ML13032A277
| ML13032A277 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 02/01/2013 |
| From: | Frank Ehrhardt NRC/RGN-II/DRP/RPB2 |
| To: | Tynan T Southern Nuclear Operating Co |
| References | |
| IR-12-005 | |
| Download: ML13032A277 (51) | |
See also: IR 05000424/2012005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
245 PEACHTREE CENTER AVENUE NE, SUITE 1200
ATLANTA, GEORGIA 30303-1257
February 1, 2013
Mr. Tom E. Tynan
Vice President - Vogtle
Southern Nuclear Operating Company, Inc.
Vogtle Electric Generating Plant
7821 River Road
Waynesboro, GA 30830
SUBJECT: VOGTLE ELECTRIC GENERATING PLANT - NRC INTEGRATED INSPECTION
REPORT 05000424/2012005 AND 05000425/2012005
Dear Mr. Tynan:
On December 31, 2012, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Vogtle Electric Generating Plant, Units 1 and 2. The enclosed integrated
inspection report documents the inspection findings, which were discussed on January 11,
2013, 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.
Three self-revealing findings of very low safety significance (Green) were identified during this
inspection. These findings were determined to involve violations of NRC requirements. Two
licensee-identified violations which were determined to be of very low safety significance are
listed in this report. The NRC is treating these violations as non-cited violations (NCVs)
consistent with Section 2.3.2 of the Enforcement Policy.
If you contest 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 Vogtle
Electric Generating Plant.
If you disagree with the cross-cutting aspects assigned 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
Vogtle Electric Generating Plant.
T. Tynan 2
In accordance with the 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosures, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of the
NRCs Agencywide Document Access and Management 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/
Frank Ehrhardt, Chief
Reactor Projects Branch 2
Division of Reactor Projects
Docket Nos.: 05000424, 05000425
License Nos.: NPF-68 and NPF-81
Enclosures: Inspection Report 05000424/2012005 and 05000425/2012005
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
T. Tynan 2
In accordance with the 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosures, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of the
NRCs Agencywide Document Access and Management 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/
Frank Ehrhardt, Chief
Reactor Projects Branch 2
Division of Reactor Projects
Docket Nos.: 05000424, 05000425
License Nos.: NPF-68 and NPF-81
Enclosures: Inspection Report 05000424/2012005 and 05000425/2012005
w/Attachment: Supplemental Information
cc w/encl: (See page 3)
X PUBLICLY AVAILABLE G NON-PUBLICLY AVAILABLE G SENSITIVE X NON-SENSITIVE
ADAMS: G Yes ACCESSION NUMBER:_________________________ X SUNSI REVIEW COMPLETE G FORM 665 ATTACHED
OFFICE RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS RII:DRS RII:DRS
SIGNATURE Via email Via email Via email Via email Via email Via email Via email
NAME MCain TChandler BCaballero ANielsen CDykes WPursley JLaughlin
DATE 01/25/2013 02/31/2013 01/25/2013 01/28/2013 01/28/2013 01/29/2013 01/28/2013
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRP RII:DRP
SIGNATURE MOM /RA/ FJE /RA/
NAME MMiller FEhrhardt
DATE 01/20/2013 02/01/2013 2/ /2013 2/ /2013 2/ /2013 2/ /2013 2/ /2013
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICIAL RECORD COPY DOCUMENT NAME: G:\DRPII\RPB2\VOGTLE\REPORTS\2012\2012-005\VOGTLE
2012005.DOCX
T. Tynan 3
cc w/encl: Dennis R. Madison
C. Russ Dedrickson Vice President
Fleet Support Supervisor Edwin I. Hatch Nuclear Plant
Southern Nuclear Operating Company, Inc. Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution Electronic Mail Distribution
S. Kuczynski Leigh Perry
Chairman, President and CEO SVP & General Counsel-Ops & SNC
Southern Nuclear Operating Company, Inc. Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution Electronic Mail Distribution
Todd L. Youngblood T. E. Tynan
Vice President Site Vice President
Fleet Oversight Vogtle Electric Generating Plant
Southern Nuclear Operating Company, Inc. Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution Electronic Mail Distribution
W. L. Bargeron M. J. Ajluni
Plant Manager Nuclear Licensing Director
Vogtle Electric Generating Plant Southern Nuclear Operating Company, Inc.
Southern Nuclear Operating Company, Inc. Electronic Mail Distribution
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B. D. McKinney, Jr.
D. G. Bost Regulatory Response Manager
Chief Nuclear Officer Southern Nuclear Operating Company, Inc.
Southern Nuclear Operating Company, Inc. Electronic Mail Distribution
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D. W. Daughhetee
N. J. Stringfellow Licensing Engineer
Licensing Manager Southern Nuclear Operating Company, Inc.
Southern Nuclear Operating Company, Inc. Electronic Mail Distribution
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Bradley J. Adams
Paula Marino Vice President
Vice President Fleet Operations Support
Engineering Southern Nuclear Operating Company, Inc.
Southern Nuclear Operating Company, Inc. Electronic Mail Distribution
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T. D. Honeycutt
T. A. Lynch Regulatory Response Supervisor
Vice President Southern Nuclear Operating Company, Inc.
Joseph M. Farley Nuclear Plant Electronic Mail Distribution
Southern Nuclear Operating Company, Inc.
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Arthur H. Domby, Esq.
Troutman Sanders
Electronic Mail Distribution
T. Tynan 4
(cc w/encl continued) Richard Haynes
Director, Division of Waste Management
L. P. Hill Bureau of Land and Waste Management
Licensing Supervisor S.C. Department of Health and
Southern Nuclear Operating Company, Inc. Environmental Control
Electronic Mail Distribution 2600 Bull Street
Columbia, SC 29201
L. L. Crumpton
Administrative Assistant, Sr. Lee Foley
Southern Nuclear Operating Company, Inc. Manager of Contracts Generation
Electronic Mail Distribution Oglethorpe Power Corporation
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Hickox, T. Mark
Vogtle Electric Generating Plant Mark Williams
Electronic Mail Distribution Commissioner
Georgia Department of Natural Resources
Mark Rauckhorst Electronic Mail Distribution
Site Vice President
Vogtle Units 3 and 4 Chuck Mueller
Southern Nuclear Operating Company Manager
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Georgia Department of Natural Resources
S. C. Swanson Electronic Mail Distribution
Site Support Manager
Vogtle Electric Generating Plant Cynthia A. Sanders
Electronic Mail Distribution Radioactive Materials Program Manager
Environmental Protection Division
Jerry Ranalli Georgia Department of Natural Resources
Municipal Electric Authority of Georgia Electronic Mail Distribution
Power
Electronic Mail Distribution James C. Hardeman
Environmental Radiation Program Manager
Sandra Threatt, Manager Environmental Protection Division
Nuclear Response and Emergency Georgia Department of Natural Resources
Environmental Surveillance Electronic Mail Distribution
Bureau of Land and Waste Management
Department of Health and Environmental Mr. Steven M. Jackson
Control Senior Engineer - Power Supply
Electronic Mail Distribution Municipal Electric Authority of Georgia
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Division of Radiological Health
TN Dept. of Environment & Conservation Reece McAlister
401 Church Street Executive Secretary
Nashville, TN 37243-1532 Georgia Public Service Commission
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Office of the Attorney General
40 Capitol Square, SW (cc w/encl continued next page)
Atlanta, GA 30334
T. Tynan 5
(cc w/encl continued)
Office of the County Commissioner
Burke County Commission
Electronic Mail Distribution
Director
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Governor's Office of Consumer Affairs
2 M. L. King, Jr. Drive
Plaza Level East; Suite 356
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Amy Whaley
Resident Manager
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T. Tynan 6
Letter to Tom E. Tynan from Frank Ehrhardt dated February 1, 2013
SUBJECT: VOGTLE ELECTRIC GENERATING PLANT - NRC INTEGRATED INSPECTION
REPORT 05000424/2012005 AND 05000425/2012005
Distribution w/encl:
C. Evans, RII
L. Douglas, RII
OE Mail
RIDSNRRDIRS
PUBLIC
RidsNrrPMVogtle Resource
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-424, 50-425
Report Nos.: 05000424/2012005 and 05000425/2012005
Licensee: Southern Nuclear Operating Company, Inc. (SNC)
Facility: Vogtle Electric Generating Plant, Units 1 and 2
Location: Waynesboro, GA 30830
Dates: October 1, 2012 through December 31, 2012
Inspectors: M. Cain, Senior Resident Inspector
T. Chandler, Resident Inspector
B. Caballero, Senior Operations Engineer
A. Nielsen Senior Health Physicist
C. Dykes Health Physicist
W. Pursley Health Physicist
J. Laughlin Emergency Preparedness Inspector
R. Williams Reactor Inspector
A. Vargas-Mendez Reactor Inspector
Approved by: Frank Ehrhardt, Chief
Reactor Projects Branch 2
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000424/2012-005, 05000425/2012-005; 10/01/2012 - 12/31/2012; Vogtle Electric
Generating Plant, Units 1 and 2; Post-Maintenance Testing, Identification and Resolution
of Problems
The report covered a three-month period of inspection by the resident inspectors. Three
non-cited violations (NCVs) with very low safety significance (Green) were identified.
The significance of inspection findings are indicated by their color (i.e., great than Green,
or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination Process dated June 19, 2012. Cross-cutting aspects
are determined using IMC 0310, Components Within The Cross-Cutting Areas dated
October 28, 2011. All violations of NRC requirements are dispositioned in accordance
with the NRCs Enforcement Policy dated June 7, 2012. The NRCs program for
overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process revision 3.
Cornerstone: Mitigating Systems
Green: A self-revealing non-cited violation (NCV) for failure to meet the requirements of
10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings was
identified for failure to provide adequate work instructions in the maintenance procedure
used to modify 480V EMAX safety-related switchgear breakers. Specifically, when
modifying the breaker by adding a higher amperage closing coil, failure to verify the
proper placement of the wire bundle on top of the closing coil following replacement
resulted in the safety-related 2A NSW cooling tower fan #3 480V EMAX breaker failing
to close when demanded. The licensee replaced the failed breaker and returned the fan
to operable status within 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
The finding was more than minor because it impacted the reactor safety mitigating
systems cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences and
affected the cornerstone attribute of equipment performance. Since the inspectors
answered No to all of the IMC 0609.04 (dated June 19, 2012) Exhibit 2, section A,
questions 1-4, mitigating systems cornerstone screening questions, the inspectors
concluded that the finding was of very low safety significance (Green). The cause of this
finding was related to the corrective action program component of the problem
identification and resolution cross-cutting area due to less-than-adequate problem
evaluation. P.1(c) Specifically, the licensees extent of cause evaluations performed on
previous 480V EMAX breaker failures (caused by restricted movement of the close
lever) did not identify the potential of the closing coil wire bundle to interfere with the
proper movement of the close lever. The licensees corrective action to the 480V EMAX
breaker issue was to revise the maintenance procedure used to perform maintenance on
EMAX breakers (procedure 28480-C, 480V EMAX Breaker Maintenance), and then
inspect all 480V EMAX breakers on site that had been modified with the higher
amperage closing coil to verify that the wire bundles were not interfering with the
operation of the close coil. The licensee entered this into their corrective action program
as CR 549999 and CR 550736. (Section 1R19)
Enclosure
3
Cornerstone: Initiating Events
Green: A self-revealing non-cited violation (NCV) for failure to meet the requirements of
10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings was
identified for failure to provide adequate work instructions in the operations and
maintenance procedures used to open main steam isolation valves (MSIVs) that were
bound in their closed seat. Specifically, the operations and maintenance procedures
used to open the loop 2 and loop 3 outboard MSIVs did not provide instructions to limit
the magnitude of the force applied to the valve stems while attempting to open the
valves, which ultimately resulted in the brittle failure of the valve stems. The licensee
conducted ultrasonic testing of the remaining six Unit 1 MSIVs to verify that the valve
stems were intact. The two failed valve stems were replaced, and the reactor was
restarted nine days later.
