ML12312A082
ML12312A082 | |
Person / Time | |
---|---|
Site: | Brunswick ![]() |
Issue date: | 11/07/2012 |
From: | Randy Musser NRC/RGN-II/DRP/RPB4 |
To: | Annacone M Carolina Power & Light Co |
Shared Package | |
ML12325A266 | List: |
References | |
IR-12-004 | |
Download: ML12312A082 (45) | |
See also: IR 05000324/2012004
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
November 7, 2012
Mr. Michael J. Annacone
Vice President
Brunswick Steam Electric Plant
P.O. Box 10429
Southport, NC 28461-0429
SUBJECT: BRUNSWICK STEAM ELECTRIC PLANT - NRC INTEGRATED INSPECTION
REPORT NOS.: 05000325/2012004 AND 05000324/2012004
Dear Mr. Annacone:
On September 30, 2012, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Brunswick Unit 1 and 2 facilities. The enclosed integrated inspection report
documents the inspection findings, which were discussed on October 11, 2012, with you 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 and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
One NRC-identified and one self-revealing finding of very low safety significance (Green) were
identified during this inspection. These findings were determined to involve a violation of NRC
requirements. Further, two licensee-identified violations were determined to be of very low
safety significance and are listed in this report. The NRC is treating these findings as non-cited
violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance of these NCVs, 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 the Regional Administrator Region II; the Director, Office of Enforcement, United
States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident
Inspector at the Brunswick Steam Electric Plant.
If you disagree with the cross-cutting aspect assignment in this report, you should provide a
response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the
Brunswick Steam Electric Plant.
M. Annacone 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
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/
Randall A. Musser, Chief
Reactor Projects Branch 4
Division of Reactor Projects
Docket Nos.: 50-325, 50-324
Enclosure: Inspection Report 05000325, 324/2012004
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
ML12312A082_________________ x SUNSI REVIEW COMPLETE x FORM 665 ATTACHED
OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP RII:DRP
SIGNATURE JSD: /RA/ RAM RA for Via e-mail Via e-mail Via e-mail Via e-mail JGW: /RA/
MPS
NAME JDodson MCatts MSchwieg PNiebaum LLake MEndress JWorosilo
DATE 10/24/2012 11/07/2012 10/24/2012 10/29/2012 10/26/2012 10/25/2012 10/15/2012
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRP RII:DRS
SIGNATURE RAM: /RA/ Via e-mail
NAME RMusser MSpeck
DATE 11/7/2012 11/06/2012
E-MAIL COPY? YES NO YES NO
M. Annacone 3
cc w/encl: Lee Grzeck
Plant General Manager Regulatory Affairs Manager
Brunswick Steam Electric Plant Brunswick Steam Electric Plant
Progress Energy Progress Energy Carolinas, Inc.
Electronic Mail Distribution Electronic Mail Distribution
Edward L. Wills, Jr. Randy C. Ivey
Director Site Operations Manager, Nuclear Oversight
Brunswick Steam Electric Plant Brunswick Steam Electric Plant
Electronic Mail Distribution Progress Energy Carolinas, Inc.
Electronic Mail Distribution
J. W. (Bill) Pitesa
Senior Vice President Paul E. Dubrouillet
Nuclear Operations Manager, Training
Duke Energy Corporation Brunswick Steam Electric Plant
Electronic Mail Distribution Electronic Mail Distribution
John A. Krakuszeski Joseph W. Donahue
Plant Manager Vice President
Brunswick Steam Electric Plant Nuclear Oversight
Electronic Mail Distribution Progress Energy
Electronic Mail Distribution
Lara S. Nichols
Deputy General Counsel Senior Resident Inspector
Duke Energy Corporation U.S. Nuclear Regulatory Commission
Electronic Mail Distribution Brunswick Steam Electric Plant
U.S. NRC
M. Christopher Nolan 8470 River Road, SE
Director - Regulatory Affairs Southport, NC 28461
General Office
Duke Energy Corporation John H. O'Neill, Jr.
Electronic Mail Distribution Shaw, Pittman, Potts & Trowbridge
2300 N. Street, NW
Michael J. Annacone Washington, DC 20037-1128
Vice President
Brunswick Steam Electric Plant Peggy Force
Electronic Mail Distribution Assistant Attorney General
State of North Carolina
Annette H. Pope P.O. Box 629
Manager-Organizational Effectiveness Raleigh, NC 27602
Brunswick Steam Electric Plant
Electronic Mail Distribution (cc w/encl - continued)
M. Annacone 4
cc w/encl contd:
Chairman
North Carolina Utilities Commission
Electronic Mail Distribution
Robert P. Gruber
Executive Director
Public Staff - NCUC
4326 Mail Service Center
Raleigh, NC 27699-4326
Anthony Marzano
Director
Brunswick County Emergency Services
Electronic Mail Distribution
Public Service Commission
State of South Carolina
P.O. Box 11649
Columbia, SC 29211
W. Lee Cox, III
Section Chief
Radiation Protection Section
N.C. Department of Environmental Commerce & Natural Resources
Electronic Mail Distribution
Warren Lee
Emergency Management Director
New Hanover County
Department of Emergency Management
230 Government Center Drive
Suite 115
Wilmington, NC 28403
M. Annacone 5
Letter to Michael J. Annacone from Randall A. Musser dated November 7, 2012
SUBJECT: BRUNSWICK STEAM ELECTRIC PLANT - NRC INTEGRATED INSPECTION
REPORT NOS.: 05000325/2012004 AND 05000324/2012004
Distribution w/encl:
J. Baptist, RII EICS
L. Douglas, RII EICS
OE Mail (email address if applicable)
RIDSNRRDIRS
PUBLIC
R. Pascarelli, NRR ((Regulatory Conferences Only))
RidsNrrPMBrunswick Resource
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-325, 50-324
Report Nos.: 05000325/2012004, 05000324/2012004
Licensee: Carolina Power and Light (CP&L)
Facility: Brunswick Steam Electric Plant, Units 1 & 2
Location: 8470 River Road, SE
Southport, NC 28461
Dates: July 1, 2012 through September 30, 2012
Inspectors: M. Catts, Senior Resident Inspector
M. Schwieg, Resident Inspector
P. Niebaum, Acting Senior Resident Inspector
J. Dodson, Senior Project Engineer (1R04, 1R05, 4OA2)
L. Lake, Senior Reactor Inspector (4OA5)
M. Endress, Reactor Inspector (1R07)
Approved by: Randall A. Musser, Chief
Reactor Projects Branch 4
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000325/2012004, 05000324/2012004; 07/01/12 - 09/30/12; Brunswick Steam Electric
Plant, Units 1 & 2; Refueling and Other Outage Activities, Identification and Resolution of
Problems
This report covers a three-month period of inspection by resident inspectors and announced
baseline inspections by regional inspectors. Two Green findings were identified by the
inspectors. The significance of most findings is indicated by their color (Green, White, Yellow,
Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process
(SDP). 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.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Barrier Integrity
Green: The inspectors identified a Green non-cited violation (NCV) of TS 3.6.4.1,
Secondary Containment because the licensee did not maintain secondary containment
operable as required during a maintenance activity considered an operation with a
potential for draining the reactor vessel (OPDRV). Once questioned by the inspectors,
the licensee restored secondary containment, developed an Operation standing
instruction (12-052) to treat the activity as an OPDRV and placed this issue into its
corrective action program (CAP) as AR 562188.
The failure to maintain secondary containment operable while Unit 1 was in Mode 4 with
an OPDRV in progress was a performance deficiency. The finding was more than minor
because it was associated with the configuration control attribute of the Barrier Integrity
Cornerstone, and adversely affected the cornerstone objective to provide reasonable
assurance that physical design barriers (fuel cladding, reactor coolant system, and
containment) protect the public from radionuclide releases caused by accidents or
events because the Unit 1 secondary containment boundary was not preserved or
maintained. The inspectors evaluated the finding using Inspection Manual Chapter
(IMC) 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings,
which required an analysis using IMC 0609 Appendix G since the reactor was in Mode 4
(cold shutdown). The finding was determined to be of very low safety significance
(Green) according to IMC 0609 Appendix G, Attachment 1, Checklist 6, since a
quantitative assessment (Phase 2 or Phase 3 evaluation) was not required. Specifically,
the inspectors determined that the licensee maintained adequate mitigation capability for
reactor vessel water level inventory and an event did not occur that could be
characterized as a loss of control. The cause of this finding was directly related to the
cross-cutting aspect of Accurate Procedures in the Resources component of the Human
Performance area, because the licensee did not consider the recirculation pump seal
replacement activity to be OPDRV based on procedural guidance that contains
exclusions to what are considered OPDRV activities. H.2(c) (Section 1R20)
3
Cornerstone: Emergency Preparedness
Green: A self-revealing Green NCV of 10 CFR 50.54(q)(2) was identified for the
licensees failure to properly evaluate or consider the impact to emergency response
facilities of design change ESR98-00436 which was implemented in 1999. This resulted
in the loss of Emergency Response Facility Information System (ERFIS), Emergency
Response Data System (ERDS), Safety Parameter Display System (SPDS), and all
displays including radiation monitors for the emergency response facilities. Specifically,
the licensee failed to ensure that adequate emergency response facilities and equipment
were available as required by the Brunswick Nuclear Plant Radiological Emergency
Plan, Section 1.3.1.3 revision 80 and 10 CFR 50.47(b)(8). This issue was captured in the
The licensees failure to properly evaluate or consider the impact to emergency
response facilities of design change ESR98-00436 which was implemented in 1999 was
a performance deficiency. Specifically, the licensee introduced a single point failure
mode which did not meet the design requirements specified in their Design Basis
Document (DBD 60) sections 3.6.7.2 and 3.6.7.3. This resulted in the licensees failure
to ensure that adequate emergency response facilities and equipment were available as
delineated in the Updated Final Safety Analysis Report (UFSAR) Section 7.7.1.9, and
required by the Brunswick Nuclear Plant Radiological Emergency Plan, Section 1.3.1.3,
revision 80, and 10 CFR 50.47(b)(8). The finding was more than minor because it
adversely affected the Emergency Preparedness Cornerstone objective of ensuring that
the licensee was capable of implementing adequate measures to protect the health and
safety of the public in the event of a radiological emergency. Specifically, the Facilities
and Equipment attribute was affected during the time when the ERFIS, ERDS, SPDS,
and all displays including radiation monitors for the emergency response facilities were
degraded, and as a result did not meet 10 CFR 50.47(b)(8) Planning Standard program
element, adequate emergency facilities and equipment to support the emergency
response are provided and maintained. The finding was assessed for significance in
accordance with NRC IMC 0609, Appendix B Emergency Preparedness Significance
Determination Process. Attachment 2 of Appendix B, Failure to Comply Significance
Logic is as follows: Failure to comply; Loss of Risk Significant Planning Standard
Function (RSPS), No; RSPS Degraded Function, No; Loss of Planning Standard
Function, No; the result is a Green finding. The inspectors determined that this resulted
in a very low safety significance finding (Green). No cross-cutting aspect was assigned
to this finding because the performance deficiency occurred more than three years ago
and is not reflective of current plant performance. (Section 4OA2.2)
B. Licensee-Identified Violations
Violations of very low safety significance that were identified by the licensee have been
reviewed by inspectors. Corrective actions taken or planned by the licensee have been
entered into the licensees CAP. These violations and corrective action tracking
numbers are listed in Section 4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Unit 1 began the inspection period at rated thermal power (RTP), and operated at or near full
power until July 22, 2012 when reactor power was lowered to 52 percent to clear a fouled
circulating water debris filter and power was returned to RTP on July 23, 2012. On August 3,
2012, power was reduced to 70 percent for a rod sequence exchange and power was returned
to RTP on August 5, 2012. On August 5, 2012, power was reduced to 90 percent for control rod
improvement and power was returned to RTP on the same day. On August 8, 2012, power was
reduced to 65 percent for offsite transmission line work and power was returned to RTP on the
same day. On September 16, 2012, the reactor was shut down for forced outage to replace the
1A and 1B recirculation pump seal assemblies. Reactor startup commenced on September 27,
2012 and the main generator was synchronized to the grid on September 28, 2012. Reactor
power was raised to RTP on September 29, 2012. On September 30, 2012 reactor power was
reduced to 75 percent for a scheduled control rod improvement. Power ascension continued to
RTP for the remainder of the inspection period.
