenance of Emergency Preparedness
10CFR50.54(q) Evaluation of TEENS augmentation hardware addition
10CFR50.54(q) Evaluation of TSC Renovation
362854; NOUE declared - Tornado
364318; Tornado event
364674; Extensive loss of ANS due to tornadoes
453700; PAR training requirement
456771; RP ERO staffing PER not closed correctly
571878; admin error on 50.54q eval of TEENS implementation
572826; EPDP-17 enhancement to add subject matter experts in 50.54q screening
95003-005, BFN NRC Column 4 Response Project and Administrative Controls - Appendix H,
Rev. 1: ERO Readiness Performance Area Report
BFN Quality Assurance - Emergency Preparedness Drill Assessment - QA-11-007 dated April
21, 2011
BFN Quality Assurance - Emergency Preparedness Equipment and Facility Readiness, QA-BF-
11-008 dated June 30, 2011
BFN Self-assessment BFN-EP-S-10-001, B5B Commitments
BFN Self-assessment BFN-EP-S-11-001, Effectiveness Reviews
Drill and exercise reports, 2010, 2011, and 2012
EPDP-1, Procedures, Maps, and Drawings, Rev. 3
EPDP-16, Designated Emergency Response Equipment (DERE), Rev. 0
EPDP-17, NPG Emergency Plan Effectiveness Review, Rev. 0
Event records of NOUE declared on 4/27/2011 - Tornado with Extended Loss of Off-site Power
Attachment
15
NPG-SPP-18.3, Emergency Preparedness, Rev. 1
REP, Radiological Emergency Plan, (Appendix A - BFN), Rev. 97
REP, Radiological Emergency Plan, (Generic Part), Rev. 97
Self-assessment CRP-EP-S-11-03, Site Tornado Procedure, BP-128, dated September 28,2011
Self-assessment CRP-EP-S-12-005; Training Program comparison
Self-assessment CRP-EP-S-12-006, REP drill
Self-assessment CRP-EP-S-12-020; EP Records
SPP-3.1, Corrective Action Program, Rev. 4
TVA Quality Assurance - Emergency Preparedness Audit Report SSA1003 dated May 20, 2010
TVA Quality Assurance - Emergency Preparedness Audit Report SSA1203 dated April 24, 2012
Section 1EP6: Drill Evaluation
Browns Ferry, Off Year Exercise Report
CECC-EPIP-1, Emergency Classification Procedure, REV. 53
EPIP-1, Emergency Classification Procedure, REV. 47
NP-REP, Radiological Emergency Plan, (Generic Part), Rev. 97
NP-REP, Radiological Emergency Plan, Appendix A, Rev. 97
PER 567663, Accountability report inaccuracy during EP drill
PER 568729, Revise EPIP-7, App. B to Indicate OSC Minimum Staffing
PER 569310, CECC ERO member failed to respond to CECC activation
PER 569374, Simulator Issues during the BFN Off Year Exercise
PER 570670, During the Unannounced Staffing Drill, TEENS System Delay
PER 571025, During EP OYE Simulator Stack Rad Simulation did not operate as expected
PER 571053, During the EP Unannounced Staffing Drill issues were observed
PER 571382, During the 2012 EP Off Year Exercise Stack Monitor Simulation was an issue
PER 572271, Focus areas found in the June 13th BFN REP OYE
Performance Indicator Data from June 2012
Section 2RS1: Radiological Hazard Assessment and Exposure Control
(Annual Inventory Of Non-Fuel SNM and Other Items (Trash) In Unit 1, 2 And 3 Spent Fuel
Pools Performed 8/10-25/2011.)