The finding was more than minor because it was associated with the procedure quality
attribute of the reactor safety - initiating events cornerstone and it adversely affected the
cornerstone objective to limit the likelihood of events that upset plant stability and
challenge critical safety functions during shutdown as well as power operations.
Specifically, the failure to provide adequate work instructions to operations and
maintenance personnel resulted in the failure of both the loop 2 and loop 3 outboard
MSIVs and the subsequent manual reactor trip. Since the inspectors answered no to
the Exhibit 1, section B, initiating events screening question, the inspectors concluded
that the finding was of very low safety significance (Green). The cause of the finding
was related to the work control component of the human performance cross-cutting area
due to less-than-adequate work planning. H.3(a) Specifically, the licensees
procedures used to open the MSIVs that were stuck on their closed seat did not contain
instructions or precautions to limit the magnitude of the force applied to the valve stems
while attempting to open the valves. The licensee entered this issue into their corrective
action program as CR 530916. (Section 4OA2)
Cornerstone: Occupational Radiation Safety
Green: The inspectors identified a Green, self-revealing, Non-cited Violation of technical specification 5.7.1, High Radiation Area, for an unauthorized entry into a High
Radiation Area (HRA). A maintenance worker entered a HRA in Unit 1 containment
without being briefed on the radiological conditions. The licensee entered this issue into
their corrective action program as CR 523976 and took immediate corrective actions
including an outage work crew stand down.
This finding was more than minor because it was associated with the occupational
radiation safety cornerstone attribute of human performance and adversely affects the
cornerstone objective of ensuring adequate protection of worker health and safety from
exposure to radiation from radioactive material during routine civilian nuclear reactor
operation. The finding was evaluated using the occupational radiation safety
significance determination process. The finding was not related to As Low As
Reasonably Achievable (ALARA) planning, nor did it involve an overexposure or
substantial potential for overexposure, and the ability to assess dose was not
compromised. Therefore, the finding was determined to be of very low safety
Enclosure
4
significance (Green). This finding involved the cross-cutting aspect of human
performance, work practices H.4.b] because the HRA event was a direct result of poor
communications during the pre-job briefing and a lack of procedure adherence on the
part of the maintenance worker. The licensee entered this issue into the Corrective
Action Program (CAP) as CR 523976. (Section 2RS1)
Violations of very low safety significance that were identified by the licensee have been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have been
entered into the licensees corrective action program. These violations and the corrective action
tracking numbers are listed in Section 4OA7 of this report.
Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 1 started the report period shutdown for a planned refueling outage. The licensee
conducted a reactor startup on October 8; however the licensee shutdown unit 1 later
that day due to issues with two main steam isolation valves (MSIVs). The licensee
restarted unit 1 on October 17 and attained full rated thermal power (RTP) power on
October 21. Unit 1 operated at essentially full RTP for the remainder of the inspection
period.
Unit 2 operated at essentially full RTP for the entire inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
.1 Seasonal Extreme Weather Conditions
a. Inspection Scope
The inspectors conducted a detailed review of the stations adverse weather procedures
written for extreme low temperatures. The inspectors verified that weather related
equipment deficiencies identified during the previous year had been corrected prior to
the onset of seasonal extremes. The inspectors evaluated the licensees
implementation of adverse weather preparation procedures and compensatory
measures before the onset of seasonal extreme weather conditions. Documents
reviewed are listed in the Attachment. The inspectors evaluated the following risk-
significant systems:
- Unit 1 emergency diesel generator (EDG) system (both trains)
- Unit 2 nuclear service cooling water (NSCW) system (both trains)
b. Findings
No findings were identified.
.2 Impending Adverse Weather Conditions
a. Inspection Scope
The inspectors reviewed the licensees preparations to protect risk-significant systems
from predicted severe weather conditions of sub-freezing temperatures expected on
December 21, 2012. The inspectors evaluated the licensees implementation of adverse
weather preparation procedures and compensatory measures, including operator
staffing, before the onset of and during the adverse weather conditions. The inspectors
Enclosure
6
reviewed the licensees plans to address the ramifications of potentially lasting effects
that may result from sub-freezing temperatures. The inspectors verified that operator
actions specified in the licensees adverse weather procedure maintained readiness of
essential systems. The inspectors verified that required surveillances were current, or
were scheduled and completed, if practical, before the onset of anticipated adverse
weather conditions. The inspectors also verified the licensee implemented periodic
equipment walk-downs or other measures to ensure that the condition of plant
equipment met operability requirements. Documents reviewed are listed in the
Attachment.
b. Findings
No findings were identified.
1R04 Equipment Alignment (71111.04)
.1 Partial Walkdown
The inspectors verified that critical portions of selected risk-significant systems were
correctly aligned. The inspectors selected systems for assessment because they were a
redundant or backup system/train, were important for mitigating risk for the current plant
conditions, had been recently realigned, or were a single-train system. The inspectors
determined the correct system lineup by reviewing plant procedures and drawings. The
inspectors verified that critical portions of the selected systems were correctly aligned by
performing partial walkdowns. Document reviewed are listed in the Attachment. The
inspectors selected the following three systems/trains to inspect:
- Diesel-driven fire water pump #2, the motor-driven firewater pump, and the
associated yard piping system while the diesel-driven firewater pump # 1 was out of
service for a scheduled maintenance outage
- Unit 1 train B containment spray (CS) system while the Unit 1 train A CS system was
out of service due to a planned maintenance outage.
- Unit 1 train B residual heat removal (RHR) system
b. Findings
No findings were identified.
.2 Complete Walkdown
a. Inspection Scope
The inspectors verified the alignment of the Unit 1 A train residual heat removal (RHR)
system. The inspectors selected this system for assessment because it is a risk-
significant mitigating system. The inspectors determined the correct system lineup by
reviewing plant procedures, drawings, the updated final safety analysis report, and other
documents. In order to identify any deficiencies that could affect the ability of the system
to perform its function(s), the inspectors reviewed records related to outstanding design
Enclosure
7
issues and maintenance work requests. The inspectors verified that the selected system
was correctly aligned by performing a complete walk down of accessible components.
To verify the licensee was identifying and resolving equipment alignment discrepancies,
the inspectors reviewed corrective action documents, including condition reports and
outstanding work orders, as well as periodic reports containing information on the status
of risk-significant systems, including maintenance rule reports and system health
reports. Document reviewed are listed in the Attachment.
b. Findings
No findings were identified.
1R05 Fire Protection (71111.05AQ)
a. Inspection Scope
Quarterly Inspection
The inspectors evaluated the adequacy of selected fire plans by comparing the fire plans
to the defined hazards and defense-in-depth features specified in the fire protection
program. In evaluating the fire plans, the inspectors assessed the following items:
(1) control of transient combustibles and ignition sources, (2) fire detection systems, (3)
water-based fire suppression systems, (4) gaseous fire suppression systems, (5) manual
firefighting equipment and capability (6) passive fire protection features, (7)
compensatory measures and fire watches, and (8) issues related to fire protection
contained in the licensees corrective action program. The inspectors toured the
following five fire areas to assess material condition and operational status of fire
protection equipment. Documents reviewed are listed in the Attachment.
- Unit 1 north and south main steam valve room, fire zones 99, 104 and 45
- Unit 2 control building level B east penetration areas (zones 62, 63 and 82) and west
penetration areas (zones 60, 61, and 64)
- Unit 1A and 1B nuclear service cooling water (NSCW) pump rooms, fire zones 160A
and 160B
- Unit 1 A, B, C, and D class 1E 125 VDC station batteries and associated switchgear
rooms, fire zones 71, 76, 77A, 77B, 78A, 78B, 79A, 79B, 56A, 56B, 83, and 152
- Unit 2 A, B, C, and D class 1E 125 VDC station batteries and associated switchgear
rooms, fire zones 71, 76, 77A, 77B, 78A, 78B, 79A, 79B, 56A, 56B, 83, and 152
b. Findings
No findings were identified.
Enclosure
8
1R06 Flood Protection Measures (71111.06)
The inspectors reviewed related flood analysis documents and walked down the areas
listed below that contain risk significant structures, systems, and components
susceptible to flooding. The inspectors verified plant design features and plant
procedures for flood mitigation were consistent with design requirements and internal
flooding analysis assumptions. The inspectors also assessed the condition of flood
protection barriers and drain systems. In addition, the inspectors verified the licensee
was identifying and properly addressing issues using their corrective action program.
Documents reviewed are listed in the Attachment.
- Unit 2 residual heat removal (RHR) and containment spray (CS) pump rooms
b. Findings
No findings were identified.
.2 Underground Cables
a. Inspection Scope
The inspectors reviewed related flood analysis documents and inspected the areas listed
below that contain cables whose failure could disable risk significant equipment. The
inspector directly observed the condition of cables and cable support structures and, as
applicable, verified that dewatering devices and drainage systems were functioning
properly. In addition, the inspectors verified the licensee was identifying and properly
addressing issues using their corrective action program. Documents reviewed are listed
in the Attachment.
- Cable Pull Boxes: 1NE7ADKEM40, 1NE7ADKEM39, 1NE9GHKEPB01
b. Findings
No findings were identified.
1R07 Heat Sink Performance (71111.07)
.1 Annual Review
a. Inspection Scope
The inspectors verified the readiness and availability of the Unit 1 emergency diesel
generator (EDG) jacket water (JW) heat exchanger to perform its design function by
observing performance tests or reviewing reports of those tests, verifying the licensee
uses the periodic maintenance method outlined in Generic Letter 89-13, Service Water
Enclosure
9
System Problems Affecting Safety Related Equipment, observing the licensees heat
exchanger inspections and verifying critical operating parameters through direct
observation or by reviewing operating data. Additionally, the inspectors verified that the
licensee had entered any significant heat exchanger performance problems into their
corrective action program and that the licensees corrective actions were appropriate.
Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program (71111.11)
.1 Resident Inspector Quarterly Review of Licensed Operator Requalification
a. Inspection Scope
The inspectors observed an evaluated simulator scenario administered to an operating
crew conducted in accordance with the licensees accredited requalification training
program. The inspectors assessed licensed operator performance, the ability of the
licensee to administer the scenario and evaluate the operators, the quality of any post-
scenario critique, any follow-up actions taken by the facility licensee, and the
performance of the simulator. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2 Resident Inspector Quarterly Review (Licensed Operator Performance):
The inspectors observed licensed operator performance in the main control room on
October 8, during the reactor startup following the 1R17 refueling outage. Documents
reviewed are listed in the Attachment. Inspectors observed licensed operator
performance to assess the following:
- Use of plant procedures
- Control board manipulations
- Communications between crew members
- Use and interpretation of instruments, indications, and alarms
- Use of human error prevention techniques
- Documentation of activities
- Management and supervision
b. Findings
No findings were identified.