Unit 2 began the inspection period at RTP, and operated at or near full power until August 18,
2012, when power was reduced to 70 percent for a rod sequence exchange and power was
returned to RTP on August 19, 2012. On August 20, 2012, power was reduced to 86 percent
for control rod improvement and power was returned to RTP on August 21, 2012. On August
21, 2012, power was reduced to 94 percent for control rod improvement and power was
returned to RTP on August 21, 2012. On September 29, 2012, reactor power was reduced to
94 percent to support a scheduled rod improvement and returned to RTP later that day and
maintained RTP for the remainder of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01 - 1 sample)
External Flooding
a. Inspection Scope
The inspectors evaluated the design, material condition, and procedures for coping with
the design basis probable maximum flood. The inspectors reviewed the Updated Final
Safety Analysis Report (UFSAR), which depicted the design flood levels and protection
areas containing safety-related equipment, to identify areas that may be affected by
external flooding. The inspectors conducted a site walk-down of the service water
building, to ensure that erected flood protection measures were in accordance with
design specifications. The inspectors reviewed the sealing of equipment below the flood
line, adequacy of watertight doors, drain systems and sumps including check valves,
and maintenance and calibration of flood protection equipment. The inspectors also
reviewed operating procedures for mitigating external flooding during severe weather to
5
determine if the licensee planned or established adequate measures to protect against
external flooding events.
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Quarterly Partial System Walk-downs (71111.04Q - 3 samples)
a. Inspection Scope
The inspectors performed partial system walk-downs of the following risk-significant
systems:
- Unit 2 A train Core Spray (CS) system while B residual heat removal/service
(RHR/SW) was inoperable for a system outage on July 11, 2012;
- Unit 1 Reactor Building Closed Cooling Water (RBCCW) on July 27, 2012; and
- Unit 1 B Standby Gas Treatment (SBGT) while the A SBGT was inoperable for a
maintenance outage on September 19, 2012.
The inspectors selected these systems based on their risk-significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could impact the function of the system, and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, UFSAR, Technical Specification (TS) requirements, outstanding work
orders, condition reports, and the impact of ongoing work activities on redundant trains
of equipment in order to identify conditions that could have rendered the systems
incapable of performing their intended functions. The inspectors also walked down
accessible portions of the systems to verify that system components and support
equipment were aligned correctly and were operable. The inspectors examined the
material condition of the components and observed operating parameters of equipment
to verify that there were no obvious deficiencies. The inspectors also verified that the
licensee had properly identified and resolved equipment alignment problems that could
cause initiating events or impact the capability of mitigating systems or barriers and
entered them into the CAP with the appropriate significance characterization.
b. Findings
No findings were identified.
.2 Semi-Annual Complete System Walk-down (71111.04S - 1 sample)
a. Inspection Scope
On September 5, 2012 the inspectors performed a complete system alignment
inspection of the Unit 1 RHR system to verify the functional capability of the system.
This system was selected because it was considered both safety-significant and risk-
6
significant in the licensees probabilistic risk assessment. The inspectors walked down
the system to review mechanical and electrical equipment line-ups, electrical power
availability, system pressure and temperature indications, as appropriate, component
labeling, component lubrication, component and equipment cooling, hangers and
supports, operability of support systems, and to ensure that ancillary equipment or
debris did not interfere with equipment operation. A review of a sample of past and
outstanding work orders (WOs) was performed to determine whether any deficiencies
significantly affected the system function. In addition, the inspectors reviewed the CAP
to ensure that system equipment alignment problems were being identified and
appropriately resolved.
b. Findings
No findings were identified.
1R05 Fire Protection (71111.05Q - 5 samples)
Quarterly Resident Inspector Tours
a. Inspection Scope
The inspectors conducted fire protection walk-downs which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- Unit 1 and 2 Control Buildings 23' Elevation 1PFP-CB-7;
- Unit 1 Reactor Building East 50 Elevation 1PFP-RB1-1h;
- Unit 1 Turbine Building South Area 38 Elevation 1PFP-TB1-1k;
- Unit 2 Reactor Building 50 Elevation 2PFP-RB2-1h; and
- Unit 2 Reactor Building North 2A Core Spray Room 2-PFP-RB2-1b.
The inspectors reviewed areas to assess if the licensee had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant, effectively maintained fire detection and suppression capability, maintained
passive fire protection features in good material condition, and had implemented
adequate compensatory measures for out-of-service, degraded or inoperable fire
protection equipment, systems, or features in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to impact equipment which could initiate or mitigate a
plant transient, or their impact on the plants ability to respond to a security event. Using
the documents listed in the attachment, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
fire detectors and sprinklers were unobstructed, that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees CAP.
7
b. Findings
No findings were identified.
1R06 Flood Protection Measures (71111.06 - 1 sample)
Annual Review of Cables Located in Underground Bunkers/Manholes
a. Inspection Scope
The inspectors conducted an inspection of underground bunkers/manholes subject to
flooding that contain cables whose failure could disable risk-significant equipment. The
inspectors performed walk-downs of risk-significant areas, including manhole 2-MH-
7SW, to verify that the cables were not submerged in water, that cables and/or splices
appear intact and to observe the condition of cable support structures. When applicable,
the inspectors verified proper dewatering device (sump pump) operation and verified
level alarm circuits are set appropriately to ensure that the cables will not be submerged.
Where dewatering devices were not installed; the inspectors ensured that drainage was
provided and was functioning properly.
b. Findings
No findings were identified.
1R07 Heat Sink Performance (71111.07T - 3 samples)
Triennial Review of Heat Sink Performance
a. Inspection Scope
The inspectors selected the Residual Heat Removal (RHR) Heat Exchanger 2A, Diesel
Generator (DG) 3 Jacket Water Cooler and the Core Spray (CS) Room Cooler 1A,
based on their risk-significance in the licensees probabilistic safety analysis and their
importance to safety-related mitigating system support functions in the NRCs model for
Brunswick Nuclear Power Plant, Units 1 and 2.
For the RHR Heat Exchanger 2A, DG 3 Jacket Water Cooler and the CS Room Cooler
1A, the inspectors reviewed the licensees inspection, maintenance, and monitoring of
biotic fouling and macro-fouling programs, to determine if they were adequate to ensure
proper heat transfer. This was accomplished by determining whether the methods used
were consistent with accepted industry practices. The inspectors also reviewed the
licensees inspection and cleaning activities had established acceptance criteria
consistent with industry standards, and the as-found results were recorded, evaluated,
and appropriately dispositioned to maintain structural integrity.
For the RHR Heat Exchanger 2A, DG 3 Jacket Water Cooler and the CS Room Cooler
1A, the inspectors reviewed the methods and results of heat exchanger performance
inspections. In addition, the inspectors reviewed the condition and operation of the RHR
Heat Exchanger 2A, DG 3 Jacket Water Cooler and the CS Room Cooler 1A to
8
determine if they were consistent with design assumptions in heat transfer calculations
and as described in the USFAR. This included determining whether the number of
plugged tubes was within pre-established limits based on capacity and heat transfer
assumptions. The inspectors also determined whether the licensee evaluated the
potential for water hammer and established adequate controls and operational limits to
prevent heat exchanger degradation due to excessive flow-induced vibration during
operation.
The inspectors determined whether the performance of the ultimate heat sink (UHS)-
Cape Fear River and its subcomponents such as piping, intake screens, pumps, valves,
etc. was appropriately evaluated by tests or other equivalent methods to ensure
availability and accessibility to the in-plant cooling water systems. The inspectors also
reviewed design changes to the service water system and the UHS.
The inspectors reviewed the licensees operation of the service water system and UHS.
This included a review of licensees procedures for a loss of the service water system or
UHS and the verification that instrumentation, which is relied upon for decision-making,
was available and functional. The inspectors also performed a system walk-down on the
service water system to determine whether the licensees assessment on structural
integrity was adequate and interviewed the respective system engineer. For buried or
inaccessible piping, the inspectors reviewed the licensees pipe testing, inspection, and
monitoring program to determine whether structural integrity was ensured and that any
leakage or degradation was appropriately identified and dispositioned by the licensee.
The inspectors performed a system walk-down of the service water intake structure to
determine whether the licensees assessment on structural integrity and component
functionality was adequate. The inspectors also determined whether service water
pump bay silt accumulation was monitored, trended, and maintained at an acceptable
level by the licensee, and that water level instruments were functional and routinely
monitored. The inspectors also determined whether the licensees ability to ensure
functionality during adverse weather conditions was adequate.