0-TI-540, Storage of Material in the spent Fuel Storage Pool (SFSP) and Transfer Canal
(U1/U2), Rev. 2
Browns Ferry Technical Specification 5.7 Administrative Controls-High Radiation Area
NPG-SPP-05.0, Radiological and Chemistry Control, Rev. 1
NPG-SPP-05.1, Radiological Controls, Rev. 2
NSTS Confirmation Form 2012 Annual Inventory Reconciliation [Browns Ferry 1 AmBe Source],
Dated 1/18/2012
NSTS Confirmation Form 2012 Annual Inventory Reconciliation [Browns Ferry 3 Cs-137
Sources], Dated 1/18/2012
PER 334211 Track and trend radworker practices in drywell U2R16
PER 334244 Radworker practices in drywell U2R16
PER 439979 RP posted area incorrectly
PER 475108 U1R9 Drywell access room improperly posted
PER 512565 worker put tie wrap in mouth in RCA
PER 512567 building scaffold in unsurveyed area
RCDP-1, Conduct of Radiological Controls, Rev. 3
RCI-1.1, Radiation Operations Program Implementation, Revision 149
Attachment
16
RCI-1.2, Radiation, Contamination and Airborne Surveys, Revision 16
RCI-17, Control of High Radiation Areas and Very High Radiation Areas, Revision 71
RCI-26, Radiation Protection Department Standards and Expectations, Revision 19
RCI-33, Diving Operations on the Refuel Floor, Rev. 9
RCI-34, Remote Monitoring, Revision 12
RCI-40.0, RP Actions for Operation's Unit 0 (Common) Procedural Hold Points, Revision 17
RCI-47, Diving Operations in the Radiologically Controlled Area, Rev. 1
RCI-9.1, Radiation Work Permits, Revision 70
RWP 1238-0001, Unit-3R15 Refueling Outage Drywell Outside Support
RWP 1238-0002, Unit-3R15 Refueling Outage Drywell Outside Support [High Rad]
RWP 1238-0003, Unit-3R15 Outage Drywell Miscellaneous System Support [Locked High Rad]
RWP 1238-0012, Unit-3R15 Outage Drywell Main Steam System Maintenance [High Rad]
RWP 1238-0033, Unit-3R15, Outage Drywell Feedwater System Maintenance [Locked High
Rad]
RWP 1238-0683, Unit-3R15, Outage, Drywell Reactor Water Recirculation System [Continuous
Coverage- Locked High Radiation Area]
RWP 1238-0693, Unit-3R15, Outage, Drywell Reactor Water Cleanup System Maintenance
[Locked High Rad]
SR 532617 Worker got separated from escort
SR 532875 Inaccurate rad tag on a box
SR 532981 Small air activity excursion on RFF during Rx disassembly
SR 534873 Coordination issues obtaining RWCU sludge sample.
SR 534880 Deterioration of padding on Knee anchors U1 593
Survey M-010612-2, Unit 3 RXB 593' RWCU BW Transfer Pump Room, 01/06/2012
Survey M-020712-13, Unit 2 RXB 519' Under Torus, 02/07/2012
Survey M-021012-10, 0-CASK-079-0100/1 (MPC SN-0237), 02/10/2012
Survey M-102411-11, Unit 2 TB 586' 2A SJAE Room, 10/24/2011
Survey M-20120306-26, ISFSI Pad, 03/06/2012
Section 2RS6: Radioactive Gases and Liquid Effluent Treatment
Procedures, Guidance Documents, and Manuals
0-ODCM-001, Offsite Dose Calculation Manual, Rev. 21
NPG-SPP-05.14, Guide for Communicating Inadvertent Radiological Spills/Leaks to Outside
Agencies, Rev. 0
NPG-SPP-05.15, Fleet Ground Water Protection Program, Rev.2
0-TI-15, Radioactive Gaseous Effluent Engineering Calculations and Measurements, Rev. 15
0-SI-4.8.A.1-1, Liquid Effluent Permit, Rev. 74