Enclosure
10
.3 Annual Review of Licensee Requalification Examination Results (71111.11A)
a. Inspection Scope
On November 29, 2012, the licensee completed administering the annual requalification
operating test and biennial written examination, which is required to be administered to
all licensed operators in accordance with 10 CFR 55.59(a)(2). The inspectors performed
an in-office review of the overall pass/fail results of the individual operating tests and the
crew simulator operating tests. These results were compared to the thresholds
established in Manual Chapter 609, Appendix I, Operator Requalification Human
Performance Significance Determination Process.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors assessed the licensees treatment of the issues listed below in order to
verify the licensee appropriately addressed equipment problems within the scope of the
Maintenance Rule (10 CFR 50.65). The inspectors reviewed procedures and records in
order to evaluate the licensees identification, assessment, and characterization of the
problems as well as their corrective actions for returning the equipment to a satisfactory
condition. Documents reviewed are listed in the Attachment. The inspectors also
interviewed system engineers and the maintenance rule coordinator to assess the
accuracy of performance deficiencies and extent of condition.
- Unit 2, 2PV3000 main steam atmospheric relief valve (ARV) exceeded MR
performance criteria for maintenance preventable functional failures (MPFFs)
- Unit 1, 1HV8702A B train RHR pump hot leg suction valve failed to open when
demanded from the control room.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
The inspectors reviewed the maintenance activities listed below to verify the licensee
assessed and managed plant risk as required by 10 CFR 50.65(a)(4) and licensee
procedures. The inspectors assessed the adequacy of the licensees risk assessments
and implementation of risk management actions. The inspectors also verified that the
licensee was identifying and resolving problems with assessing and managing
maintenance-related risk using the corrective action program. Additionally, for
maintenance resulting from unforeseen situations, the inspectors assessed the
Enclosure
11
effectiveness of the licensees planning and control of emergent work activities.
Documents reviewed are listed in the Attachment.
- Week of 10/22: Unit 1A EDG 18 month surveillance in conjunction with ongoing
work in the high voltage switchyard
- Week of 10/29: maintenance outage on the 2A EDG concurrent with maintenance
activities on various breakers in the high-voltage switchyard.
C Week of 11/12: performing testing on the 2B EDG concurrent with maintenance
activities in the high-voltage switchyard.
C Week of 12/3: rendering the Unit 1 B train NSCW cooling tower fans inoperable to
perform inspections of the 480V EMAX breakers.
C Week of 12/17: rendering the Unit 1 A train NSCW cooling tower fans inoperable to
perform inspections of the 480V EMAX breakers.
b. Findings
No findings were identified.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed the following five evaluations to verify they met the
requirements of procedure NMP-GM-002, Corrective Action Program, and NMP-GM-
002-001, Corrective Action Program Instructions. The scope of these inspections
included a review of the technical adequacy of the evaluations, the adequacy of
compensatory measures, and the impact on continued plant operation. Inspectors
reviewed licensee procedures and conducted walkdowns in accordance with OpESS FY
2012-02 Technical Specification Interpretation and Operability Determination,
inspection guidance to ensure that the licensee properly interprets Technical
Specification (TS) Bases, properly develops operability determinations, and adheres to
the NRCs guidance on the temporary use of manual actions to support operability
determinations. Documents reviewed are listed in the Attachment.
C CR 501741, Unit 1A EDG unit available light is not working
C CR 533822, Unit 1 and 2 main steam isolation valves (MSIVs)
C CR 550736, Unit 2A NSCW fan #3 failed to close
C CR 479829, Unit 2B RHR oil sample analysis
C CR 562504, Gas void discovered at 1-1208-X4-05
b. Findings
No findings were identified.
Enclosure
12
1R18 Plant Modifications
a. Inspection Scope
Temporary Modification: The inspectors reviewed minor design change package
SNC439341/1.0 which allowed the temporary installation of clamping device on valve 1-
1208-U4-A11, auxiliary pressurizer spray valve, to address body-to-bonnet leakage of
the non-isolable valve. Inspectors also reviewed the associated 10 CFR Part 50.59
screening criteria against the system design bases documentation and procedure
00307-C, Temporary Modifications. The inspectors reviewed implementation,
configuration control, post-installation test activities, drawing and procedure updates,
and operator awareness for this temporary modification.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors either observed post-maintenance testing or reviewed the test results for
the following six maintenance activities to verify that the testing met the requirements of
procedure 29401-C, Work Order Functional Tests, for ensuring equipment operability
and functional capability was restored. The inspectors also reviewed the test
procedures to verify the acceptance criteria were sufficient to meet the Technical
Specification (TS) operability requirements.
C MWO SNC410412, Unit 1 Train B Safety Injection Accumulator Vent Header
Isolation Valve 1HV-0943B
C MWO SNC439646, 1HV-3016B Disassemble MSIV and MWO SNC439647, 1HV-
3026B Disassemble MSIV
C MWO SNC43990, 1HV-3006A Disconnect Actuator and Perform Ultra-Sonic Testing
C MWO SNC424222- 2A EDG Replace O-Rings in 2R, 2L, 7R, and 7L Fuel Injectors
C MWO SNC127251 - Clean, Inspect, and Lubricate Valve Stem
C MWO SNC447741 - 2A NSCW Fan #3 Not Operating
b. Findings
Introduction: A Green self-revealing non-cited violation (NCV) of 10 CFR 50 Appendix B
Criterion V, Instructions, Procedures, and Drawings was identified for failure to provide
adequate work instructions in the maintenance procedure used to modify 480V EMAX
safety-related switchgear breakers. Specifically, when modifying the breaker by adding
a higher amperage closing coil, failure to verify the proper placement of the wire bundle
on top of the closing coil following replacement resulted in the safety-related 480V
EMAX breaker failing to close when demanded. Investigation revealed that the wire
bundle on top of the closing coil was positioned such that it prevented the close lever
from fully seating, thus preventing the breaker from closing.
Enclosure
13
Description: At 0700 on November 19, 2012, the control room operators noted that the
2A NSCW cooling tower fan #3 did not start when the associated tower return valve
opened. On Unit 2, NSCW cooling tower fan #3 is interlocked to automatically start
when the associated tower return valve opens and initiates flow to the tower spray lines.
At the direction of the shift supervisor, the operators attempted to start fan #3 manually
from the main control room, but the fan did not start. The green light for the hand switch
indicated that control power was available. The licensee declared the fan inoperable
and an incident response team (IRT) was assembled. The 480V EMAX breaker was
replaced and the monthly surveillance test was successfully completed. The fan was
declared operable at 4:30 a.m. on November 20, 2012. The investigation of the failed
breaker revealed that the wire bundle on top of the closing coil was in physical contact
with the side of the close lever, preventing the close lever from fully seating, thus
preventing the breaker from closing. When the wire bundle was properly anchored away
from the close lever, the breaker operated normally. The IRT determined that the wire
bundle on this breaker had been improperly repositioned following modification of the
breaker. The 480V EMAX breakers that supply specific safety-related loads at Vogtle
must be modified by installing a higher amperage closing coil because with the current
plant design, the closing coils are continuously energized. Maintenance procedure
28480-C Rev.26, 480V EMAX Breaker Maintenance, did not provide any guidance on
how to anchor the wire bundle following closing coil replacement so that the wire bundle
did not interfere with the operation of the close lever. The licensee entered this issue
into their corrective action program as condition reports (CR) 549999 and CR 550736.
Analysis: The failure of the breaker to close when demanded and start the safety-
related NSCW cooling tower fan was a performance deficiency. The inspectors
concluded that the finding was more than minor because it impacted the reactor safety
mitigating systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable
consequences and affected the cornerstone attribute of equipment performance.
Specifically, failure of the 480V EMAX breaker to close when demanded resulted in the
2A NSCW cooling tower fan #3 being inoperable for approximately 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Using IMC 0609, Attachment 4, Initial Characterization of Findings dated
June 19, 2012, the inspectors determined that finding affected the mitigating systems
cornerstone. The inspectors evaluated the finding using IMC 0609, Appendix A, The
Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012.
The inspectors used the Initial Screening and Characterization of Findings (IMC 0609.04
Exhibit 2, dated June 19, 2012) to characterize the finding. Since the inspectors
answered No to all of the Exhibit 2, section A, questions 1-4, mitigating systems
cornerstone screening questions, the inspectors concluded that the finding was of very
low safety significance (Green).
The inspectors determined the primary cause of the performance deficiency was that the
maintenance technicians who were modifying the 480V EMAX breakers were unaware
that the improper placement of the closing coil wire bundle could interfere with the
operation of the close lever. The inspectors determined that the cause of this finding
was related to the corrective action program component of the problem identification and
resolution cross-cutting area due to less-than-adequate problem evaluation P.1(c).]
Enclosure
14
Specifically, the licensees extent of cause evaluations performed on previous 480V
EMAX breaker failures (caused by restricted movement of the close lever) did not
identify the potential of the closing coil wire bundle to interfere with the proper movement
of the close lever.
Enforcement: The inspectors determined that the finding represents a violation of
regulatory requirements because it involved inadequate maintenance procedures used
to modify safety-related plant equipment. 10 CFR Part 50 Appendix B Criterion V,
Instructions, Procedures, and Drawings, requires, in part, that procedures shall include
appropriate quantitative or qualitative acceptance criteria for determining that important
activities have been satisfactorily accomplished. Contrary to the above, maintenance
procedure 28480-C Rev.26, 480V EMAX Breaker Maintenance, did not provide
guidance on how to anchor the wire bundle following closing coil replacement so that the
wire bundle did not interfere with the operation of the close lever. As a result of the
violation, the 2A NSCW cooling tower fan #3 was rendered inoperable for a period of
approximately 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on November 19 and 20. Because this violation was of very
low safety significance and was entered into the licensees corrective action program as
CR 549999 and CR 550736, this violation is being treated as a NCV, consistent with
Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000425/2012005-01, Inadequate
Maintenance Procedure Results in Inoperability of NSCW Cooling Tower Fan.)
1R20 Refueling and Other Outage Activities
a. Inspection Scope
The inspectors performed the activities described below for the 1R17 refueling outage
that ended on October 18, 2012. The inspectors confirmed that, when the licensee
removed equipment from service, the licensee maintained defense-in-depth
commensurate with the outage risk control plan for key safety functions and applicable
TS and that configuration changes due to emergent work and unexpected conditions
were controlled in accordance with the outage risk control plan. Inspection activities
included:
C Observed refueling activities for compliance with TS, to verify proper tracking of fuel
assemblies from the spent fuel pool to the core, and to verify foreign material
exclusion was maintained
C Performed containment closure activities, including a detailed containment walkdown
prior to startup, to verify no evidence of leakage and that debris had not been left
which could affect the performance of the containment sump
C Observed heat up and startup activities to verify that TS, license conditions, and
other requirements, commitments, and administrative procedure prerequisites for
mode changes were met prior to changing modes or plant conditions. Reactor
coolant system (RCS) integrity was verified by reviewing RCS leakage calculations
and containment integrity was verified by reviewing the status of containment
penetrations and containment isolation valves
Enclosure
15
b. Findings
No findings were identified.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors reviewed the following eight surveillance test procedures and either
observed the testing or reviewed test results to verify that testing was conducted in
accordance with the procedures and that the acceptance criteria adequately
demonstrated that the equipment was operable. Additionally, the inspectors reviewed
the CR database to verify that the licensee had adequately identified and implemented
appropriate corrective actions for surveillance test problems.