The inspectors reviewed condition reports related to the heat exchangers and heat sink
performance issues to determine whether the licensee had an appropriate threshold for
identifying issues and to evaluate the effectiveness of the corrective actions. Records
were also reviewed to verify that the licensee actions were consistent with Generic Letter
(GL) 89-13 licensee commitments, Electric Power Research Institute (EPRI) and other
industry guidelines. These inspection activities constituted three heat sink inspection
samples as defined in IP 71111.07-05.
b. Findings
No findings were identified.
9
1R11 Licensed Operator Requalification Program (71111.11Q - 2 samples)
.1 Quarterly Review of Licensed Operator Requalification Testing and Training
a. Inspection Scope
On August 13, 2012, the inspectors observed a crew of licensed operators in the plants
simulator during licensed operator requalification examinations to verify that operator
performance was adequate, evaluators were identifying and documenting crew
performance problems, and to ensure that training was being conducted in accordance
with licensee procedures. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms;
- correct use and implementation of abnormal and emergency procedures;
- control board manipulations;
- oversight and direction from supervisors; and
- ability to identify and implement appropriate TS actions and Emergency Plan actions
and notifications.
The crews performance in these areas was compared to pre-established operator action
expectations and successful critical task completion requirements.
b. Findings
No findings were identified.
.2 Quarterly Review of Licensed Operator Performance in the Main Control Room
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. Specifically, on September 16th, the inspectors
observed the Unit 1 shutdown and cooldown evolutions leading up to the forced outage
to repair the recirculation pump seals. The inspectors reviewed various licensee policies
and procedures listed in the Attachment.
- 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 and crew briefs
10
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 (71111.12Q - 3 samples)
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk-
significant systems:
- 1B Nuclear Service Water Pump smoking with vibration and strainer leakage on
pump start on June 26, 2012;
- 2A Standby Liquid Cooling accumulator failure before operability run on September
10, 2012 (AR560026); and
- Performance (unavailability and unreliability) history of the Severe Accident
Mitigation Alternatives (SAMA) diesels
The inspectors reviewed events where ineffective equipment maintenance may have
resulted in equipment failure or invalid automatic actuations of Engineered Safeguards
Systems and independently verified the licensee's actions to address system
performance or condition problems in terms of the following:
- implementing appropriate work practices;
- identifying and addressing common cause failures;
- scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;
- characterizing system reliability issues for performance;
- charging unavailability for performance;
- trending key parameters for condition monitoring; and
- ensuring 10 CFR 50.65(a)(1) or (a)(2) classification or re-classification; and verifying
appropriate performance criteria for structures, systems and components
(SSCs)/functions classified as (a)(2) or appropriate and adequate goals and
corrective actions for systems classified as (a)(1).
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the corrective action program with the appropriate
significance characterization.
b. Findings
No findings were identified.
11
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 4 samples)
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the
maintenance and emergent work activities affecting risk-significant equipment listed
below to verify that the appropriate risk assessments were performed prior to removing
equipment for work:
- Unit 2 yellow risk during emergent work on 2-E21-F015A, 2A Core Spray Full Flow
Test Bypass Valve, and scheduled maintenance on 2B RHR/residual heat removal
service water (RHRSW) on July 11, 2012;
- Unit 1 yellow risk during 1B Recirculation Pump Variable Frequency Drive power
recovery, and planned maintenance on 1A RHR/RHRSW on July 26, 2012;
- Unit 1 yellow risk during planned maintenance on 1B RHR/RHRSW September 4 to
September 6, 2012;
- Unit 1 integrated risk during repair of 1B recirculation pump seal September 17 to
September 25, 2012;
These activities were selected based on their potential risk-significance relative to the
reactor safety cornerstones. As applicable for each activity, the inspectors verified that
risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate
and complete. When emergent work was performed, the inspectors verified that the
plant risk was promptly reassessed and managed. The inspectors reviewed the scope
of maintenance work, discussed the results of the assessment with the licensee's
probabilistic risk analyst or shift technical advisor, and verified plant conditions were
consistent with the risk assessment. The inspectors also reviewed TS requirements and
walked down portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met.
b. Findings
No findings were identified.
1R15 Operability Evaluations (71111.15 - 5 samples)
a. Inspection Scope
The inspectors reviewed the following five issues:
- Unit 2 High Pressure Coolant Injection (HPCI) elevated thrust bearing temperature
on July 6, 2012 (AR548370);
(AR542025);
- Emergency Diesel Generator (EDG) #3 debris in bearing oil site glass on July 15,
2012 (AR549420);
- Reactor Building Close Cooling Water (RBCCW) piping corrosion in rattle space on
August 21, 2012 (AR557151); and
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- EDG #4 alternate safe shutdown switch contact continuity indications on August 27,
2012 (AR558810)
The inspectors selected these potential operability issues based on the risk-significance
of the associated components and systems. The inspectors evaluated the technical
adequacy of the evaluations to ensure that TS operability was properly justified and the
subject component or system remained available such that no unrecognized increase in
risk occurred. The inspectors compared the operability and design criteria in the
appropriate sections of the UFSAR and TS to the licensees evaluations, to determine
whether the components or systems were operable. Where compensatory measures
were required to maintain operability, the inspectors determined whether the measures
in place would function as intended and were properly controlled. The inspectors
determined, where appropriate, compliance with bounding limitations associated with the
evaluations. Additionally, the inspectors also reviewed a sampling of corrective action
documents to verify that the licensee was identifying and correcting any deficiencies
associated with operability evaluations.
b. Findings
No findings were identified.
1R18 Plant Modifications (71111.18 - 2 samples)
a. Inspection Scope
The inspectors reviewed the two modifications listed below to determine whether the
modifications affected the safety functions of systems that are important to safety. The
inspectors reviewed 10 CFR 50.59 documentation and post-modification testing results
and conducted field walk-downs of the modifications to verify that the modifications did
not degrade the design bases, licensing bases, and performance capability of the
affected systems.
- Design leak tight barriers at reactor building rattle spaces (EC86304);
- Service water building drain hub baffle plate installation (EC 88431)
b. Findings
No findings were identified.
1R19 Post Maintenance Testing (71111.19 - 7 samples)
a. Inspection Scope
The inspectors reviewed the following seven post-maintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- 0PT-12.2D, No. 4 Diesel Generator Monthly Load Test after replacement of the 60X
relay on July 23, 2012;
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- 0PT-08.1.4B, Residual Heat Removal (RHR) Service Water (SW) System Operability
Test - Unit 2 RHRSW Loop B after the maintenance outage on July 12, 2012;
- 0PT-08.2.2c, Low Pressure Coolant Injection/RHR System Operability Test - Unit 1
RHR Loop A after the maintenance outage on July 27, 2012;
August 16, 2012;
- 0PT-15.6, Standby Gas Treatment Operability Test, Unit 1 B after relay replacement
on August 15, 2012;
- 0PT-10.1.1, Reactor Core Isolation Cooling System Operability Test, Unit 2 after
replacement of Electronic Governor - Magnetic (EGM) on August 23, 2012; and
- 0PT-80.5, Reactor Pressure Vessel Pressure Test - Unit 1 after repair of 1B
recirculation pump seal on September 26, 2012
These activities were selected based upon the structure, system, or component's ability
to impact risk. The inspectors evaluated these activities for the following, as applicable:
the effect of testing on the plant had been adequately addressed; testing was adequate
for the maintenance performed; acceptance criteria were clear and demonstrated
operational readiness; test instrumentation was appropriate; tests were performed as
written in accordance with properly reviewed and approved procedures; equipment was
returned to its operational status following testing, and test documentation was properly
evaluated. The inspectors evaluated the activities against the UFSAR and TS to ensure
that the test results adequately ensured that the equipment met the licensing basis and
design requirements. In addition, the inspectors reviewed corrective action documents
associated with post-maintenance tests to determine whether the licensee was
identifying problems and entering them in the CAP and that the problems were being
corrected commensurate with their importance to safety.
b. Findings
No findings were identified.
1R20 Refueling and Other Outage Activities (71111.20 - 1 sample)
Other Outage Activities
a. Inspection Scope
The inspectors evaluated licensee outage activities for an unscheduled forced outage to
replace the 1B recirculation pump seal assembly. During the outage, the licensee made
the decision to replace the 1A recirculation pump seal assembly to address the potential
extent of cause/condition. The outage began on September 16, 2012 and concluded on
September 28, 2012. The inspectors reviewed activities to ensure that the licensee
considered risk in developing, planning, and implementing the outage schedule.
Additionally, the inspectors observed or reviewed the reactor shutdown and cool down,
outage equipment configuration and risk management, electrical lineups, control and
monitoring of decay heat removal, control of containment activities, performed a drywell
close out inspection, observed reactor startup and heat up activities, and identification
and resolution of problems associated with the outage. Documents reviewed are listed
in the Attachment.
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b. Findings
Introduction: The inspectors identified a Green NCV of TS 3.6.4.1, Secondary
Containment because the licensee did not maintain secondary containment operable as
required during an activity considered an operation with a potential for draining the
reactor vessel (OPDRV).
Description: On September 19, 2012, the licensee was replacing the 1B recirculation
pump seal assembly while Unit 1 was in Mode 4 (cold shutdown). In an effort to properly
isolate the work area, the recirculation suction and discharge isolation valves were
tagged closed. Due to seat leakage across the isolation valves, the 1B recirculation
pump drain valve was uncapped and opened to maintain the pump body partially empty
to prevent water from impacting the work area while the pump seal was removed. The
pump drain leakage was sent to the drywell floor drain system. The 1B recirculation
pump seal replacement activity had the potential to drain the reactor vessel below the
top of the fuel because the recirculation loops penetrate the reactor vessel below the top
of active fuel. An OPDRV is described in the licensees technical specifications as an
operation with a potential for draining the reactor vessel. However, the licensee did not
recognize or consider this activity as an OPDRV due to inadequate procedural guidance
that was used to exclude this activity as an OPDRV. Specifically, the licensee adopted
the definition of an OPDRV in procedure 0OI-01.01 as provided in Enforcement
Guidance Memorandum (EGM)11-003 as any activity that could potentially result in
draining or siphoning the RPV water level below the top of the fuel, without taking credit
for mitigating measures. However, section 9.16.15.b.(2) of licensee procedure 0OI-
01.01, BNP Conduct of Operations Supplement, stated leakage through mechanical
joints (for example valve or flange packing leaks, seat leakage through an isolation
valve, flange leakage, etc) is not considered an OPDRV. On September 19, 2012, the
licensee relaxed Unit 1 secondary containment from 03:30 a.m. until 09:20 p.m. by
opening the reactor building air lock doors on the 20-foot elevation to increase ventilation
to the recirculation pump seal replacement work area in the Unit 1 drywell. This resulted
in Secondary Containment inoperability while Unit 1 was in Mode 4 during an OPRDV
activity. The inspectors questioned the licensees Operations staff on the decision to
make secondary containment inoperable during an OPDRV activity. Following this, the
licensee restored secondary containment, developed an Operation standing instruction
12-052 to treat this activity as an OPDRV and placed this issue into its CAP as AR
562188.