0-SI-4.8.B.1.a.2, Airborne Effluent Release Rate by Manual Sampling When a Gaseous Effluent
Monitor is Inoperable, Rev. 31
0-SI-4.8.B.2-1, Airborne Effluent Analysis - Particulate and Charcoal Filter Analysis, Rev. 37
0-SI-4.8.B.2-5, Airborne Effluent Analysis - Monthly Tritium, Rev. 30
0-SI-4.8.B.2-8, Airborne Effluent Analysis - Stack Noble Gas, Rev. 12
0-SI-4.8.B.2-4, Airborne Effluent Analysis - Monthly Gamma Isotopic, Rev. 30
CI-714, Particulate and Charcoal Filter Sampling and Analysis, Rev. 30
CI-738, Sampling Effluent Monitors (CAMS) for Tritium and Gamma Isotopics, Rev. 31
0-SI-2.1-2, Airborne Effluent Radiation Monitor Source Checks, Rev. 45
1-SIMI-90B, Radiation Monitoring System Scaling and Setpoint Documents, Rev. 41
2010 Radiological Effluent Release Report
Attachment
17
2011 Radiological Effluent Release Report
2002 Radiological Effluent Release Report - Abnormal Release Addendum
Records and Data Reviewed
Browns Ferry UFSAR Chapter 9
0-SR-3.6.4.3.2(A)-SBGTS Iodine Removal Efficiency (Train-A), 8/23/2010
0-SR-3.6.4.3.2(A)-SBGTS Iodine Removal Efficiency (Train-A), 7/13/2011
Gaseous Release Permits: 120323.030.020.G, 120315.037.020.G, 120350.030.021.G,
20328.032.020.G, 120333.043.019.G, 120340.046.020.G, 120330.040.025.G
Surveillance Task Sheet: 0-SI-4.8.B.2-1- Airborne Effluent Analysis- Particulate & Charcoal
Filter Analysis, 5/1/2012
System Health Reports, Each Unit System 66 - Off-Gas, 2/1/2011-1/31/2012
System Health Report, System 77 -Radwaste, 10/1/2011-1/31/2012
System Health Report, Each Unit System 90- Radiation Monitoring, 10/1/2011-1/31/2012
Cross-Check Analysis Data: 1st Quarter 2010 through 2nd Quarter 2011
Chemistry Focused Self Assessment Report - BFN-CEM-F-11-001, Performed 6/6-17/2011
White paper documenting Ground Water Monitoring in 2010 and 2011 with results
CAP Documents
PER 257903 2-RM-090-013D, RCW Effluent Offline Rad Monitor alarmed on Hi Rad Setpoint
PER 313929 1Q FY11 Radwaste water processing and effluents continues to be problem areas.
PER 324700 Unit 3 Station Sump tritium results from the sample obtained 1/18/2011
PER359503 Unmonitored release at the gas stack
PER 367604, Insufficient sample equipment for inop Effluent CAM monitors
PER 532416, Possible release path to Waters of the US
Section 2RS7: Radiological Environmental Monitoring Program (REMP)
Procedures and Guidance Documents
Cl-420, Collection of Radiological Environmental Monitoring Samples, Revision 03
EPFS-8, Servicing of Radiological Water Samplers, Revision 2
EPFS-12, Repair and Preventative Maintenance Procedure for Radiological
EPFS-03, Servicing of Meteorological Equipment at Environmental Data Stations, Rev 15
EPFS-07, Radio and Meteorological Tower Inspection, Rev 4
EPFS-06, Calibration of Environmental Data Station Data Logger and Sonic Channels, Rev 16
Environmental Monitoring Air Sampling System, Rev 01
EMSTD-01, Environmental Radiological Monitoring Program, R25
Records and Data Reviewed
Annual Radiological Environmental Operating Report 2010 & 2011
Field Collection Sheets for June 4, 2012 Environmental Run
EPFS-6 Data sheet 1 for Cal dates 3/21/12; 10/04/11; 04/13/11; 10/14/10; 08/24/10