Surveillance Tests
C 14666-1 Rev. 34.1, Train A Diesel Generator and ESFAS Test
C 14670A-1 Rev. 1.2, Diesel Generator 1A Hot Restart Test
C 14802A-1 Rev. 4.1, Train A NSCW Pump/Check Valve IST and Response Time Test
C 14804A-1 Rev. 4, Safety Injection Pump A Inservice and Response Time Test
Containment Isolation Valve
C 14372B-1 Rev. 9, Containment Penetration No. 72B RCS loop 3 Accumulator
Sample Line Local Leak Rate Test
In-Service Tests (IST)
C 14808A-2 Rev. 2.2, Train A Centrifugal Charging Pump and Check Valve IST and
Response Time Test
RCS Leak Detection
C 14905-1 Rev. 67.2, RCS Leakage Calculation (Inventory Balance)
C 14905-2 Rev. 50.1, RCS Leakage Calculation (Inventory Balance)
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes
a. Inspection Scope
The NSIR headquarters staff performed an in-office review of the latest revisions of
various Emergency Plan Implementing Procedures (EPIPs) and the Emergency Plan
located under ADAMS accession numbers ML12188A352, ML122430598,
ML12081A211, ML12093A204, and ML12115A174.
Enclosure
16
The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in
the revisions resulted in no reduction in the effectiveness of the Plan, and that the
revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to
10 CFR Part 50. The NRC review was not documented in a safety evaluation report and
did not constitute approval of licensee-generated changes; therefore, these revisions are
subject to future inspection. This inspection activity satisfied one inspection sample for
the emergency action level and emergency plan changes on an annual basis.
Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
2. RADIATION SAFETY [RS]
2RS1 Radiological Hazard Assessment and Exposure Controls
a. Inspection Scope
Hazard Assessment and Instructions to Workers: During facility tours, the inspectors
directly observed labeling of radioactive material and postings for radiation areas, high
radiation areas (HRA)s, Very High Radiation Areas (VHRA)s and airborne radioactivity
areas established within the radiologically controlled area (RCA) of the auxiliary building,
Unit 1 (U1) reactor containment building, and radioactive waste (radwaste) processing
and storage locations. The inspectors independently measured radiation dose rates or
directly observed conduct of licensee radiation surveys for selected RCA areas. The
inspectors reviewed survey records for several plant areas including surveys for alpha
emitters, discrete radioactive particles, airborne radioactivity, gamma surveys with a
range of dose rate gradients, and pre-job surveys for upcoming tasks. The inspectors
also discussed changes to plant operations that could contribute to changing radiological
conditions since the last inspection. For selected outage jobs, the inspectors attended
pre-job briefings and reviewed radiation work permit (RWP) details to assess
communication of radiological control requirements and current radiological conditions to
workers.
Hazard Control and Work Practices: The inspectors evaluated access barrier
effectiveness for selected Locked High Radiation Area (LHRA) locations and discussed
changes to procedural guidance for LHRA and VHRA controls with health physics (HP)
supervisors. The inspectors reviewed implementation of controls for the storage of
irradiated material within the spent fuel pool (SFP). Established radiological controls
(including airborne controls) were evaluated for selected Unit 1 Refueling Outage 17
(1R-17) tasks including maintenance activities in the transfer canal, removal of
pressurizer code safety valves, and steam generator (S/G) eddy current testing. In
addition, the inspectors reviewed licensee controls for areas where dose rates could
change significantly as a result of plant shutdown and refueling operations.
Enclosure
17
Through direct observations and interviews with licensee staff, inspectors evaluated
occupational workers adherence to selected RWPs and HP technician (HPT) proficiency
in providing job coverage. Electronic dosimeter (ED) alarm set points and worker stay
times were evaluated against area radiation survey results for selected 1R17 job tasks.
As part of Inspection Procedure (IP) 71124.04, inspectors reviewed the use of personnel
dosimetry (ED alarms, extremity dosimetry, multibadging in high dose rate gradients,
etc.). The inspectors also evaluated worker responses to dose and dose rate alarms
during selected work activities.
Control of Radioactive Material: The inspectors observed surveys of material and
personnel being released from the RCA using small article monitor (SAM), personnel
contamination monitor (PCM), and portal monitor (PM) instruments. The inspectors
reviewed the last two calibration records for selected release point survey instruments
and discussed equipment sensitivity, alarm setpoints, and release program guidance
with licensee staff. The inspectors compared recent 10 Code of Federal Regulations
(CFR) Part 61 results for the Dry Active Waste (DAW) radioactive waste stream with
radio nuclides used in calibration sources to evaluate the appropriateness and accuracy
of release survey instrumentation. The inspectors also reviewed records of leak tests on
selected sealed sources and discussed nationally tracked source transactions with
licensee staff.
Problem Identification and Resolution: The inspectors reviewed and assessed Condition
Reports (CR)s associated with radiological hazard assessment and control. The
inspectors evaluated the licensees ability to identify and resolve the issues in
accordance with procedure NMP-GM-002, Corrective Action Program, Ver. 12.1. The
inspectors also reviewed recent self-assessment results.
Radiation protection activities were evaluated against the requirements of Final Safety
Analysis Report (FSAR) Section 12; Technical Specifications (TS) Sections 5.4 and 5.7;
10 CFR Parts 19 and 20; and approved licensee procedures. Licensee programs for
monitoring materials and personnel released from the RCA were evaluated against 10
CFR Part 20 and IE Circular 81-07, Control of Radioactively Contaminated Material.
Documents reviewed are listed in Sections 2RS1, 2RS2, 2RS3, and 2RS5 of the
Attachment.
b. Findings
Introduction: The inspectors identified a Green, self-revealing, Non-cited Violation
(NCV) of TS 5.7.1, High Radiation Area, for an unauthorized entry into a HRA.
Description: On September 26, 2012, a maintenance worker entered U1 containment to
locate and walk down two Nuclear Service Cooling Water check valves in preparation for
inspection. Prior to the entry, the worker received a pre-job briefing from HP to allow
access into containment. Neither the worker nor the HPT performing the briefing were
confident of the valve locations, however they determined that Quadrant 1 of the 184
elevation was the most likely area to find the valves. Since the assigned RWP (as well
as TS 5.7.1) did not allow entry into any HRAs without first obtaining a briefing on the
radiological conditions, the worker was briefed on conditions in HRAs on the 184
Enclosure
18
elevation. The worker was not briefed on conditions in HRAs in other parts of
containment. During the entry, the worker was unable to locate one of the check valves
on the 184 elevation and proceeded to exit out a stairwell to the 197 elevation. On the
197 elevation, the worker observed the other check valve inside a posted HRA. Without
knowledge of dose rates in the area, the worker proceeded past the HRA swing gate to
verify the valve tag number and subsequently received an ED dose rate alarm of 173
mrem/hr. The worker acknowledged the ED alarm and backed out of the HRA. In
addition, HP central control noted the ED alarm via telemetry and sent an HPT to
intercept the worker. The workers ED alarm setpoint was 150 mrem/hr and dose rates
in the area were as high as 445 mrem/hr on contact and 117 mrem/hr at 30 cm. The
licensee entered this issue into the Corrective Action Program (CAP) as CR 523976 and
took immediate corrective actions including an outage work crew stand down.
Analysis: The inspectors determined that the unauthorized entry into a HRA was a
performance deficiency. This finding is greater than minor because it is associated with
the Occupational Radiation Safety Cornerstone attribute of Human Performance and
adversely affects the cornerstone objective of ensuring adequate protection of worker
health and safety from exposure to radiation from radioactive material during routine
civilian nuclear reactor operation. Workers who enter HRAs without prior knowledge of
current radiological conditions could receive unintended occupational exposures. The
finding was evaluated using the Occupational Radiation Safety Significance
Determination Process (SDP). The finding was not related to As Low As Reasonably
Achievable (ALARA) planning, nor did it involve an overexposure or substantial potential
for overexposure, and the ability to assess dose was not compromised. Therefore, the
inspectors determined the finding to be of very low safety significance (Green). The
inspectors noted that the maintenance worker responded properly to the ED dose rate
alarm and was monitored by HP central control via telemetry, thereby limiting his
potential for unintended exposure. This finding involved the cross-cutting aspect of
Human Performance, Work Practices H.4.b] because the HRA event was a direct result
of poor communications during the pre-job briefing and a lack of procedure (RWP)
adherence on the part of the maintenance worker.
Enforcement: TS 5.7.1, High Radiation Area, requires individuals entering HRAs to
meet one or more of the following criteria: 1) be provided with a survey meter; 2) wear
an ED and be made aware of radiological conditions in the area; or 3) be escorted by a
HPT. Contrary to the above, on September 26, 2012, a worker entered a HRA without a
survey meter, without being made aware of radiological conditions in the area, and
without HPT escort. Because this violation was of very low safety significance and it
was entered into the licensees corrective action program (CR 523976), this violation is
being treated as an NCV, consistent with the Enforcement Policy: NCV 05000424/425,
2012005-02, Unauthorized Entry into a High Radiation Area.
Enclosure
19
2RS2 Occupational ALARA Planning and Controls
a. Inspection Scope
Work Planning and Exposure Tracking: The inspectors reviewed planned work activities
and their collective exposure estimates for the current 1R17 outage. The inspectors
reviewed ALARA planning packages for the following high collective exposure tasks:
S/G 1-4 job evolutions (nozzle dam removal and installation, eddy current testing, and
sludge lance) and repair of unit 1 transfer canal transfer tube isolation valve work. For
the selected tasks, the inspectors reviewed established dose goals and discussed
assumptions regarding the bases for the current estimates with responsible ALARA
planners. The inspectors evaluated the incorporation of exposure reduction initiatives
and operating experience, including historical post-job reviews, into RWP requirements.
Day-to-day collective dose data for the selected tasks were compared with established
dose estimates and evaluated against procedural criteria (work-in-progress review limits)
for additional ALARA review. Where applicable, the inspectors discussed changes to
established estimates with ALARA planners and evaluated them against work scope
changes or unanticipated elevated dose rates.
Source Term Reduction and Control: The inspectors reviewed the collective exposure
three-year rolling average from 2009 - 2011 and reviewed historical collective exposure
trends from 1988 - 2012. The inspectors evaluated historical dose rate trends for
reactor coolant system piping and compared them to current 1R17 data. Crud burst
evolution during the first week of the 1R17 outage and source term reduction initiatives
were reviewed and discussed with Chemistry and HP staff.