Analysis: The inspectors determined that the failure to maintain secondary containment
operable while Unit 1 was in Mode 4 with an OPDRV in progress was a performance
deficiency. The performance deficiency was more than minor because it was associated
with the configuration control attribute of the Barrier Integrity Cornerstone, and adversely
affected the cornerstone objective to provide reasonable assurance that physical design
barriers (fuel cladding, reactor coolant system, and containment) protect the public from
radionuclide releases caused by accidents or events because the Unit 1 secondary
containment boundary was not preserved or maintained. The inspectors evaluated the
finding using Inspection Manual Chapter (IMC) 0609, Attachment 4, Phase 1 - Initial
Screening and Characterization of Findings, which required an analysis using IMC 0609
Appendix G since the reactor was in Mode 4 (cold shutdown). The finding was
determined to be of very low safety significance (Green) according to IMC 0609
15
Appendix G, Attachment 1, Checklist 6, since a quantitative assessment (Phase 2 or
Phase 3 evaluation) was not required. Specifically, the inspectors determined that the
licensee maintained adequate mitigation capability for reactor vessel water level
inventory and an event did not occur that could be characterized as a loss of control.
The cause of this finding was directly related to the cross-cutting aspect of Accurate
Procedures in the Resources component of the Human Performance area, because the
licensee did not consider the recirculation pump seal replacement activity to be OPDRV
based on procedural guidance that contains exclusions to what are considered OPDRV
activities. H.2(c)
Enforcement: Unit 1 TS 3.6.4.1, Secondary Containment, required secondary
containment to be operable during modes one, two, three, during movement of recently
irradiated fuel assemblies in the secondary containment and during operations with a
potential for draining the reactor vessel (OPDRVs). Contrary to the above, on
September 19, 2012, Unit 1 secondary containment was not maintained operable during
an OPDRV activity. The licensee entered this issue in its CAP as AR 562188, and
restored secondary containment during the OPDRV activity. Because the licensee
entered the issue into its CAP and the finding is of very low safety significance (Green),
this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRCs
Enforcement Policy: NCV 05000325/2012004-01, Failure to Maintain Secondary
Containment Operable during an OPDRV activity.
1R22 Surveillance Testing
.1 Routine Surveillance Testing (71111.22 - 4 samples)
a. Inspection Scope
The inspectors either observed surveillance tests or reviewed the test results for the
following activities to verify the tests met TS surveillance requirements, UFSAR
commitments, in-service testing requirements, and licensee procedural requirements.
The inspectors assessed the effectiveness of the tests in demonstrating that the SSCs
were operationally capable of performing their intended safety functions.
- 0PT-07.2.4A, Core Spray System Operability Test - Loop A on July 5, 2012;
- 0MST-RHR21Q, RHR-LPCI, CSS and HPCI Hi Drywell Pressure Trip Unit Inst Chan
Cal on July 10, 2012;
- 0MST-RCIC42R, RCIC Auto-actuation and Isolation Logic Sys Functional on July 24,
2012; and
- 0PT-12.12D, No. 4 Diesel Generator Monthly Load Test on August 17, 2012;
b. Findings
No findings were identified.
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.2 In-Service Testing (IST) Surveillance (71111.22 - 1 sample)
a. Inspection Scope
The inspectors reviewed the performance of Unit 1 LPCI/RHR System Operability Test -
Loop B on August 9, 2012 to evaluate the effectiveness of the licensees American
Society of Mechanical Engineers (ASME)Section XI testing program for determining
equipment availability and reliability. The inspectors evaluated selected portions of the
following areas: 1) testing procedures, 2) acceptance criteria, 3) testing methods, 4)
compliance with the licensees IST program, TS, selected licensee commitments, and
code requirements, 5) range and accuracy of test instruments, and 6) required corrective
actions.
b. Findings
No findings were identified.
.3 Reactor Coolant System Leak Detection Inspection Surveillance (71111.22 - 1 sample)
a. Inspection Scope
The inspectors observed and reviewed the test results for a reactor coolant system leak
detection surveillance, 0PT-80.5, Mid-Cycle Maintenance Outage Reactor Pressure
Vessel Pressure Test, on September 28, 2012. The inspectors observed in-plant
activities and reviewed procedures and associated records to determine whether:
effects of the testing were adequately addressed by control room personnel or engineers
prior to the commencement of the testing; acceptance criteria were clearly stated,
demonstrated operational readiness, and were consistent with the system design basis;
plant equipment calibration was correct, accurate, and properly documented; and the
calibration frequency was in accordance with TSs, the UFSAR, procedures, and
applicable commitments; applicable prerequisites described in the test procedures were
satisfied; test frequencies met TS requirements to demonstrate operability and reliability;
tests were performed in accordance with the test procedures and other applicable
procedures; and test data and results were accurate, complete, within limits, and valid.
Inspectors verified that test results not meeting acceptance criteria were addressed with
an adequate operability evaluation or the system or component was declared
inoperable; equipment was returned to a position or status required to support the
performance of its safety functions; and all problems identified during the testing were
appropriately documented and dispositioned in the corrective action program.
b. Findings
No findings were identified.
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1EP6 Emergency Planning Drill Evaluation (71114.06 - 2 samples)
a. Inspection Scope
The inspectors observed site emergency preparedness training drill/simulator scenarios
conducted on July 9, 2012 and July 25, 2012. The inspectors reviewed the drill scenario
narrative to identify the timing and location of classifications, notifications, and protective
action recommendations development activities. During the drill, the inspectors
assessed the adequacy of event classification and notification activities. The inspectors
observed portions of the licensees post-drill. The inspectors verified that the licensee
properly evaluated the drills performance with respect to performance indicators and
assessed drill performance with respect to drill objectives.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification (71151 - 6 samples)
.1 Mitigating Systems Cornerstone
a. Inspection Scope
- Mitigating Systems Performance Index, Residual Heat Removal - Unit 1
- Mitigating Systems Performance Index, Residual Heat Removal - Unit 2
The inspectors sampled licensee submittals for the Mitigating Systems Performance
Index (MSPI) performance indicators listed above for the period from the third (3rd)
quarter 2011 through the second (2nd) quarter 2012. The inspectors reviewed the
licensees operator narrative logs, issue reports, MSPI derivation reports, event reports
and NRC Integrated Inspection reports for the period to validate the accuracy of the
submittals.
b. Findings
No findings were identified.
.2 Barrier Integrity Cornerstone
a. Inspection Scope
- Reactor Coolant System (RCS) Specific Activity - Unit 1
- Reactor Coolant System (RCS) Specific Activity - Unit 2
The inspectors reviewed licensee submittals for the Reactor Coolant System Specific
Activity performance indicator for the period from the third (3rd) quarter 2011 through the
second (2nd) quarter 2012. The inspectors reviewed the licensees RCS chemistry
18
samples, TS requirements, issue reports, and event reports for the period to validate the
accuracy of the submittals. In addition to record reviews, the inspectors observed a
chemistry technician obtain and analyze a reactor coolant system sample.
- Reactor Coolant System Leakage - Unit 1
- Reactor Coolant System Leakage - Unit 2
The inspectors sampled licensee submittals for the Reactor Coolant System Leakage
performance indicator for the period from the third (3rd) quarter 2011 through the second
(2nd) quarter 2012. The inspectors reviewed the licensees operator logs, RCS leakage
tracking data, issue reports, and event reports for the period to validate the accuracy of
the submittals.
b. Findings
No findings were identified.
4OA2 Identification and Resolution of Problems (71152 - 2 samples)
.1 Routine Review of Items Entered Into the Corrective Action Program
a. Inspection Scope
To aid in the identification of repetitive equipment failures or specific human performance
issues for follow-up, the inspectors performed frequent screenings of items entered into
the licensees corrective action program. The review was accomplished by reviewing
daily action request reports.
b. Findings
No findings were identified.
.2 Assessments and Observations
Selected Issue Follow-up Inspection: UPS-A Failure and Loss of Emergency Response
Facility Information System (ERFIS), Plant Process Computer (PPC), Business Network
a. Inspection Scope
The inspectors selected AR 542704, UPS-A Failure and Loss of ERFIS, PPC, Business
Network, for detailed review. This AR identified that a single failure caused the loss of
ERFIS and Safety Parameter Display System (SPDS) on both units. The inspectors
reviewed the licensees CAP for ERFIS and SPDS failures in the past. The inspectors
reviewed these reports to verify that the licensee identified the full extent of the issue,
performed an appropriate evaluation, and specified and prioritized appropriate corrective
actions. The inspectors evaluated the reports against the requirements of the licensees
CAP as delineated in corporate procedure CAP-NGGC-0200, Corrective Action
Program, 10 CFR 50.47, and 10 CFR 50 Appendix E.
19
b. Findings
No findings were identified
a. Inspection Scope
The inspectors selected AR 542704, UPS-A Failure and Loss of ERFIS, PPC, Business
Network, for detailed review. This AR identified that a single failure caused the loss of
ERFIS and Safety Parameter Display System (SPDS) on both units. The inspectors
reviewed the licensees CAP for ERFIS and SPDS failures in the past. The inspectors
reviewed these reports to verify that the licensee identified the full extent of the issue,
performed an appropriate evaluation, and specified and prioritized appropriate corrective
actions. The inspectors evaluated the reports against the requirements of the licensees
CAP as delineated in corporate procedure CAP-NGGC-0200, Corrective Action
Program, 10 CFR 50.47, and 10 CFR 50 Appendix E.
b. Findings
Introduction: A self-revealing Green NCV of 10 CFR 50.54(q)(2) was identified for the
licensees failure to properly evaluate or consider the impact to emergency response
facilities of design change ESR98-00436 which was implemented in 1999. As a result,
a number of temporary losses of ERFIS, Emergency Response Data System (ERDS),
SPDS, and all displays including radiation monitors for the emergency response facilities
occurred. Specifically, the licensee failed to ensure that adequate emergency response
facilities and equipment were available as required by the Brunswick Nuclear Plant
Radiological Emergency Plan, Section 1.3.1.3, revision 80, and 10 CFR 50.47(b)(8).