EPFS-6 Data sheet 6 for dates 03/21/12; 10/31/11; 10/04/11; 04/12/11; 10/14/10
EPFS-6 Data sheet 5 for dates 03/22/12; 04/12/11; 10/04/11; 10/20/10
EPFS-6 Data sheet 4 for dates 03/21/12; 10/04/11; 04/13/11; 10/14/10
EPFS-6 Data sheet 3 for dates 03/21/12; 10/04/11; 04/13/11; 10/14/10
EPFS-6 Data sheet 2 for dates 03/21/12; 10/04/11; 04/13/11; 10/14/10
Calibration Data Sheets for REMP Air Sampler Gas meter 2010 & 2011
Attachment
18
EPFS 1 Attachment 2 Trouble Report: 10BFN538, 10BFN536, 10BFN560, 10BFN561,
10BFN557, 10BFN549, 10BFN506
QA Record L17111221800, TVA Quality Assurance- Nuclear Power Group- Fleet Comparative
Report SSA1107, 12/20/11
CAP Documents
PER 259776- The BFN REMP air filter and charcoal cartridge samples invalid
PER 366333- Loss of power to REMP air samplers
PER 450297- REMP sample not analyzed and not recorded in PER
PER 515446- REMP sample
Section 2RS8: Radioactive Material Processing and Transportation
Procedures, Manuals, and Guides
Energy Solutions Procedure, FO-OP-022, Ecodex Precoat/Powdex/Solka-Floc/Diatomaceous
Earth/Zeolite Dewatering Procedure for Energy Solutions14-215 or Smaller Liners, Rev. 23
Radioactive Material Shipment Manual (RMSM), Volume I, Rev. 40
Radioactive Material Shipment Manual (RMSM), Volume II, Rev. 42
Radioactive Material Shipment Manual (RMSM), Volume III, Rev. 39
RWI-001, Administration of the Radioactive Material and Radwaste Packaging and
Transportation Program, Rev 9
RWTP-102, Use of Casks, Rev. 2
RWI-111, Storage of Radioactive Waste and Materials, Rev. 18
RWI-112, Container Markings, Rev. 2
0-OI-77G, Duratek Procedure FO-OP-32, Set Up and Operating Procedure for the RDS-1000
Unit at TVA Browns Ferry, Rev. 2
0-PCP-001, Process Control Program Manual (PCP), Rev. 4
NPG-SPP-3.1, Corrective Action Program, Rev. 2 and Rev. 3
Shipping Records and Radwaste Data
Certificate of Compliance No. 9168 for the Model No. 8-120B, 5/25/12
Certificate of Compliance No. 9204 for the Model No. 10-160B, 5/25/12
Gamma Isotopic Analysis Results - ID # 20120227-29 [For survey 022712-29, trash dumpster],
2/27/12
Gamma Isotopic Analysis Results - ID # 20100607-23 [NCDM Coupon 101], 6/7/10
Gamma Isotopic Analysis Results - ID # 20100607-25 [NCDM Coupon 103], 6/7/10
Gamma Isotopic Analysis Results - ID # 20100607-27RC [NCDM Coupon 047], 6/7/10
Gamma Isotopic Analysis Results - ID # 20100607-26 [NCDM Coupon 192], 6/7/10
Letter to File, Browns Ferry Nuclear Plant - Personnel Qualified to Ship Radioactive
Material/Waste, 3/19/12
List of Radioactive Material Storage Areas [Spreadsheet]
List of Red System 077 Issues
List of Outstanding Work Orders for System 077 [Radwaste]
Liquid Radwaste System (System 077) Health Report (2/1/12 - 5/31/12), 6/19/12
Liquid Radwaste System (System 077) Health Report (10/1/2011 - 1/31/2012), 5/17/12
Project Plan, BFN Radwaste Legacy Project, Project ID: 100533, Rev. 1, 2/1/12
Qualification Matrix Report for selected individuals to verify Subpart H training
Radioactive Material Shipping logs for the period 7/10/10 to 5/17/12
Attachment
19
Radiological Survey M-20120517-23, Pre-Shipment Survey on HIC# CL40524-9
Radiological Survey M-20120620-17, Down Post, HIC transfer complete.