Radiation Worker Performance: The inspectors observed radiation worker performance
for S/G job evolutions such as the diaphragm and manway removal and nozzle dam
removal and installation. The inspectors observed ALARA briefings for multiple steam
generator jobs and emerging jobs such as U1 transfer canal valve work. Radiation
worker performance was also evaluated as part of IP 71124.01. While observing job
tasks, the inspectors evaluated the use of remote technologies to reduce dose including
teledosimetry and remote visual monitoring.
Problem Identification and Resolution: The inspectors reviewed and discussed selected
CAP documents associated with ALARA program implementation. The inspectors
evaluated the licensees ability to identify and resolve the issues in accordance with
licensee procedure NMP-GM-002, Corrective Action Program, Ver. 12.1. The
inspectors also evaluated the scope and frequency of the licensees self-assessment
program and reviewed recent assessment results.
ALARA program activities were evaluated against the requirements of FSAR Section 12,
TS Section 5.4, 10 CFR Part 20, and approved licensee procedures. Records reviewed
are listed in the Attachment.
Enclosure
20
b. Findings
No findings were identified.
2RS3 In-Plant Airborne Radioactivity Control and Mitigation
a. Inspection Scope
Engineering Controls: The inspectors reviewed the use of temporary and permanent
engineering controls to mitigate airborne radioactivity during the 1R17 refueling outage.
The inspectors observed the use of portable air filtration units for work in contaminated
areas of the containment building and reviewed filtration unit testing certificates. The
inspectors evaluated the effectiveness of continuous air monitors and air samplers placed
in work area breathing zones to provide indication of increasing airborne levels.
Respiratory Protection Equipment: The inspectors reviewed the use of respiratory
protection devices to limit the intake of radioactive material. This included review of
devices used for routine tasks and devices stored for use in emergency situations. As part
of IP 71124.02, the inspectors reviewed ALARA evaluations for the use of respiratory
protection devices during work in the lower cavity near the transfer canal. Selected Self-
Contained Breathing Apparatus (SCBA) units and negative pressure respirators (NPR)s
staged for routine and emergency use in the Main Control Room and other locations were
inspected for material condition, SCBA bottle air pressure, number of units, and number of
spare masks and air bottles available. The inspectors reviewed maintenance records for
selected SCBA units for the past two years and evaluated SCBA and NPR compliance with
National Institute for Occupational Safety and Health certification requirements. The
inspectors also reviewed records of air quality testing for supplied-air devices and SCBA
bottles.
The inspectors observed the use of air-supplied suits during S/G maintenance work. The
inspectors discussed training for various types of respiratory protection devices with HP
staff and interviewed radworkers and control room operators on use of the devices
including SCBA bottle change-out and use of corrective lens inserts. The inspectors
reviewed respirator qualification records (including medical qualifications) for several Main
Control Room operators and emergency responder personnel in the Maintenance and HP
departments. In addition, inspectors evaluated qualifications for individuals responsible for
testing and repairing SCBA vital components.
Problem Identification and Resolution: The inspectors reviewed and assessed CRs
associated with airborne radioactivity mitigation and respiratory protection. The
inspectors evaluated the licensees ability to identify and resolve the issues in
accordance with procedure NMP-GM-002, Corrective Action Program, Ver. 12.1. The
inspectors also reviewed recent self-assessment results.
Licensee activities associated with the use of engineering controls and respiratory
protection equipment were reviewed against TS Section 5.4; 10 CFR Part 20; Regulatory
Guide 8.15, Acceptable Programs for Respiratory Protection; and applicable licensee
procedures. Documents reviewed are listed in the Attachment.
Enclosure
21
b. Findings
No findings were identified.
2RS4 Occupational Dose Assessment
a. Inspection Scope
External Dosimetry: The inspectors reviewed the licensees National Voluntary
Accreditation Program (NVLAP) certification data for accreditation for the current year for
Ionizing Radiation Dosimetry. The inspectors reviewed program procedures for
processing active personnel dosimeters (ED)s and onsite storage of Optically Stimulated
Luminescent Dosimeters (OSLD)s. Comparisons between ED and OSLD results,
including correction factors, were discussed in detail. The inspectors also reviewed
dosimetry occurrence reports regarding alarming dosimeters.
Internal Dosimetry: Inspectors reviewed and discussed the in vivo bioassay program
with the licensee. Inspectors reviewed procedures that addressed methods for
determining internal or external contamination, releasing contaminated individuals, the
assignment of dose, and the frequency of measurements depending on the nuclides.
Inspectors reviewed and evaluated Whole Body Counter (WBC) records selected from
March 2010 to September 2012. The inspectors evaluated the licensees program for in
vitro monitoring, however there were no dose assessments using this method to review.
There were no internal dose assessments for internal exposure greater than 10 millirem
committed effective dose equivalent to review.
Special Dosimetric Situations: The inspectors reviewed records for declared pregnant
workers (DPW)s from March 2010 through September 2012 and discussed guidance for
monitoring and instructing DPWs. Inspectors reviewed and witnessed the licensees
practices for monitoring external dose in areas of expected dose rate gradients,
including the use of multi-badging and extremity dosimetry. The inspectors evaluated
the licensees neutron dosimetry program including instrumentation which was evaluated
under procedure 71124.05. In addition, the inspectors evaluated the adequacy of
shallow dose assessments for selected Personnel Contamination Events occurring
between September 2011 and September 2012.
Problem Identification and Resolution: The inspectors reviewed and discussed licensee
CAP documents associated with occupational dose assessment. Inspectors evaluated
the licensees ability to identify and resolve the identified issues in accordance with
procedure NPM-GM-002, Corrective Action Program, Version 12.1. The inspectors
also discussed the scope of the licensees internal audit program and reviewed recent
assessment results.
HP program occupational dose assessment activities were evaluated against the
requirements of FSAR Section 12; TS Section 5.4; 10 CFR Parts 19 and 20; and
approved licensee procedures. Records reviewed are listed in the Attachment.
Enclosure
22
b. Findings
No findings were identified.
2RS5 Radiation Monitoring Instrumentation
a. Inspection Scope
Radiation Monitoring Instrumentation: During walk-downs of the auxiliary building and
the RCA exit point, the inspectors observed installed and portable radiation detection
equipment. These included area radiation monitors (ARM)s, continuous air samplers,
liquid and gaseous effluent monitors, PCMs, SAMs, PMs, a WBC, count room
equipment, and portable survey instruments. The inspectors observed the physical
location of the components, noted their material condition, observed the currency of
calibration and source check stickers, and discussed performance of equipment with RP
personnel.
In addition to equipment walk-downs, the inspectors observed source functional checks
of portable detection instruments, including ion chambers and telepoles. For the
portable instruments, the inspectors observed the use of a high-range calibrator and
discussed periodic output value testing, calibration, and source check processes with
health physics technicians. The inspectors reviewed calibration records and discussed
with chemistry personnel alarm setpoint values for PCMs, PMs, effluent monitors,
WBCs, and an ARM. This included a sampling of instruments used for post-accident
monitoring such as a containment high-range radiation monitor and effluent monitors for
noble gas and iodine. The inspectors reviewed the most recent 10 CFR Part 61 analysis
for DAW to determine if calibration and check sources are representative of the plant
source term. The inspectors observed computerized performance check calibration
efficiency information for count room gamma detectors and a liquid scintillation detector.
The inspectors also observed the currency of calibration for selected EDs at the RCA
entry point.
Problem Identification and Resolution: The inspectors reviewed selected CAP reports in
the area of radiological instrumentation. The inspectors evaluated the licensees ability
to identify and resolve the issues in accordance with procedure NMP-GM-002-001,
Corrective Action Program Instructions, Ver. 29.0. Documents reviewed are listed in
the Attachment.
Effectiveness and reliability of selected radiation detection instruments were reviewed
against details documented in the following: 10 CFR Part 20; NUREG-0737,
Clarification of TMI Action Plan Requirements; FSAR Chapters 11 and 12; and
applicable licensee procedures. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
Enclosure
23
4. OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification
a. Inspection Scope
The inspectors sampled licensee submittals for the listed mitigating systems cornerstone
PIs during the period from October 1, 2011, through September 30, 2012, for Unit 1 and
Unit 2. The inspectors verified the licensees basis in reporting each data element using
the PI definitions and guidance contained in procedures 00163-C, NRC Performance
Indicator and Monthly Operating Report Preparation and Submittal, and Nuclear Energy
Institute 99-02, Regulatory Assessment Indicator Guideline.
Mitigating Systems Cornerstone
C Emergency AC Power Systems
C Cooling Water Systems
C Safety System Functional Failures
The inspectors reviewed Unit 1 and Unit 2 operator log entries, the CR data base, the
Vogtle mitigating system performance indicator basis document, the monthly operating
reports and monthly PI summary reports to verify that the licensee had accurately
submitted the PI data.
Occupational Radiation Safety Cornerstone: The inspectors reviewed the Occupational
Exposure Control Effectiveness PI results for the Occupational Radiation Safety
Cornerstone from June 2011 through September 2012. For the assessment period, the
inspectors reviewed electronic dosimeter alarm logs and CRs related to controls for
exposure significant areas. Documents reviewed are listed in the Attachment.
Public Radiation Safety Cornerstone: The inspectors reviewed the Radiological Control
Effluent Release Occurrences PI results for the Public Radiation Safety Cornerstone
from June 2011 through September 2012. For the assessment period, the inspectors
reviewed cumulative and projected doses to the public contained in liquid and gaseous
release permits and CRs related to Radiological Effluent Technical Specifications/Offsite
Dose Calculation Manual issues. The inspectors also reviewed licensee procedural
guidance for collecting and documenting PI data. Documents reviewed are listed in the
Attachment.
The inspectors completed five of the required samples specified in IP 71151.
b. Findings
No findings were identified.
Enclosure
24
4OA2 Identification and Resolution of Problems
.1 Daily Condition Report Review. As required by inspection procedure 71152,
Identification and Resolution of Problems, and in order to help identify repetitive
equipment failures or specific human performance issues for follow-up, the inspectors
performed a daily screening of items entered into the licensees corrective action
program. This review was accomplished by either attending daily screening meetings
that briefly discussed major CRs, or accessing the licensees computerized corrective
action database and reviewing each CR that was initiated.
.2 Focused Review
a. Inspection Scope
The inspectors performed a detailed review of the following CR which addressed the
failure of the Unit 1 Loop 2 & 3 outboard MSIVs in the closed position during startup.
The goal of the review was to verify that the full extent of the issue was identified, an
appropriate evaluation was performed, and appropriate corrective actions were specified
and prioritized. The inspectors evaluated the CR against the licensees corrective action
program as delineated in licensee procedure NMP-GM-002, Corrective Action Program,
and 10 CFR 50, Appendix B, Criterion XVI, Corrective Action. Documents reviewed are
listed in the Attachment.
- 530916 - Unit 1 steam generators 2&3 do not indicate steam flow
b. Findings and Observations
Introduction: A Green self-revealing non-cited violation (NCV) of 10 CFR 50 Appendix B
Criterion V, Instructions, Procedures, and Drawings was identified for failure to provide
adequate work instructions in the operations and maintenance procedures used to open
main steam isolation valves (MSIVs) that were bound in their closed seat. Specifically,
the operations and maintenance procedures used to open the loop 2 and loop 3
outboard MSIVs did not provide instructions to limit the magnitude of the force applied to
the valve stems while attempting to open the valves. Investigation revealed that the
cause of the stem failures was excessive force applied to the thermally embrittled stems.