This issue was captured in the licensees CAP as AR 542704.
Description: In 1999, the licensee implemented design change ESR98-00436 for the
power supply to the ERFIS, ERDS, SPDS, and all displays including RMS for the
emergency response facilities. The licensee did not properly evaluate or consider the
impact to emergency response facilities and equipment prior to implementation of this
design change. As a result, the ERFIS, ERDS, and SPDS systems, and all radiation
monitoring system (RMS) displays were susceptible to a single point power failure mode.
The implementation of the design change introduced a single point failure mode which
did not meet the design requirements specified in their Design Basis Document (DBD
60) sections 3.6.7.2 and 3.6.7.3. Prior to the licensees implementation of design
change ESR98-00436 in 1999, this single point vulnerability did not exist as the power
supply system had automatic switching capability on loss of one power source. When
the design change was implemented, the ERFIS, ERDS, and SPDS systems and RMS
displays were degraded as demonstrated by the resulting failures of those systems on
multiple occasions including July 17, 2004 and June 12, 2012. Additionally, all displays
for those systems were lost in all of the emergency facilities including the radiation
monitoring system.
20
On June 13, 2012, the licensee made an event notification to the NRC Operations
Center, 50.72(b)(3)(xiii) Loss of Emergency Assessment Capability, Offsite Response
Capability, or Offsite Communications Capability for the emergency response facilities.
The report delineated that at 5:57 p.m. EDT on June 12, 2012, Brunswick Nuclear Plant
experienced a fault on the Emergency Response Facility Information System (ERFIS)
uninterruptible power supply (UPS) electrical bus A. This resulted in a loss of site
Safety Parameter Display System (SPDS), Emergency Response Data System (ERDS)
and Plant Process Computer (PPC) for both Unit 1 and Unit 2.
During the loss of SPDS, the emergency response capability of that system was lost to
the site. During the loss of ERDS, the automatic data transfer feature of that system
was lost for transmissions to the NRC, however manual data transfer was still available.
During the loss of the PPC, automatic core thermal power averaging and automatic core
thermal limit monitoring was lost. Manual calculations were available for these functions.
Unit 1 SPDS was restored to the Emergency Operations Facility (EOF) at 7:49 p.m. on
June 12, 2012. Unit 2 SPDS was restored to the EOF at 8:30 p.m. on June 12, 2012.
The inverter was restored to service on June 17, 2012 at 12:00 noon.
Inspectors determined that the licensee did not properly evaluate or consider the impact
to all emergency response facilities and equipment prior to implementation of the
ESR98-00436 design change. The inspectors concluded that the ERFIS, ERDS, and
SPDS systems required by the Brunswick Nuclear Plant Radiological Emergency Plan
were degraded from 1999 when the design change was installed to present.
Compensatory measures were put in place during the June 2012 event to manually
obtain and log the required data from the instrumentation in the control room and
transmit to the emergency response facilities, and after the June 2012 event, the
licensee initiated a design change to restore the power configuration to those systems
back to the original design which would remove this failure mechanism.
Analysis: The licensees failure to properly evaluate or consider the impact to
emergency response facilities of design change ESR98-00436 which was implemented
in 1999 was a performance deficiency. Specifically, the licensee introduced a single
point failure mode which did not meet the design requirements specified in their Design
Basis Document (DBD 60) sections 3.6.7.2 and 3.6.7.3. This resulted in the licensees
failure to ensure that adequate emergency response facilities and equipment were
available as delineated in the Updated Final Safety Analysis Report (UFSAR) Section
7.7.1.9, and required by the Brunswick Nuclear Plant Radiological Emergency Plan,
Section 1.3.1.3, revision 80, and 10 CFR 50.47(b)(8).
The finding was more than minor because it adversely affected the Emergency
Preparedness Cornerstone objective of ensuring that the licensee was capable of
implementing adequate measures to protect the health and safety of the public in the
event of a radiological emergency. Specifically, the Facilities and Equipment attribute
was affected during the time when the ERFIS, ERDS, SPDS, and all displays including
radiation monitors for the emergency response facilities were degraded, and as a result
did not meet 10 CFR 50.47(b)(8) Planning Standard program element, adequate
emergency facilities and equipment to support the emergency response are provided
and maintained. The finding was assessed for significance in accordance with NRC IMC 0609, Appendix B Emergency Preparedness Significance Determination Process.
21
Attachment 2 of Appendix B, Failure to Comply Significance Logic is as follows: Failure
to comply; Loss of Risk Significant Planning Standard Function (RSPS), No; RSPS
Degraded Function, No; Loss of Planning Standard Function, No; the result is a Green
finding. The inspectors determined that this resulted in a low safety significance finding
(Green). No cross-cutting aspect was assigned to this finding because the performance
deficiency occurred more than three years ago and is not reflective of current plant
performance.
Enforcement: 10 CFR 50.54(q)(2) requires, in part, a licensee to follow and maintain the
effectiveness of an emergency plan that meets the requirements in Appendix E to this
part and, for nuclear power reactor licensee, the planning standards of 10 CFR 50.47(b).
The Brunswick Nuclear Plant Radiological Emergency Plan, Section 1.3.1.3, revision 80,
states in part that special provisions have been made to assure that ample space and
proper equipment are available to effectively respond to a full range of possible
emergencies. Contrary to the above, from 1999, when design change ESR98-00436
was installed, until the compensatory measures were put in place in June 2012, the
licensee failed to maintain adequate emergency facilities and equipment to support
emergency response when the ERFIS, ERDS, SPDS, and all displays including radiation
monitors for the emergency response facilities were degraded due to the implementation
of the design change. This resulted in failures of those systems on July 17, 2004 and
June 12, 2012. The licensee has compensatory measures in place, entered this issue
their CAP as AR 542704, and initiated a design change to restore the power
configuration back to the original design. Because the licensee entered the issue into its
CAP and the finding is of very low safety significance (Green), this violation is being
treated as an NCV, consistent with Section 2.3.2 of the NRCs Enforcement Policy: NCV
05000325; 324/2012004-02, Failure to Maintain Reliability and Availability of Emergency
Response Equipment for Emergency Response Facilities.
.3 Assessments and Observations
Selected Issue Follow-up Inspection: EDG 2 wiring associated with Alternate Safe
Shutdown (ASSD) Switch 2-DG-SS-A1
a. Inspection Scope
The inspectors performed a detailed review of AR 557897 associated with the wiring for
the EDG 2 Alternate Safe Shutdown (ASSD) Switch 2-DG-SS-A1. The issue was
discovered during a planned system outage for EDG2 during the week of August 26.
The inspectors verified that the issue was captured completely and accurately in the
CAP. The inspectors evaluated the licensees operability determinations and performed
walk-downs with licensee staff of applicable fire areas as needed. The inspectors
followed the licensees actions to restore the wiring to its proper configuration and also
verified the extent of condition inspections for the remaining EDGs 1, 3 and 4 were
completed in a timely manner. The inspectors reviewed the licensees reportability
evaluation and subsequent 8-hour report made to the NRC in accordance with 10 CFR
50.72(b)(3)(ii)(B). Additional documents reviewed are listed in the Attachment.
b. Findings
22
Introduction: The inspectors opened an unresolved item (URI) for this issue of concern
to determine if a performance deficiency existed.
Description: A wiring discrepancy was identified during inspection of the EDG 2 ASSD
switch 2-DG-SS-A1. A contact in the circuit was determined to be bypassed that would
have the potential to prevent proper isolation of the EDG2 control circuits from the Main
Control Room (MCR) during an Appendix R fire event. The inspectors plan to review the
licensees cause evaluation for this event and determine if a performance deficiency
existed. This issue is being tracked as URI 05000325; 324/2012004-03, EDG2 wiring on
ASSD switch.
4OA3 Follow-up of Events (71153 - 2 samples)
.1 Notice of Unusual Event for Fire in the Protected Area
a. Inspection Scope
For the plant event listed below, the inspectors reviewed plant parameters, reviewed
personnel performance, and evaluated performance of mitigating systems. The
inspectors communicated the plant events to appropriate regional NRC personnel, and
compared the event details with criteria contained in IMC 0309, Reactive Inspection
Decision Basis for Reactors, for consideration of potential reactive inspection activities.
As applicable, the inspectors verified that the licensee made appropriate emergency
classification assessments and properly reported the event in accordance with 10 CFR
50.72. The inspectors reviewed the licensees follow-up actions related to the events to
assure that the licensee implemented appropriate corrective actions commensurate with
their safety significance.
- On August 2, 2012, a fire existed in the protected area on the Units 1 and 2 turbine
building roof for approximately two hours, meeting the criteria for a Notice of Unusual
Event declaration.
b. Findings
One licensee identified violation is documented in Section 4OA7 of this report.
.2 (Closed) LER 05000325/2012-004-00, High Pressure Coolant Injection (HPCI)
Inoperable Due to Erratic Governor Operation
a. Inspection Scope
On May 2, 2012, Unit 1 HPCI was declared inoperable due to erratic governor operation
during Surveillance Test 0PT-09.2, HPCI System Operability Test. The erratic governor
operation was due to the failure of the Ramp Generator Signal Convertor (RGSC). The
licensee determined that the root cause of the RGSC failure was due to a lack of a
replacement preventative maintenance (PM) for the RGSC, which had been installed for
at least 22 years. The corrective actions included replacing the RGSC and creating a
PM task to replace the RGSCs. The licensee documented the root cause evaluation in
23
NCR 534364. The inspectors reviewed the LER, the NCR, and corrective actions to
determine whether the station adequately evaluated the condition.
b. Findings
One licensee identified violation is documented in Section 4OA7 of this report. This LER
is closed.
4OA5 Other Activities
.1 (Discussed) NRC Temporary Instruction (TI) 2515/187, Inspection of Near-Term Task
Force Recommendation 2.3 Flooding Walk-downs, and NRC TI 2515/188, Inspection of
Near-Term Task Force Recommendation 2.3 Seismic Walk-downs
a. Inspection Scope
Inspectors accompanied the licensee on a sampling basis, during their flooding and
seismic walk-downs, to verify that the licensees walk-down activities were conducted
using the methodology endorsed by the NRC. These walk-downs are being performed at
all sites in response to a letter from the NRC to licensees, entitled Request for
Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding
Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights
from the Fukushima Dai-Ichi Accident, dated March 12, 2012 (ADAMS Accession No.