Radiological Survey M-20120620-19, Pre-Shipment on cask # 14-170-35
Radiological Survey M-022412-4, Other - Trash Dumpster
Radiological Survey M-022712-29, Job Coverage [Trash Dumpster]
Radiological Survey M-20120312-12, Trash Dumpster from PA
RWP12040086, Legacy Radwaste Project (LHRA), Rev. 0
Shipment 100618, Corrosion coupons in a DOT 7A container, Type A
Shipment 120401, Liquid tanker, Low Specific Activity (LSA-I)
Shipment 120455, Control Rod Drives (2 boxes), Type A
Shipment 110804, Empty 8-120A cask, Excepted package-empty
Shipment 110318, DAW (2 sealand containers), Low Specific Activity (LSA-II)
Shipment 101111, DAW (1 sealand container), Low Specific Activity (LSA-II)
Shipment 110902, Surveillance Capsule, Type A
Shipment 100326, Control Rod Drives (2 boxes), Type A
Shipment 100327, Control Rod Drives (2 boxes), Type A
Shipment 100328, Control Rod Drives (2 boxes), Type A
Shipment 120616, Dewatered Resin, Low Specific Activity (LSA-II)
10 CFR Part 61 Analyses, DAW 2012; CWPS 2012; RWCU 2010 and 2012 Preliminary;
Thermex 2010 and 2012 Preliminary,
CAP Documents
PER 513962, Non-RCA Trash dumpster alarms truck monitor
PER 520927, Non-RCA Trash dumpster alarms truck monitor
PER 409367, Equipment Sump over flowed contaminating RW 546
PER 425240, Radwaste El. 546 posted CA due to flooding from floor drains
PER 433904, RW 546 C-zone due to Equipment Sump overflow
PER 429803, Trend of flooding RW 546 elevation
PER 451830, Entire 546 elevation of the Rad waste building flooded
PER 456136, RW elevation 546 was flooded again spreading more contamination
PER 533414, 10CFR61 samples do not include a RWCU Sample
PER 441666, Intruder brakin at Low Level Radwaste yard
PER 254001, ATIS Radwaste Shipping Task tracking problem
PER 343736, Radioactive Material stored for years without disposition determination
PER 431466, Received notification that torque values were incorrect upon receipt of ISP
capsule
PER 236118, Two boxes of Used Control Rod Drives Shipped to GEH Improperly
PER 453834, Adverse Trend of flooding RW 546 elevation
Apparent Cause Evaluation Report, PER 453834, 10/28/11
PERs written by licensee during inspection activities:
SR 568025, O-OI-77E needs to be revised to correct references to procedures that are no
longer in existence.
SR 570902, PER 236118 needs to be revisited. Upon review, the corrective actions were
inadequate.
SR 571151, PER 431466 needs to be revisited. Upon review, the corrective actions were
inadequate.
Attachment
20
Section 4OA1: Performance Indicator Verification
3-47E812-1, Flow Diagram for HPCI, Rev. 64
3-OI-73, High Pressure Coolant Injection System, Rev. 52
571936; improve DEP PI advance scheduling
572831; PAR development in licensed operator training PI opportunities
BFN-50-7073, Design Criteria Document for the HPCI system, Rev. 22
CECC EPIP-3, Operations Duty Specialist Procedure for Alert, Rev. 41
Consolidated Date Entry Sheets for Units 1, 2 and 3 for the Safety System Functional Failures
Documentation of ANS tests for 4th quarter 2011 - 1st quarter 2012
Documentation of DEP opportunities for 4th quarter 2011 - 1st quarter 2012
EPDP-11, Emergency Preparedness Performance Indicators, Rev. 3
EPIP-2, Notification of Unusual Event, Rev. 31
EPIP-3, Alert, Rev. 34
EPIP-4, Site Area Emergency, Rev. 33
LER 259/2011-006-00, Loss of Safety Function (HPCI) Due to Primary Containment Isolation.