Description: On October 8, with Unit 1 in Mode 2, the operators had begun preparations
for power ascension. At 1616, as the main feed pump was being placed on line, the
control room operators noted a divergence in RCS loop differential temperatures (Ts),
steam pressures, and steam flows between loops 1 & 4 and loops 2 & 3. Loops 1 & 4
showed increasing loop Ts, lowering steam pressure, and some minimal steam flow,
while loops 2 & 3 showed no loop T, increasing steam pressures (to the point of lifting
the loop 2 & 3 atmospheric relief valves), and no steam flow. The Main Control Board
hand switches indicated that all MSIVs and associated bypass valves were open. The
operators identified the potential impact to the core neutron flux and stopped power
ascension. Following discussions with plant management and engineering, the
operators placed the plant in a safe condition by inserting a manual trip of the reactor at
Enclosure
25
2155. The licensee subsequently assembled an Issue Response Team (IRT) and a root
cause team to investigate the cause of the diverging indications and to determine the
required corrective actions.
The investigations revealed that the outboard MSIVs on both loops 2 & 3 were failed in
the closed position. Upon disassembly, it was discovered the stems of both the failed
MSIVs had undergone brittle fracture just above the T-head, where the valve stem is
connected to the valve disk. Westinghouse representatives were consulted on the MSIV
issue. They conveyed to the licensee that the material used for the MSIV stems, ASME
SA564 Gr. 630PH T 17-4 PH heat treated to 1100oF, is susceptible to embrittlement
when exposed to temperatures above 500oF for a sustained period (after about 10
years). Metallurgical analysis performed on the sheared stems validated that thermal
embrittlement was the failure mechanism. The failure analysis concluded that both stem
fractures were the result of sudden brittle failures from single tensile stress events.
Further investigation by the IRT revealed that the loop 2 outboard MSIV stem failed
during main steam line warming evolutions conducted on October 6 by operations
personnel. The IRT also determined that that the loop 3 outboard MSIV stem failed on
the night of October 7 following activities performed by maintenance personnel to lift the
valve disk off its closed seat.
The root cause team determined that the root cause of the MSIV stem failures was
temperature aging embrittlement of the stem material. The team also determined that
the major contributing causes of the event were thermal binding of the valve disks in the
closed seat and inadequate procedural guidance, i.e. procedures used to open the
MSIVs did not provide instructions or guidance to limit the magnitude of the force applied
to the valve stems while attempting to open the valves, which ultimately resulted in the
brittle failure of the valve stems. The inadequate procedures specified by the root cause
team were operating procedures 12001-C, Unit Heat Up to Hot Shutdown (Mode 5 to
Mode 4), and 14850-1/2, Cold Shutdown Valve In-Service Test, and maintenance
procedure 26854-C, MSIV Actuator Maintenance. The licensee conducted ultrasonic
testing on the remaining six Unit 1 MSIVs to verify that the valve stems were intact. The
two failed valve stems were replaced, and the reactor was restarted on October 17. The
licensee entered this issue into their corrective action program as CR 530916.
Analysis: The failure to provide adequate work instructions in the operations and
maintenance procedures used to open main steam isolation valves (MSIVs) that were
stuck on their closed seat was a performance deficiency. The inspectors concluded that
the finding was more than minor because it was associated with the procedure quality
attribute of the reactor safety - initiating events cornerstone and it adversely affected the
cornerstone objective to limit the likelihood of events that upset plant stability and
challenge critical safety functions during shutdown as well as power operations.
Specifically, the failure to provide adequate work instructions to operations and
maintenance personnel resulted in the failure of both the loop 2 and loop 3 outboard
MSIVs and the subsequent manual reactor trip.
Using IMC 0609, Attachment 4, Initial Characterization of Findings dated June 19,
2012, the inspectors determined that finding affected the Initiating Events cornerstone.
The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance
Enclosure
26
Determination Process (SDP) for Findings At-Power, dated 06/19/12. The inspectors
used the Initial Screening and Characterization of Findings (IMC 0609.04 Exhibit 1,
dated June 19, 2012) to characterize the finding. Since the inspectors answered No to
the Exhibit 1, section B, Initiating Events screening question, the inspectors concluded
that the finding was of very low safety significance (Green).
The primary cause of the performance deficiency, as determined by the inspectors, was
less than adequate work planning and coordination. The inspectors determined that the
cause of this finding was related to the work control component of the human
performance cross-cutting area due to less-than-adequate work planning [H.3 (a)].
Specifically, the licensees procedures used to open the MSIVs that were stuck on their
closed seat did not contain instructions or precautions to limit the magnitude of the force
applied to the valve stems while attempting to open the valves.
Enforcement: The inspectors determined that the finding represents a violation of
regulatory requirements because it involved inadequate operations and maintenance
procedures used to operate safety-related plant equipment. 10 CFR 50 Appendix B
Criterion V requires, in part, that procedures shall include appropriate quantitative or
qualitative acceptance criteria for determining that important activities have been
satisfactorily accomplished. Contrary to the above, the licensees procedures used to
open the loop 2 and loop 3 outboard MSIVs did not provide instructions to limit the
magnitude of the force applied to the valve stems while attempting to open the valves.
As a result of the violation, the loop 2 and loop 3 MSIVs failed in the closed position, and
the reactor was manually tripped on October 8, extending the 1R17 refueling outage for
an additional nine days. The licensee conducted ultrasonic testing on the remaining six
Unit 1 MSIVs to verify that the valve stems were intact. The stems of the loop 2 and
loop 3 outboard MSIVs were replaced, and the Unit 1 reactor was restarted on October
17. Because this violation was of very low safety significance and it was entered into the
licensees corrective action program as CR 530916, this violation is being treated as an
NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000424/2012005-03, Inadequate Operations and Maintenance Procedures Results in
Brittle Failure of the Loop 2 and Loop 3 Outboard MSIV Stems.)
.3 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a review of the licensees Corrective Action Program and
associated documents to identify trends which could indicate the existence of a more
significant safety issue. The review was focused on repetitive equipment issues, but
also considered the results of inspector daily CR screening and the licensees trending
efforts. The review nominally considered the six month period of April 2012 through
September 2012 although some examples extended beyond those dates when the
scope of the trend warranted. The inspectors also reviewed several CRs associated
with operability determinations which occurred during the period. Corrective actions
associated with a sample of the issues identified in the licensees trend reports were
reviewed for adequacy. The inspectors also evaluated the trend reports against the
Enclosure
27
requirements of the licensees corrective action program as specified in licensee
procedure NMP-GM-002, Corrective Action Program, and 10 CFR 50, Appendix B.
b. Findings and Observations
No findings were identified.
4OA3 Follow-up of Events and Notices of Enforcement Discretion (71153)
.1 (Closed) LER 05000424/2012-003-00 Failure to Comply with Technical Specification
LCOs 3.7.14 and 3.0.3
a. Inspection Scope
On August 17, 2012, 1A ESF Chiller condenser vacuum was noted to be 12 inches of
mercury, with a vacuum of 15 inches of mercury specified as the low limit on operating
logs. The shift supervisor mistakenly believed condenser pressure was one of the
parameters which engineering had evaluated and was continuing to monitor with a
recorder. Condenser pressure was not one of the parameters being monitored and
recorded on a recorder. When the condenser pressure was recorded as out of
specification on the operator rounds log sheet, the shift supervisor failed to initiate
operability and reportability determination processes. This misinformation was carried
forward through subsequent shifts via logs. During the next five days, 1A ESF Chiller
condenser vacuum decreased to 4 inches of mercury and stabilized for an additional
four days prior to initiation of a CR on August 26, 2012. Subsequent investigation and
consultation with the vendor determined the 1A ESF Chiller was inoperable and the TS
LCO was entered at 1437 on August 26, 2012. As a result of the delay in recognition of
the status of the subject chiller, appropriate actions of LCOs 3.7.14 and 3.0.3 were not
taken. The inspectors reviewed the LER, the associated CR and enhanced apparent
cause determination, and subsequent action items.
b. Findings
One licensee-identified violation was identified, and is documented in section 4OA7 of
this report. This LER is closed.
4OA5 Other Activities
.1 Quarterly Resident Inspector Observations of Security Personnel and Activities
a. Inspection Scope
During the inspection period, the inspectors conducted observations of security force
personnel and activities to ensure that the activities were consistent with licensee
security procedures and regulatory requirements relating to nuclear plant security.
These observations took place during both normal and off-normal plant working hours.
Enclosure
28
These quarterly resident inspector observations of security force personnel and activities
did not constitute any additional inspection samples. Rather, they were considered an
integral part of the inspectors normal plant status review and inspection activities.
b. Findings and Observations
No findings were identified.
.2 (Discussed) Temporary Instruction 2515/187 - Inspection of Near-Term Task Force
Recommendation 2.3 Flooding Walkdowns
a. Inspection Scope
Inspectors conducted independent walkdowns to verify that the licensee completed the
actions associated with the flood protection feature specified in paragraph 03.02.a.2 of
this TI. Inspectors are performing walkdowns at all sites in response to a letter from the
NRC to licensees, entitled Request for Information Pursuant to Title 10 of the Code of
Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the
Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident, dated
March 12, 2012 (ADAMS Accession No. ML12053A340).
Enclosure 4 of the letter requested licensees to perform external flooding walkdowns
using an NRC-endorsed walkdown methodology (ADAMS Accession No.
ML12056A050). Nuclear Energy Industry (NEI) document 12-07 titled, Guidelines for
Performing Verification Walkdowns of Plant Protection Features, (ADAMS Accession
No. ML12173A215) provided the NRC-endorsed methodology for assessing external
flood protection and mitigation capabilities to verify that plant features, credited in the
CLB for protection and mitigation from external flood events, and are available,
functional, and properly maintained.
b. Findings and Observations
Findings or violations associated with the flooding, if any, will be documented in the 1st
quarter integrated inspection report of 2013.
.3 Temporary Instruction 2515/188 - Inspection of Near-Term Task Force
Recommendation 2.3 Seismic Walkdowns
a. Inspection Scope
The inspectors accompanied the licensee on their seismic walkdowns of the following
- Unit 1 Diesel Fuel Oil Transfer Pump A, SWEL 1 item #60, on August 15 in the
Diesel Fuel Oil Storage Tank Building
- Unit 1B Diesel Generator Control Panel, SWEL 1 item #61, on August 15 in the
Diesel Generator Building
Enclosure
29
16 in the AFW Pump House
- Unit 2 Spent Fuel Pool Heat Exchanger B, SWEL 2 item #1, on August 21 in the
Auxiliary Building
The inspectors verified that the licensee confirmed that the following seismic features
associated with the above listed components were free of potential adverse seismic
conditions:
- Anchorage was free of bent, broken, missing or loose hardware
- Anchorage was free of corrosion that is more than mild surface corrosion
- Anchorage is free of visible cracks in the concrete near the anchors
- Anchorage configuration was consistent with plant documentation
- SSCs will not be damaged from impact by nearby equipment or structures
- Overhead equipment, distribution systems, ceiling tiles and lighting, and masonry
block walls are secure and not likely to collapse onto the equipment
- Attached lines have adequate flexibility to avoid damage
- The area appears to be free of potentially adverse seismic interactions that could
cause flooding or spray in the area
- The area appears to be free of potentially adverse seismic interactions that could
cause a fire in the area
- The area appears to be free of potentially adverse seismic interactions associated
with housekeeping practices, storage of portable equipment, and temporary
installations (e.g., scaffolding, lead shielding)
The inspectors independently performed their walkdowns and verified that the following
components were free of the potential adverse seismic conditions listed above:
- Unit 2A Diesel Generator Air Start Receiver #1, SWEL 1 item #55, on December 17
in the Diesel Generator Building
- Unit 1 Spent Fuel Pool Pump B, SWEL item #2, on December 17 in the Auxiliary
Building
Observations made during the walkdowns that could not be determined to be acceptable
were entered into the licensees corrective action program for evaluation.