Enclosure 3 of the March 12, 2012, letter requested licensees to perform seismic walk-
downs using an NRC-endorsed walk-down methodology. Electric Power Research
Institute (EPRI) document 1025286 titled, Seismic Walk-down Guidance, (ADAMS
Accession No. ML12188A031) provided the NRC-endorsed methodology for performing
seismic walk-downs to verify that plant features, credited in the current licensing basis
(CLB) for seismic events, are available, functional, and properly maintained.
Enclosure 4 of the letter requested licensees to perform external flooding walk-downs
using an NRC-endorsed walk-down methodology (ADAMS Accession No.
ML12056A050). Nuclear Energy Industry (NEI) document 12-07 titled, Guidelines for
Performing Verification Walk-downs of Plant Protection Features, (ADAMS Accession
No. ML12173A215) provided the NRC-endorsed methodology for assessing external
flood protection and mitigation capabilities to verify that plant features, credited in the
CLB for protection and mitigation from external flood events, are available, functional,
and properly maintained.
b. Findings
Findings or violations associated with the flooding and seismic walk-downs, if any, will
be documented in future reports.
24
.2 (Discussed) Temporary Instruction (TI) 2515/182 - Review of the Implementation of the
Industry Initiative to Control Degradation of Underground Piping and Tanks, Phase 1
a. Inspection Scope
Leakage from buried and underground pipes has resulted in ground water contamination
incidents with associated heightened NRC and public interest. The industry issued a
guidance document, Nuclear Energy Institute (NEI) 09-14, Guideline for the
Management of Buried Piping Integrity, (ADAMS Accession No. ML 1030901420), to
describe the goals and required actions (commitments made by the licensee) resulting
from this underground piping and tank initiative. On December 31, 2010, NEI issued
Revision 1 to NEI 09-14, Guidance for the Management of Underground Piping and
Tank Integrity, (ADAMS Accession No. ML 110700122), with an expanded scope of
components which included underground piping that was not in direct contact with the
soil and underground tanks. On November 17, 2011, the NRC issued TI-2515/182,
Review of the Industry Initiative to Control Degradation of Underground Piping and
Tanks, to gather information related to the industrys implementation of this initiative.
The instructors reviewed the licensees programs for buried pipe and underground piping
and tanks in accordance with TI-2515/182 to determine if the program attributes and
completion dates identified in Section 3.3 A and 3.3 B of NEI 09-14, Revision 1, were
contained in the licensees program and implementing procedures. For the buried pipe
and underground piping program attributes, with completion dates that had passed, the
inspectors reviewed records to determine if the attribute was in fact complete and to
determine if the attribute was accomplished in a manner which reflected good or poor
practices in management.
b. Observations
The licensees buried piping and underground piping and tanks program was inspected
in accordance with paragraphs 03.01.a through 03.01.c of TI-2515/182 and was found to
meet all applicable aspects of NEI 09-14 Revision 1, as set forth in Table 1 of the TI.
Based upon the scope of the review described above, Phase I of TI-2515/182 was
completed.
c. Findings
No findings were identified.
4OA6 Management Meetings
Exit Meeting Summary
On July 19, 2012, the inspectors presented inspection results of the triennial heat sink
inspection to Mr. Michael Annacone and other members of the licensee staff. The
25
inspectors confirmed that none of the potential report input discussed was considered
proprietary.
On September 18, 2012, the inspector presented inspection results of the TI-182, Phase
1 of the Underground Piping and Tanks Inspection by conference call to Mr. James
Burke, Site Director of Engineering, and other members of the licensee staff. The
inspector verified that all proprietary information was returned to the licensee.
On October 11, 2012, the inspectors presented inspection results from the quarterly
inspection to Mr. Annacone and other members of the licensee staff. The inspectors
confirmed that any proprietary information received during the inspection period were
properly controlled or returned to licensee staff.
4OA7 Licensee-Identified Violations
The following violations of very low significance (Green) were identified by the licensee
and are violations of NRC requirements which meet the criteria of the NRC Enforcement
Policy, for being dispositioned as NCVs.
- 10 CFR 50.54(q) requires, in part, a licensee authorized to possess and operate
a nuclear power reactor shall follow and maintain in effect emergency plans
which meet the standards of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4)
requires, in part, a standard emergency classification and action level scheme be
used by the licensee. Procedure 0PEP-02.1.1, Emergency Control - Notification
of Unusual Event, Alert, Site Area Emergency, and General Emergency, Step
5.7.2 states, that the emergency declaration will be made within 15 minutes after
the availability of indications to plant operators that an emergency action level
has been exceeded. Procedure 0PEP-02.1, Initial Emergency Actions, HU2.1,
requires the declaration of an Unusual Event when a fire is not extinguished
within 15 minutes of control room notification or verification of a control room fire
alarm in any Table H-1 or Table H-3 areas. Table H-1 includes the turbine
building. Contrary to the above, on August 2, 2012, a Notice of Unusual Event
(NOUE) was not classified within 15 minutes of a fire within the protected area
not being extinguished within 15 minutes of detection. Specifically, when a fire
was reported on the Turbine Building roof to the Control Room and was not
extinguished within 15 minutes, conditions were met for classification of EAL
HU2.1 in accordance with Procedure 0PEP-02.1; however, the EAL was not
classified until approximately eight hours after the fire started. This issue was
entered into the licensees CAP as NCR 552984 and the licensee is performing a
root cause evaluation. Corrective actions included making a one hour report to
the NRC for discovery of a condition that met the EAL classification for an NOUE
after the fact. The inspectors determined the finding was associated with an
actual event implementation problem, and assessed the significance using IMC 0609, Appendix B, "Emergency Preparedness Significance Determination
Process." Using the Emergency Preparedness SDP, Sheet 1, "Failure to
Implement (Actual Event) Significance Logic" the inspectors determined the
finding was of very low safety significance (Green) because the licensee failed to
implement a risk significant planning standard (10 CFR 50.47(b)(4)) during an
actual Notice of Unusual Event.
26
- 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
requires that activities affecting quality shall be prescribed by documented
instructions, procedures, or drawings, of a type appropriate to the circumstances
and shall be accomplished in accordance with these instructions, procedures, or
drawings. Licensee procedure ADM-NGGC-0107, Equipment Reliability Process
Guideline, steps 9.4.9 and 9.4.10 required component experts and preventive
maintenance (PM) optimization to determine if there was a cost effective PM to
prevent failure and then to develop the PM model. Contrary to the above, the
Unit 1 high pressure coolant injection (HPCI) ramp generator signal converter
(RGSC) did not have the appropriate preventive maintenance to prevent failure.
As a result, the Unit 1 high pressure coolant injection (HPCI) system failed the
HPCI System Operability Test performed on April 30, 2012 and was declared
inoperable. The licensee entered this issue into the CAP as NCR 534364.
Corrective actions included replacing the RGSC and creating a PM task to
replace the RGSCs on a specified frequency. Using IMC 0609, Appendix A,
"Phase 1 Initial Screening and Characterization of Findings," the inspectors
determined this finding required a Phase 2 analysis. The Phase 1 screened this
Mitigating Systems Cornerstone finding to Phase 2 because the finding
represented a loss of HPCI system and/or function. The inspectors, with the
assistance of the regional Senior Risk Analyst, performed a Phase 2 analysis
using the Saphire 8 Model. 109 hours0.00126 days <br />0.0303 hours <br />1.802249e-4 weeks <br />4.14745e-5 months <br /> of unavailability time was used for the
analysis since HPCI was not required during the refueling outage from February
23, 2012 through April 29, 2012. Based on the results of the Phase 2 analysis,
the inspectors determined the finding was of very low safety significance (Green).