Licensed Operator Training Scenarios 04, 17, 06, 18, 30, and 05 from 4th quarter 2011
Maintenance Rule Function Failure Report from April 1, 2011 to March 31, 2012
NPG-SPP-02.2, Performance Indicator Program, Rev. 3
NPG-SPP-03.4, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting
10 CFR 50.65, Rev. 01
PER 439338 RP tech posted an area incorrectly
PER 533834 Contractor receives uptake during hydrolaze activities
PER 534086 Laborer contaminated while working in an area near where CRD header was
being hydrolased.
RCI-39, Radiation Protection Cornerstones, Rev. 9
SR 532755, Dosimetry alarms due to being run through x-ray machine
Section 4OA2: Identification and Resolution of Problems
0-47E820-1, Flow Diagram Control Rod Drive Hydraulic System, Rev. 32
0-OI-2B, Condensate Storage and Transfer System, Rev. 76
1-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity, Rev. 04
2-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity, Rev. 14
3-47E804-1, Flow Diagram Condensate, Rev. 45
3-47E818-1, Flow Diagram Condensate Storage and Supply System, Rev. 27
3-47E820-2, Flow Diagram Control Rod Drive Hydraulic System, Rev. 19
3-47E855-1, Flow Diagram Fuel Pool Cooling System, Rev. 24
3-GOI-100-3A, Refueling Operations (RX Vessel Disassembly and Floodup), Rev. 53
3-OI-78, Fuel Pool Cooling and Cleanup System, Rev. 60
3-OI-85, Control Rod Drive System, Rev. 75
3-POI-200.5, Operations with Potential for Draining the Reactor Vessel/Cavity, Rev. 11
3-POI-78, Reactor Water Letdown During Refueling Outages Using Submersible Pump/Filter,
Rev. 06
Engineering trend report data from January 1, 2011 to December 1, 2011
Integrated Trend Report, Q1FY12, October 1 December 31, 2012
Integrated Trend Report, Q2FY12, January 1 March 31, 2011
PE-P4461A, Recirculation System Suction Plug Installation/Removal Procedure for Browns
Ferry Nuclear Station under Project PE 00-829/1299 & 09-1614, Rev. 4
Attachment
21
PE-P4462A, Jet Pump Plug Procedure for Browns Ferry Nuclear Station under Project PE 00-
829, Rev. 0
PE-P4850, Operating and Maintenance Instructions for the Main Steam Line Plugs and
Installation/Removal Tools for Browns Ferry Station - Project PE 998, Rev. 2
PER 471366, CAP gaps to excellence plan
PER 491985, Human Performance gaps to excellence plan
PER 512589, Cross-functional issue on outage-related worker practices
PER 539854, Engineering has documented several inappropriate action closures
PER 563559, QA identified trend on BFN Fire Operations Training
RPT-CAP011, Gognos PER Word Search report from Jan 1, 2012 to June 29, 2012
Section 4OA3: Event Follow-up
0-TI-230V, Vibration Program, Rev. 10
0-TI-346, Maintenance Rule Performance Indicator Monitoring, Trending, and Reporting -
10CFR50.65, Rev. 38
1-SR-3.3.8.2.1(A), RPS Circuit Protector Calibration/Functional Test For 1A1 and 1A2, Rev. 6
3-AOI-100-1, Reactor Scram, Rev. 58
Browns Ferry - Emergency Diesel Generator System Vulnerability to Functional Failure
Assessment, dated May 7, 2009
Design Criteria BFN-50-7082, Standby Diesel Generator, Rev. 16
Drawing 1-45E641-3, Instr & Controls Power Sys Schematic Diagram SH-3, Rev. 5
Drawing, 0104D3695-1, Isolated Phase Bus Return Air Duct, dated 1/20/12
Electro-Motive Vibration Guidelines Industrial Power Units, letter dated October 29, 1982
EMD Power Systems Owners Group Meeting, Diesel Generator Vibration Acceptable Criteria,
dated June 26-28, 1991
FSAR Section 11, Power Conversion Systems, BFN-24