Additionally, inspectors verified that items that could allow the spent fuel pool to drain
down rapidly were added to the SWEL and these items were walked down by the
licensee.
b. Findings and Observations
No findings were identified.
Enclosure
30
4OA6 Meetings, Including Exit
.1 Exit Meeting
On January 11, the resident inspectors presented the inspection results to Mr. Tom
Tynan and other members of his staff, who acknowledged the findings. The inspectors
confirmed that proprietary information was not provided or examined during the
inspection.
4OA7 Licensee-Identified Violations
The following violations of very low significance (Green) or Severity Level IV were
identified by the licensee and are violations of NRC requirements which meet the criteria
of the NRC Enforcement Policy for being dispositioned as Non-cited Violations.
.1 Failure to Comply with Technical Specification LCOs 3.7.14 and 3.0.3
TS 3.0.3 requires, in part, that when a limiting condition of operation (LCO) is not met
and the associated actions are not met, an associated action is not provided, or if
directed by the associated actions, the unit shall be placed in a mode or other specified
condition in which the LCO is not applicable. TS 3.7.14 require that two engineered
safety feature (ESF) room cooler and safety-related chiller trains shall be operable.
Contrary to the above, on August 17, 2012, at approximately midnight, the unit 1 shift
supervisor failed to enter the required action statement for TS LCO 3.7.14, Condition A
when the unit 1A ESF chiller condenser purge pressure was noted to be out of
specification high. Inoperability of the chiller was not recognized until August 26, 2012,
and the LCO entered at 1437. Further, during the extended period during which the 1A
ESF chiller was inoperable (albeit unrecognized as inoperable), opposite train supported
components as well as redundant room coolers on the train B ESF Chiller and room
cooler train were removed from service for unrelated activities which resulted in two
occasions during which TS 3.0.3 should have been applied. The licensee documented
this event in their corrective action program as CR 507143. Using IMC 0609, dated
June 19, 2012, Attachment 4, Table 2, the inspectors verified that the finding affected
the mitigation systems cornerstone. IMC 0609 Attachment 4 Table 3 directed the
inspectors to use IMC 0609 Appendix A to characterize the finding. Because the finding
represented an actual loss of function of one train of ECCS for greater than its TS
Allowed Outage Time, a detailed risk evaluation was required. A detailed risk evaluation
was performed by a regional senior reactor analyst in accordance with IMC 0609
Appendix A guidance using the NRC Vogtle SPAR model and the Saphire 8 risk analysis
code. An Event/Condition Analysis module in Saphire was run with the unit 1A train ESF
chiller failed with no recovery allowed for a 9 day exposure period. The dominant
sequence was a loss of offsite power with success of reactor trip and emergency power
with late failure of feedwater and failure to implement feed and bleed cooling due to
failure of the Unit 1B train chiller and loss of the safety related switchgear. The detailed
risk evaluation determined that the risk due to the performance deficiency was an
increase in core damage frequency of <1E-6/year, a GREEN finding of very low safety
significance. The risk was mitigated by the availability of alternate train components and
the short exposure period.
Enclosure
31
.2 Failure to Conduct Required ASME Code Section XI Inspections
On April 12, 2012, Vogtle staff identified that in-service inspections for the second 10-
year ISI period were missed for eight ASME Code Class 1 valves. Valves 1/2
1208U6035, 1/2 1208U6036, 1/2 1208U6037 and 1/2 1208U6038 are chemical and
volume control system normal and alternate charging check valves to the reactor coolant
system. Leakage control devices (seal encapsulation devices) were installed on the Unit
1 valves in 1987 to address recurring body-to-bonnet leakage per an industry approved
Westinghouse design change. The seal caps were subsequently installed on the unit 2
valves in 1989. Title 10 CFR 50.55a(g)(4) requires, in part, that licensees follow the
pressure test requirements of ASME Code Section XI. ASME Code,Section XI, IWA-
5240, requires visual examinations as part of system pressure tests. ASME Code
Section XI, IWA-5242, 1998 Edition through 2000 addenda, requires VT-2 visual
examinations for pressure retaining bolted connections in borated water systems.
Contrary to the above, from October, 1987, to the present, Vogtle did not perform a
visual inspection of the valve body-to-bonnet studs. This finding was more than minor
because it impacted the initiating events cornerstone and its attribute of equipment
performance. Specifically, it affected the objective to limit the likelihood of those events
that upset plant stability and challenge critical safety functions during shutdown as well
as power operations. Using Inspection manual chapter 0609, dated June 19, 2012,
Appendix A, The Significance Determination Process (SDP) for Findings At-Power, this
finding was determined to be of very low safety significance because the licensees
evaluation was able to demonstrate structural integrity. Specifically, stud stress was not
sufficiently close to the yield stress to cause a loss of integrity. Therefore, the finding
does not contribute to both the likelihood of a reactor trip and the likelihood that
mitigation equipment will not be available. The licensee has entered this issue into their
corrective action program as CRs 438268, 458567, 505111 and 547078. To address
the issue for the short term, the licensee plans to follow the needed and good practice
recommendations detailed by the PWROG in letter OG-12-330 which was issued on
August 16, 2012. The long term corrective actions will be to remove all of the existing
seal caps and install a bonnet with a canopy seal weld to remove the need for a seal cap
as a way to mitigate the effects of leakage and to allow visual examination of the bolted
connections.
.3 Failure to Post High Radiation Area
10 CFR 20.1902(b) requires licensees to post each HRA with a conspicuous sign or
signs bearing the radiation symbol and the words CAUTION, HIGH RADIATION AREA
or DANGER, HIGH RADIATION AREA. Contrary to this, on September 18, 2012, the
entryway into the Unit 1 spent fuel pool cooling system demineralizer valve gallery was
discovered to be missing a conspicuous sign bearing the radiation symbol and the words
CAUTION, HIGH RADIATION AREA or DANGER, HIGH RADIATION AREA.
Accessible areas inside the Valve Gallery room contained dose rates up to 327 mrem/hr
at 30 cm. A HP foreman discovered this violation while performing a walkdown of HRA
postings in the auxiliary building. The licensee took immediate corrective actions upon
discovery including restoration of the HRA posting (CR 519818). There was no
evidence of unauthorized worker entry into the affected area. Although this event
involved the failure to maintain proper control for a HRA, this finding is of very low safety
Enclosure
32
significance because it was not related to ALARA planning, nor did it involve an
overexposure or substantial potential for overexposure, and the ability to assess dose
was not compromised.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel:
S. Bargeron, Plant Manager
B. Brown, Training Manager
R. Collins, Chemistry Manager
K. Dyar, Security Manager
G. Gunn, Licensing
I. Kochery, Health Physics Manager
D. McCary, Maintenance Manager
K. Molina, Heat Exchanger Engineer
D. Puckett, Performance Analysis Supervisor
J. Robinson, Engineering Programs Manager
S. Swanson, Site Support Manager
T. Tynan, Site Vice-President
NRC personnel:
M. Cain, Senior Resident Inspector
T. Chandler, Resident Inspector
F. Ehrhardt, Chief, Region II Reactor Projects Branch 2
T. Lighty, Project Engineer
LIST OF ITEMS OPENED AND CLOSED
Opened and Closed
050000425/2012005-01 NCV Inadequate Maintenance Procedure Results in
Inoperability of NSCW Cooling Tower Fan
(Section 1R19)
05000424, 425/2012005-02 NCV Unauthorized Entry into a High Radiation Area.