ATTACHMENT: SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
M. Annacone, Site Vice President
A. Brittain, Manager - Security
J. Burke, Director - Site Engineering
K. Croker, Supervisor - Emergency Preparedness
C. Dunsmore, Manager - Shift Operations
P. Dubrouillet, Manager - Training
G. Galloway, Acting Manager, Nuclear Oversight
C. George, Manager - BOP Systems
S. Gordy, Manager - Maintenance
L. Grzeck, Manager - Regulatory Affairs
M. Hamm, Superintendent - Mechanical Maintenance
F. Jefferson, Manager - Reactor Systems Engineering
J. Kalamaja, Manager - Operations
J. Krakuszeski, Plant General Manager
R. Mosier, Communication Specialist
A. Padleckas, Superintendent - Nuclear Operations Performance
D. Petrusic, Superintendent - Environmental and Chemistry
A. Pope, Manager - Nuclear Support Services
J. Price, Manager- Design Engineering
W. Richardson, Engineering
T. Roeder, Supervisor - Chemistry
T. Sherrill, Licensing Senior Technical Specialist
P. Smith, Superintendent - Electrical, Instrumentation, and Controls Maintenance
M. Talon, Buried Piping Program Manager
J. Terrell, Corporate Buried Piping Program Manager
M. Turkal, Lead Engineer - Technical Support
J. Vincelli, Manager - Environmental and Radiological Controls
B. Wilder, Engineering
E. Wills, Director - Site Operations
NRC Personnel
R. Musser, Chief, Reactor Projects Branch 4, Division of Reactor Projects Region II
Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000325/2012004-01 NCV Failure to Maintain Secondary Containment Operable
During an OPDRV Activity. (Section 1R20)
05000325;324/2012004-02 NCV Failure to Maintain Reliability and Availability of
Emergency Response Equipment for Emergency
Response Facilities. (Section 4OA2.2)
Opened
05000325;324/2012004-03 URI EDG2 Wiring on ASSD Switch (Section 4OA2.3)
Closed
05000325/2012-004-00 LER High Pressure Coolant Injection (HPCI) Inoperable
Due to Erratic Governor Operation (Section 4OA3.2)
Discussed
Temporary Instruction TI Inspection of Near-Term Task Force Recommendation
2515/187 2.3 Flooding Walk-downs (Section 4OA5.1)
Temporary Instruction TI Inspection of Near-Term Task Force Recommendation
2515/188 2.3 Seismic Walk-downs (Section 4OA5.1)
Temporary Instruction TI Review of the Implementation of the Industry Initiative
2515/182 to Control Degradation of Underground Piping and
Tanks, Phase 1 (Section 4OA5.2)
Attachment
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
0AOP-13.0, Operation During Hurricane, Flood Conditions, Tornado, or Earthquake
0PEP-02.6, Severe Weather
2APP-UA-01, Annunciator Procedure for Panel UA-01
2APP-UA-28, Annunciator Procedure for Panel UA-28
2OP-43, Service Water System Operating Procedure
OPS-NGGC-1305, Operability Determinations
Nuclear Condition Reports
556860 556861 556862 556863 556864 556865
556866 556867 556868 556869 556870 557375
555023 545354 553946
Work Orders
550098 550100 550102 550015 545859 545861
1828825 1828826 1643223 1775054
Drawings
D-02778, Reactor Building Floor and Wall Sleeves Tabulation - Sheet No 1 Unit No 2
D-02779, Reactor Building Floor and Wall Sleeves Tabulation and Details - Sheet No 2
D-11597, Backdraft Damper with Extra Deep Frame
F-0424, Service Water Intake Structure Units 1 & 2 Ventilation System & Drainage Piping
LL-FB-02103, Reactor Building, Elevation -170, Fire Barrier Penetrations, RHR-HPCI Room
North Wall
Miscellaneous
0PIC-LS001, Omnitrol (Valrec) Level Control Switch Model 613, Single Actuator
DBD-106, Hazards Analysis
Engineering Change 80408R0, Flooding Design Basis Update
Individual Plant Examination for External Events Submittal, June 1995
Link Seal Vendor Manual
Quick Hit Self-Assessment 541666-15, Emergency Action Level Functionality
SD-43, Service Water System
URS List of Flood Features Inspected
URS Near Term Force Recommendations 2.3: Flooding, Project Number 30703-007
Section 1R04: Equipment Alignment
Procedures
Procedure 2OP-18, Core Spray System Operating Procedure
1OP-17, RHR System Operating Procedure
2OP-10, Standby Gas Treatment System Operating Procedure
Attachment
4
Drawings
D-25024, Reactor Building Core Spray System Piping Diagram
9527-D-2025, sheets 1A and 1B, RHR System, Unit 1
F-04073, Reactor Building Standby Gas Treatment Piping Diagram
Miscellaneous
DBD-10, Design Basis Document Standby Gas Treatment System
SD-10, System Description Standby Gas Treatment System
Section 1R05: Fire Protection
Procedures
0FPP-014, Control of Combustible, Transient Fire Loads, and Ignition Sources
0PFP-CB, Control Building Pre-Fire Plans
OPLP-01, Fire Protection Program Document
OPLP-01.2, Fire Protection System Operability, Action, and Surveillance Requirements
0PFP-013, General Fire Plan
1PFP-RB, Reactor Building Pre-Fire Plans Unit 1
2PFP-RB, Reactor Building Prefire Plans Unit 2
OPT-34.11.2.0, Portable Fire Extinguisher Inspection
1PFP-TB, Turbine Building Prefire plans
Section 1R06: Flood Protection
Nuclear Condition Reports
490292
Drawings
F-03347, East Yard Area - Units No. 1 & 2 Electrical Underground Duct Runs Manholes
F-03343, East Yard Area - Units No. 1 & 2 Electrical Underground Duct Runs Plan
Section 1R07: Heat Sink Performance
Procedures
0ENP-2704, Administrative Control of NRC Generic Letter 89-13 Requirements
0ENP-2705, Service Water Heat Exchanger Thermal Performance Testing
0PM-ACU500, Inspection and Cleaning of the RHR/Core Spray Room Aerofin Cooler Air Filters
and Coolers
0PM-STU500, Service Water Intake Structure Inspection and Cleaning
0CM-ENG521, Perfex Cooler Inspection and Repair
0E&RC-3212, Service/Circulating Water Chlorine Sampling
1PM-MEC502, Nuclear Service Water Header Inspection
1PM-MEC506, Conventional Service Water Header Inspection
2PM-MEC501, Nuclear Service Water Header Inspection
2PM-MEC505, Conventional Service Water Header Inspection
0PT-08.1.4a, RHR Service Water System Operability Test - Loop A
0AOP-18.0, Nuclear Service Water system Failure
0AOP-19-0, Conventional Service Water System Failure
Attachment
5
0AOP-37.1, Intake System Blockages
0O1-03.4, Unit 0 Outside Auxiliary Operator Daily Check Sheets
IPT-24.1-1, Service Water Pump and Discharge Valve Operability Test
0AI-81, Water Chemistry Guidelines
0A1-86, Service/Circulating Water Treatment Strategic Plan
0SMP-SW1500, Sodium Hypochlorite Injection to the SW System
Nuclear Condition Reports
392541 507589 339272 539775 497132 542399
Work Orders
01582632 01324149
Drawings
BN 43.0.01, Service Water System
Calculations
OSW-0096, Calculation for Tube Plugging and Fouling of Service Water Safety Related Heat
Exchangers
OSW-0097, RHR and Core Spray Room Cooler Performance
G0050C-04, Design Basis Heat Loads from Vital Heat Exchangers
Miscellaneous
LTAM-BNP-12-0009, Formal Water Hammer Analysis for Service Water
DBD-43, Service Water System
DBD-17, Residual Heat Removal System
System Health Report, Q1-2012, RBCCW Unit 1
System Health Report, Q1-2012, Service Water
System Health Report, Q1-2012, Emergency Diesel Generators
Program Health Report, Q1-2012, GL 89-13 Program
EC-84365, Temporary Removal of Degraded Coating on Internal Surfaces of Service Water
Pump Discharge Pipe Spools and Elbows
EC-85258, Replace Nuclear and Conventional Service Water Pump Discharge Elbow
2-E11-B002A, Final Eddy Current Inspection Report for RHR Heat Exchanger 2A,
March 15, 2011
EDG-3-JWC-2010, Final Eddy Current Inspection Report for EDG-3 Jacket Water Cooler
May 18, 2010
SD-63, Sodium Hypochlorite Injection System
Procedure Revision Requests
00549906 00549915 00549919 00549920 00549923 00549924
00550041 00550333
Section 1R11: Licensed Operator Requalification
Procedures
0PEP-2.1.1, Emergency Control - Notification of Unusual Event, Alert, Site Area Emergency, or
General Emergency
Attachment
6
0PEP-02.1, Initial Emergency Actions
AOP-17, Turbine Building Closed Cooling Water System
AOP-19, Conventional Service Water System Failure
EM-78, Nuclear Power Facility Emergency Notification Form
ENP-24.5, Reactivity Control Planning
2EOP-01-LPC, Level/Power Control
2EOP-01-RSP, Reactor Scram Procedure
OPS-NGGC-1000, Fleet Conduct of Operations
TRN-NGGC-0420, Conduct of Simulator Training and Evaluation
Miscellaneous
LORX-IPO-003 Scenario
Technical Specifications 3.7.1, Residual Heat Removal Service Water System
Technical Specifications 3.7.2.E, Service Water System and Ultimate Heat Sink
Section 1R12: Maintenance Effectiveness
Procedures
1OP-43, Service Water System Operating Procedure
MNT-NGGC-0001, Maintenance Rework Program
0PT-06.1, SLC System Operability Test
0AOP-36.2, Station Blackout
0PT-12.22, Load Test for SAMA Diesels
ADM-NGGC-0101, Maintenance Rule Program
Nuclear Condition Reports
546346 554488 549265 519703 477622 436705
436703 409663 408997 401149 477561 477622
401149
Work Orders
1802757 2104000 1868030 1746181
Drawings
Miscellaneous
FP-20234, R.P Adams CO, Inc, Strainers, Poro-Edge Automatic
Technical Specification 3.7.2, Service Water System and Ultimate Heat Sink
SD-05, Standby Liquid Control System
Maintenance Rule Unavailability Reports, January 2012 through August 2012
SAMA Diesels System Health Report, Q2-2012
Section 1R13: Maintenance Risk Assessment and Emergent Work Control
Procedures
0AI-144, Risk Management
0AP-022, BNP Outage Risk Management
0AP-025, BNP Integrated Scheduling
Attachment
7
ADM-NGGC-0006, Online EOOS Model
ADM-NGGC-0104, Work Management Process
WCP-NGGC-0500, Work Activity Integrated Risk Management Program
OPS-NGGC-1311, Protected Equipment
Nuclear Condition Reports
559242
Miscellaneous
BNP EOOS Risk Assessment
BNP EOOS Risk Assessment Report for Work Week 36
Section 1R15: Operability Evaluations
Procedures
0PT-12.2C, No. 3 Diesel Generator Monthly Load Test
FP-20322, Diesel Generator Instruction Manual
OPS-NGGC-1305, Operability Determinations
OPS-NGGC-1307, Operational Decision making
Nuclear Condition Reports
250203 310500 318607 548370 549420 558810
Work Orders
542970
Drawings
D-25028, Reactor Building Closed Cooling Water System
F-09348, Diesel Generator No. 4 Circuits Control Wiring Diagram
Miscellaneous
EDG 1-4 Generator Bearing Oil Analysis
SD-39, Emergency Diesel Generators
Section 1R18: Plant Modifications
Procedures
EGR-NGGC-0028 Engineering Evaluation
0AI-68 Brunswick Nuclear Plant Response to Severe Weather Warnings
Engineering Changes