FSAR Section 8.4, Normal Auxiliary Power System, BFN-24
FSAR Section 8.5, Standby AC Power Supply and Distribution, BFN-24
Main Control Room Logs
NPG-SPP-06.2, Preventive Maintenance, Rev.0
NPG-SPP-06.2, Preventive Maintenance, Rev.04
NPG-SPP-09.18, Integrated Equipment Reliability Program, Rev. 02
NPG-SPP-09.18.1, System Vulnerability Review Process (MCIP Reviews), Rev. 4
NPG-SPP-09.18.2, Equipment Reliability Classification, Rev. 0
NPG-SPP-09.18.2, Equipment Reliability Classification, Rev. 01
NPG-SPP-2.3, Operating Experience Program, Rev. 3
OE25284 - Emergency Diesel Generator Governor Drive Oil Supply Line Sheared, North Anna
1 and 2
Operations Standing Order 174, Rev. 1, To establish Operations Department expectation when
as-found data is outside of acceptable regulatory or programmatic requirements
PER 131365, Out of Tolerance Time Delay Relay
PER 151812, RPS Circuit Protector Failed Acceptance Criteria
PER 178286, Acceptance Criteria Failed
PER 248513, Failed Acceptance Criteria Step 7.2 (28)
PER 362395, Oil Leak Resulting in Emergency Shutdown of C DG
PER 391479, Classification of System 55 Power Supplies
PER 413140, 1A1 RPS Circuit Protector Undervoltage Trips
PER 438808, Unknown Object Found in U3 Phase Bus Duct
Attachment
22
PER 440359, U3 Scrammed on September 28, 2011 at 0414
PER 442914, Evaluation of Surveillance Data from Past Performances
PER 486780, 3C1 Relay Results Below Acceptance Criteria
PER 496592, Fire in Annunciator Panel 3-XA-55-5A
SPP-3.9, Operating Experience Program, Revs. 4 and 5
SPP-6.2, Preventive Maintenance, Rev.09
SPP-9.18.2, Equipment Reliability Classification, Rev. 00
SR 496007, U-3 Annunciator Panel 9-5A Fire and AOI entry
Technical Specification and Bases 3.3.8.2, Reactor Protection System (RPS) Electric Power
Monitoring, Amendment 263 and Rev. 43, respectively
Technical Specifications and Bases 3.8, Electrical Power System, Amendment 266
Technical Specifications and Bases Section 3.8, Electrical Power Systems, Amendment 280
and Rev. 52 respectively
TVA-NQA-PLN89-A, Nuclear Quality Assurance Plan (NQAP), Rev. 23, 24, 25 and 26
Attachment
LIST OF ACRONYMS
ADAMS - Agencywide Document Access and Management System
ADS - Automatic Depressurization System
ALARA As Low As Reasonably Achievable
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
DAC Derived Air Concentration
DCN - design change notice
ED Electronic Dosimeter
EDG - emergency diesel generator
EECW - emergency equipment cooling water
FE - functional evaluation
FPR - Fire Protection Report
FSAR - Final Safety Analysis Report
HP Health Physics
IMC - Inspection Manual Chapter
JOG Joint Owners Group
LER - licensee event report
LHRA Locked High Radiation Area
NCV - non-cited violation
NRC - U.S. Nuclear Regulatory Commission
NSTS National Source Tracking System
OA Other Activity
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
RS Radiation Safety
RTP - rated thermal power
RPS - reactor protection system
RWP - radiation work permit
SDP - significance determination process
SBGT - standby gas treatment
SNM - special nuclear material
Attachment
24
SSC - structure, system, or component
TI - Temporary Instruction
TIP - transverse in-core probe
TLD Thermoluminescent Dosimeter
TRM - Technical Requirements Manual
TS - Technical Specification(s)
U1 Unit 1
U2 Unit 2
U3 Unit 3
UFSAR - Updated Final Safety Analysis Report
URI - unresolved item
VHRA Very High Radiation Area
WO - work order
Attachment