EC 88431, Service Water Building Drain Hub Baffle Plate Installation
EC 86304, Design Leak Tight Barriers at Reactor Bldg Rattle Spaces
Nuclear Condition Reports
559173 490292
Attachment
8
Drawings
D-02041, Service Water System Piping Diagram
F-04024, Service Water Intake Structure Ventilation System & Draining Piping
F-01027, Seismic Isolation Space
Miscellaneous
UFSAR Updated Final Safety Analysis Report
Section 1R19: Post Maintenance Testing
Procedures
0PT-08.2.2C, LPCI/RHR System Operability Test
0PT-80.5, Mid-Cycle Maintenance Outage Reactor Pressure Vessel Pressure Test
Nuclear Condition Reports
551048
Work Orders
1951825 2028895 2034614 2112268
Drawings
D-25026, Sheet 2A, Residual Heat Removal System, Unit 1
Miscellaneous
Technical Specifications 3.5.1, Emergency Core Cooling System - Operating
Section 1R20: Outage Activities
Procedures
0GP-01, Prestartup Checklist
0GP-02, Approach to Criticality and Pressurization of the Reactor
0GP-03, Unit Startup and Synchronization
0GP-05, Unit Shutdown
0GP-10, Rod Sequence Checkoff Sheets
0AI-127, Primary Containment Inspection and Closeout
0AP-22, BNP Outage Risk Management
0OI-01-01, BNP Conduct of Operations Supplement
0SP-12-001, EGM 11-003 OPDRV Activities
Nuclear Condition Reports
561831 561899 561173 562188
Drawings
D-20022 Sheet 1, Piping Diagram Extraction Steam System, Unit 1
Miscellaneous
Main Control Room (MCR) Logs
Outage Control Center (OCC) Logs
Attachment
9
Unit 1 Key Safety Function Component Status Sheets
Operations Standing Instruction 12-052
Section 1R22: Surveillance Testing
Procedures
0PT-07.2.4a, Core Spray System Operability Test - Loop A
0MST-RHR21Q, CSS and HPCI Hi Drywell Pressure Trip Unit Chan Cal
0MST-RCIC42R, RCIC Auto-actuation and Isolation Logic Sys Functional
0PT-12.12D, No. 4 Diesel Generator Monthly Load Test
0PT-08.2.2B, LPCI/RHR System Operability Test - Loop B
0PT-80.5, Mid-Cycle Maintenance Outage Reactor Pressure Vessel Pressure Test
Nuclear Condition Reports
547945
Work Orders
2107649
Drawings
D-25024, Reactor Building Core Spray System Piping Diagram
Miscellaneous
Technical Specification 3.5.1, Emergency Core Cooling System - Operating
UFSAR Section 6.3.3.7, Lag Times
Section 1EP6: Drill Evaluation
Procedures
0PEP-2.1.1, Emergency Control - Notification of Unusual Event, Alert, Site Area Emergency, or
General Emergency
0PEP-02.1, Initial Emergency Actions
0PEP-02.6.20, Dose Projection Coordinator
0PEP-03.4.8, Offsite Dose Projections for Monitored Releases
2EOP-01-RSP, Reactor Scram Procedure
EM-78, Nuclear Power Facility Emergency Notification Form
EMG-NGGC-0002, Offsite-Dose Assessment
OPS-NGGC-1000, Fleet Conduct of Operations
Nuclear Condition Reports
551255 551620 551698 552439
Section 4OA1: Performance Indicator Verification
Procedures
0E&RC-1006, Operation of the Reactor Building Sample Stations
0E&RC-2212, Calibration/Operation of Genie Gamma Spectroscopy System
REG-NGGC-0009, NRC Performance Indicators and Monthly Operating Report Data
Attachment
10
Miscellaneous
BNP-PSA-069, NRC Mitigating System Performance Index (MSPI) Basis Document
Unit 1 RHR MSPI Margin Reports, July 2011 to June 2012
Unit 2 RHR MSPI Margin Reports, July 2011 to June 2012
Unit 1 RHR MSPI Derivation Reports, July 2011 to June 2012
Unit 2 RHR MSPI Derivation Reports, July 2011 to June 2012
REG-NGGC-0009, Attachment 4 - MSPI Unavailability Data Sheets, July 2011 to June 2012
REG-NGGC-0009, Attachment 6 - MSPI Unreliability Data Sheets, July 2011 to June 2012
Section 4OA2: Identification and Resolution of Problems
Procedures
CAP-NGGC-0200, Condition Identification and Screening Process
CAP-NGGC-0205, Condition Evaluation and Corrective Action Process
CAP-NGGC-0206, Performance Assessment and Trending
OERP, Radiological Emergency Response Plan
OPLP-37, Equipment Important to Emergency Preparedness and ERO Response
OPEP-02.6.21, Emergency Communicator
OPEP-04.2, Emergency Facilities and Equipment
ADM-NGGC-0119, Nuclear Safety Culture Program, Revision 01
Nuclear Condition Reports
AR 00201153, Adverse Trend - Failed ERFIS Multiplexer Modules
ACE CR 542704, UPS-A Failure and Loss of ERFIS, PPC, Business Network
Miscellaneous
Down Time by Computer System Log
NIT Key performance indicators
ESR 98-00436, RAINS 99-0045, 50.59 Evaluation
ESR 98-00436, RAINS 99-0045, 50.54q Evaluation
Section 4OA3: Event Followup
Procedures
0PT-09.2, HPCI System Operability Test
0PT-09.3, HPCI System - 165 PSIG Flow Test
ADM-NGGC-0107, Equipment Reliability Process Guideline
0PEP-02.1, Initial Emergency Actions
0PEP-02.1.1, Emergency Control - Notification of Unusual Event, Alert, Site Area Emergency,
and General Emergency
0PEP-02.2.1, Emergency Action Level Bases
Nuclear Condition Reports
534364 552815 552984
Work Orders
2107224 2107264 2107271 2107313
Attachment
11
Drawings
1-FP-02039, General Electric Gas Control Piping Diagram
D-02055, Piping Diagram, Carbon Dioxide & Hydrogen Systems, Units 1 & 2
Miscellaneous
10 CFR 50.72 Event Report 47893, High Pressure Coolant Injection Inoperable due to Erratic
Governor Operation, May 2, 2012
LER 1-2012-004-00, High Pressure Coolant Injection Inoperable due to Erratic Governor
Operation, June 29, 2012
System Description 19, High Pressure Coolant Injection System
Technical Specification 3.5.1, Emergency Core Cooling Systems and Reactor Core Isolation
Cooling
Event Notification, Discovery of a Condition that Met the EAL Classification of an Unusual Event
(After-the-Fact), August 2, 2012
NUREG-1022, Event Reporting Guidelines
Operator Logs, August 2, 2012
SD-59, Hydrogen Water Chemistry System
Section 4OA5: Other Activities
Procedures
EGR-NGGC-0209, Buried Piping Program, Rev. 3
EGR-NGGC-0513, License Renewal Buried Piping and Tanks Inspection Program, Rev. 3
0AOP-13.0, Operation During Hurricane, Flood Conditions, Tornado, or Earthquake
0PEP-02.6, Severe Weather
2APP-UA-01, Annunciator Procedure for Panel UA-01
2APP-UA-28, Annunciator Procedure for Panel UA-28
2OP-43, Service Water System Operating Procedure
OPS-NGGC-1305, Operability Determinations
MNT-NGGC-004, Scaffolding Control
0PT-34.2.2.1, Fire Door, Pressure Boundary Door, ASSD Access/Egress Door, and Severe
Weather/Flood Control Door Inspections
0AI-68, Brunswick Nuclear Plant Response to Severe Weather Warnings
0PEP-02.1.1, Emergency Control-Notification of Unusual Event, Alert, Site Area Emergency,
and General Emergency
0PEP-02.6, Severe Weather
0AOP-13.0, Operation During Hurricane, Flood Conditions, Tornado, or Earthquake
Nuclear Condition Reports
551646 551838 551964 550469 559173 556860
556861 556862 556863 556864 556865 556866
556867 556868 556869 556870 557375 555023
545354 553946
Work Orders
550098 550100 550102 550015 545859 545861
1828825 11828826 1643223 1775054 2113607
Attachment
12
Work Requests
546632 546540 546541 546543 544971 546174
546823 546824 546203 546274 546278
Drawings
D-11099, Reactor Building Miscellaneous Steel Pool Liners
D-2274, Diesel Cooling Water
D-25049, Reactor Building Piping Diagram Fuel Pool Cooling & Filtering System, Unit 1
D-26007, Reactor Building Fuel Pool Cooling & Filter System Plan EL 80-0 & Sections
D-26009, Reactor Building Fuel Pool Cooling & Filter System Miscellaneous Plans & Sections
D-27010, Supplemental Spent Fuel Pool Cooling System
F-25008, Reactor Building Arrangement & Details, Fuel Pool
D-02778, Reactor Building Floor and Wall Sleeves Tabulation - Sheet No 1 Unit No 2
D-02779, Reactor Building Floor and Wall Sleeves Tabulation and Details - Sheet No 2
D-11597, Backdraft Damper with Extra Deep Frame
F-0424, Service Water Intake Structure Units 1 & 2 Ventilation System & Drainage Piping
LL-FB-02103, Reactor Building, Elevation -170, Fire Barrier Penetrations, RHR-HPCI Room
North Wall
1-FP-09319, Reactor Building Railroad Doors
Corrective Action Document
PRR 562261, Revise EGR-NGGC-0209 to strengthen the tie to the License Renewal Program
Miscellaneous
Calculation 2RB2-0012, Analysis for Spent Fuel Pool - Elevation of Top of Active Fuel
Engineering Change 80408R0, Flooding Design Basis Update
EPRI Report 1025286, Seismic Walk-down Guidance for Resolution of Fukushima Near-Term
Task Force Recommendation 2.3: Seismic
FP-75090, International Instruments INC, Instruments, Switchboard, Edgewise
System Description SD-43, Service Water System
UFSAR Section 9.1.3.3, Fuel Pool Cooling and Cleanup System, Safety Evaluation
Units 1 and 2, Flood Protection Feature 6BL, Service Water Building, 4 Elevation, Pipe
Penetration Seal\20-8 Pipe Sleeves
Unit 1, SWEL 1 List
Unit 1, SWEL 2 List
Unit 2, SWEL 1 List
Unit 2, SWEL 2 List
URS Post Fukushima Project, NTTF Recommendation 2.3 Seismic Walk-down Training Record
URS Project Number 30703-007, Near Term Task Force Recommendation 2.3 Seismic Walk-
down Procedure
0PIC-LS001, Omnitrol (Valrec) Level Control Switch Model 613, Single Actuator
DBD-106, Hazards Analysis
Engineering Change 80408R0, Flooding Design Basis Update
Individual Plant Examination for External Events Submittal, June 1995
Link Seal Vendor Manual
Quick Hit Self-Assessment 541666-15, Emergency Action Level Functionality
SD-43, Service Water System
Attachment
13
URS List of Flood Features Inspected
URS Near Term Force Recommendations 2.3: Flooding, Project Number 30703-007
Report Number 110311.401, Summary of Progress Energy Fleet Underground Piping and
Tanks with the Scope of NEI 09-14 (Rev. 1), prepared by Structural Integrity Associates,
Inc., dated 12/07/2011
Assessment Number 531636, Quick Hit Self Assessment for HNP and BNP Buried Piping
Program and the NRC TI-2515/182 Inspection, 08/15/2012
Specification 024-001 for Special Doors
Section 4OA7: Licensee-Identified Violations
Procedures
0PEP-02.1, Initial Emergency Actions
0PEP-02.1.1, Emergency Control - Notification of Unusual Event, Alert, Site Area Emergency,
and General Emergency
0PEP-02.2.1, Emergency Action Level Bases
Nuclear Condition Reports
552815 552984
Drawings
1-FP-02039, General Electric Gas Control Piping Diagram
D-02055, Piping Diagram, Carbon Dioxide & Hydrogen Systems, Units 1 & 2
Miscellaneous
Event Notification, Discovery of a Condition that Met the EAL Classification of an Unusual Event
(After-the-Fact), August 2, 2012
NUREG-1022, Event Reporting Guidelines
Operator Logs, August 2, 2012
SD-59, Hydrogen Water Chemistry System
Attachment