(Section 2RS1)05000424/2012005-03 NCV Inadequate Operations and Maintenance
Procedures Results in Brittle Failure of the Loop
2 and Loop 3 Outboard MSIV Stems (Section
4OA2)
Closed
05000424/2012-003-00 LER Failure to Comply with Technical Specification
LCOs 3.7.14 and 3.0.3
05000424/425/2515/188 TI Inspection of Near-Term Task Force
Recommendation 2.3 Seismic Walkdowns
(Section 4OA5.3)
Attachment
2
Discussed
05000424/425/2515/187 TI Inspection of Near-Term Task Force
Recommendation 2.3 Flooding Walkdowns
(Section 4OA5.2)
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
11877-1 Rev. 23, Cold Weather Checklist
11877-2 Rev. 21, Cold Weather Checklist
11901-1 Rev. 17.2, Heat Tracing System Alignment
11901-2 Rev. 12.3, Heat Tracing System Alignment
Section 1R04: Equipment Alignment
Procedures
11903-C Rev. 30.4, Fire Protection System Alignment, Section A
11011-1 Rev. 14.2, Residual Heat Removal System Alignment
13011-1 Rev. 70.3, Residual Heat Removal System
11115-1 Rev. 10.2, Containment Spray System Alignment
Drawings
CX4DB173-1 Rev. 41.0, P&I Diagram Fire Protection - Pump House No. 1 & 2, System 2301
CX4DB173-2 Rev. 29.0, P&I Diagram Fire Protection - Yard Piping System, System 2301
1X4DB122 Rev. 51.0, P&I Diagram Residual Heat Removal System 1205
1X4DB131 Rev. 35.0, P&I Diagram Containment Spray System - System No. 1206
Section 1R05: Fire Protection
Procedures
92799-1 Rev. 3.2, Zone 99 - Control Building Level A Fire Fighting Preplan
92804-1 Rev. 4.2, Zone 104 - MSIV Room Level 1 Fire Fighting Preplan
92745-1 Rev. 2.2, Zone 45 - Auxiliary Building Level 1 Fire Fighting Preplan
92760-2 Rev. 1.0, Zone 60 - Control Building - Level B Fire Fighting Preplan
92761-2 Rev. 1.1, Zone 61 - Control Building - Level B Fire Fighting Preplan
92764-2 Rev. 1.1, Zone 64 - Control Building - Level B Fire Fighting Preplan
92762-2 Rev. 2.0, Zone 62 - Control Building - Level B Fire Fighting Preplan
92763-2 Rev. 0.2, Zone 63 - Control Building - Level B Fire Fighting Preplan
92782-2 Rev. 1.2, Zone 82 - Control Building - Level B Fire Fighting Preplan
92860A-1 Rev. .2, Zone 160A - NSCW Pumphouse - Train A Fire Fighting Preplan
92860B-1 Rev. .2, Zone 160B - NSCW Pumphouse - Train B Fire Fighting Preplan
92771-1, Rev. 4.1, Zone 71 - Control Building - Level B Fire Fighting Preplan
92776-1, Rev. 2.1, Zone 76 - Control Building - Level B Fire Fighting Preplan
92777A-1, Rev. 1.1, Zone 77A - Control Building - Level B Fire Fighting Preplan
92777B-1, Rev. 1.2, Zone 77B - Control Building - Level B Fire Fighting Preplan
92778A-1, Rev. 2.1, Zone 78A - Control Building - Level B Fire Fighting Preplan
92778B-1, Rev. 1.2, Zone 78B - Control Building - Level B Fire Fighting Preplan
Attachment
3
92779A-1, Rev. 0.2, Zone 79A - Control Building - Level B Fire Fighting Preplan
92779B-1, Rev. 1.2, Zone 79B - Control Building - Level B Fire Fighting Preplan
92756A-1, Rev. 0.2, Zone 56A - Control Building - Level B Fire Fighting Preplan
92756B-1, Rev. 1.2, Zone 56B - Control Building - Level B Fire Fighting Preplan
92783-1, Rev. 2.2, Zone 83 - Control Building - Level B Fire Fighting Preplan
92852-1, Rev. 2.2, Zone 152 - Control Building - Level B Fire Fighting Preplan
92771-2, Rev. 1.1, Zone 71 - Control Building - Level B Fire Fighting Preplan
92776-2, Rev. 1.1, Zone 76 - Control Building - Level B Fire Fighting Preplan
92777A-2, Rev. 1.1, Zone 77A - Control Building - Level B Fire Fighting Preplan
92777B-2, Rev. 1.2, Zone 77B - Control Building - Level B Fire Fighting Preplan
92778A-2, Rev. 0.2, Zone 78A - Control Building - Level B Fire Fighting Preplan
92778B-2, Rev. 1.2, Zone 78B - Control Building - Level B Fire Fighting Preplan
92779A-2, Rev. 0.2, Zone 79A - Control Building - Level B Fire Fighting Preplan
92779B-2, Rev. 1.2, Zone 79B - Control Building - Level B Fire Fighting Preplan
92756A-2, Rev. 1.1, Zone 56A - Control Building - Level B Fire Fighting Preplan
92756B-2, Rev. 0.2, Zone 56B - Control Building - Level B Fire Fighting Preplan
92783-2, Rev. 0.2, Zone 83 - Control Building - Level B Fire Fighting Preplan
92852-2, Rev. 0.2, Zone 152 - Control Building - Level B Fire Fighting Preplan
Section 1R06: Flood Protection Measures
Work Orders
SNC 395851
Calculations
X6CXC-27 Rev.8, Flooding Analysis Auxiliary Building Level D
Procedures
25512-C Rev.1.0, General Inspection Outdoor Electrical Duct Run Pull Boxes/Manholes
Drawings
2X4DB147-1 Rev. 18.0, P&I Diagram Aux BLDG. Flood Retaining Rooms Alarms & Drains
System No. 1218
Section 1R07: Heat Sink Performance
Work Orders:
SNC 121039
Procedures
83309-C Rev. 6.4, Safety-Related Heat Exchanger Inspection
83305-C Rev. 7.6, Heat Exchanger Test/Maintenance Program
Condition Reports:
CR 523190
Other
Heat Exchanger as-found and as-left photos taken 09/24/2012
Attachment
4
Section 1R11: Licensed Operator Requalification Program
Other
V-RQ-SE-12705, Simulator Exercise Guide: Steam Generator Tube Leak
Procedures
18009-C Rev. 29.2, Steam Generator Tube Leak
18007-C Rev. 23, Chemical and Volume Control System Malfunction
Section 1R12: Maintenance Effectiveness
Engineering Documents
System 1301-2, Main Steam System 2012 health reports
System 1205 RHR System, 4th quarter 2012 system health report
Condition Reports:
348759, 2PV3000 has exceeded its MR OOS time
343991, 2PV3000 SG-1 ARV Failure
518228, 1HV8702A will not open
534190, MR evaluation identified FF
Procedures
50028-C Rev. 18.1, Engineering Maintenance Rule Implementation
Section 1R15: Operability Evaluations
Condition Reports:
501741, U-1 D/G A unit available light is not working
533822, Unit 1 and 2 main steam isolation valves (MSIVs)
530916, Unit 1 steam generator loops 2 & 3 do not indicate steam flow
Engineering Documents
RER SNC440108, MSIV stem failure evaluation
DOEJ-VRSNC440108-M001, Unit 1 MSIVs ((1HV3016B & 1HV3026B) disk separation
Other
VNP-12-008 REV. 1.0, ODMI worksheet for degraded MSIV stems
Operating Experience Smart Sample (OpESS) 2012/02, Revision 1, Technical Specification
Interpretation and Operability Determination
Section 1R18: Plant Modifications
Other
MDC SNC439341/1.0, 1-1208-U4-A11, Auxiliary Pressurizer Spray Valve Seal Leak
Section 1R19: Post Maintenance Testing
Procedures
25022-C Rev. 7, General Small Bore Piping Installation
14240-1 Rev. 5.2, Manual SLI TADOT
14850-1 Rev. 52, Cold Shutdown Valve Inservice Test
14980A-2 Rev. 23, Diesel Generator Operability Test
14825-1 Rev. 97, Quarterly Inservice Valve Test
14430-2 Rev. 9.0, NSCW Cooling Tower Fans Monthly Test
Attachment
5
Condition Reports
523221
530916
550378
Work Orders
SNC410412
SNC439646
SNC439647
SNC439900
SNC424222
SNC387201
SNC127251
SNC447741
Section 1R22: Surveillance Testing
Procedures
14666-1 Rev. 34.1, Train A Diesel Generator and ESFAS Test
14372B-1 Rev. 9, Containment Penetration No. 72B RCS Loop #3 Accumulator Sample Line
Local Leak Rate Test
14670A-1 Rev. 1.2, Diesel Generator 1A Hot Restart Test
14808A-2 Rev. 2.2, Train A Centrifugal Charging Pump and Check Valve IST and Response
Time Test
14804A-1 Rev. 4, Safety Injection Pump A Inservice and Response Time Test
14905-1 Rev. 67.2, RCS Leakage Calculation (Inventory Balance)
14905-2 Rev. 50.1, RCS Leakage Calculation (Inventory Balance)
14802A-1 Rev. 4.1, Train A NSCW Pump/Check Valve IST and Response Time Test
Work Orders
SNC356759
SNC335801
SNC371576
SNC385697
SNC380426
SNC392870
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Change Packages
Emergency Plan, Revision 57
NMP-EP-110, Emergency Classification Determination and Initial Action, Version 3.0
NMP-EP-110, Emergency Classification Determination and Initial Action, Version 4.0
NMP-EP-111, Emergency Notifications, Version 7.0
NMP-EP-112, Protective Action Recommendations, Version 2.0
91201-C, Activation and Operation of the Technical Support Center, Revision 17
91202-C, Activation and Operation of the Operations Support Center, Revision 22
Attachment
6
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Procedures, Guidance Documents, and Manuals
43014-C, Special Radiological Controls, Ver. 46
46024-C, Release of Materials from the RCA, Ver. 9.3
93240A-C, Reactor Vessel Disassembly Instructions, Ver. 33
NMP-HP-300, Radiation and Contamination Surveys, Ver. 2.0
NMP-HP-302, Restricted Area Classification, Postings, and Access Control, Ver. 2.1
NMP-HP-303, Personnel Decontamination, Ver. 1.4
NMP-HP-305, Alpha Radiation Monitoring, Ver. 4.0
NMP-CH-013-001, Vogtle RCS Chemistry During Scheduled Plant Shutdowns with Fuel
Defects Suspected, Ver. 1.0
NMP-GM-002, Corrective Action Program, Ver. 12.1
Records and Data
NMP-HP-109, Investigation, Evaluation and Management of Damaged, Lost, Malfunctioning or
Alarming Dosimetry, Ver. 1.1, Data Sheet 2, Workers Statement for Dose/Rate Alarm or
Malfunctioning Electronic Dosimeter, 9/26/12
Unit 1 Spent Fuel Pool Inventory Log Non Fuel Radioactive Material Stored in Unit 1 Spent Fuel
Pool
Contingency Plan for: Health Physics Controls for High Noble Gas Activity During Code Safety
Valve Removal During 1R17
ALARA Briefing Notes: Rx Head/ Upper Internals Lift and Set
Alpha Levels, Comparing 100 cm2 Samples
RWP 12-1618, Repair of Unit 1 Transfer Canal Transfer Tube ISO Valve 1-1213-U6-086, Rev. 0
RWP 12-1505, Under Rx Vessel Work, Rev. 0
RWP 12-1608, Remove/Install Pzr Code Safety Valves, Rev. 0
RWP 12-1600, CM/PM, VLVS, Motors, Misc Activities in U1 CTMT, Rev. 0
Radiological Survey 124805, U1 SFP Transfer Canal
Radiological Survey 124876, U1 SFP Transfer Canal
Radiological Survey 152222, Unit 1 Transfer Canal - Top View (1FHBC1)
Radiological Survey 151559, Valve Gallery SFP Cooling Demin (1AXA46)
Radiological Survey 153033, Inside Bioshield Area (1RXC)
Radiological Survey 152915, Inside Bioshield Area (1RXC)
Radiological Survey 152836, Inside Bioshield Area (1RXC)
Radiological Survey 152699, Seal Table Rack Pulled Back (1RXA3)
Radiological Survey 152711, Seal Table (1RXA3)
Radiological Survey 151834, Auxiliary Building Level D (CAXD)
Radiological Survey 9/18/2012 15:38, Under Vessel
Radiological Survey 152143, Top of PZR - 252 el. (1RX32)
Radiological Survey 152054, Top of PZR - 252 el. (1RX32)
Radiological Survey 135176, Top of PZR - 252 el. (1RX32)
Radiological Survey 152283, S/G 4 Channel Head Cold Leg (1RXB27)
Radiological Survey 152282, S/G 4 Channel Head Hot Leg (1RXB26)
Radiological Survey 152281, S/G 1 Channel Head Cold Leg (1RXB19)
Radiological Survey 152281, S/G 1 Channel Head Cold Leg (1RXB19)
Radiological Survey 151964, U1 Containment Building Level 1 (1RX1)
Radiological Survey 152549, U1 Containment Level A Lower Overview (1RXA)
Radiological Survey 151986, U1 Containment 184 El. Quadrant 1 (1RXB)
Attachment
7
Radiological Survey 152662, U1 Containment 197 El. Quadrant 1 (1RXA)
Radiological Survey 151910, U1 Containment 197 El. Quadrant 1 (1RXA)
Radioactive Sealed Source Leak Test Certification, Source ID 0413-00-00, 6/22/12, 12/5/11,
6/28/11, 1/19/10
Air Sample Log, 9/18/12 - 9/20/12
CAP Documents
V-HP-2011, Fleet Oversight Audit of Health Physics, 8/15/11
CR 382120
CR 335736
CR 519818
CR 527832
CR 523976
CR 434496
CR 434498
CR 481012
Section 2RS2: Occupational ALARA Planning and Controls
Procedures, Guidance Documents, and Manuals
16035-1, Chemistry Operations Interface for RCS Chemistry Control During Scheduled Plant
Shutdowns, Ver. 13
43031-C, Steam Generator Job Coverage, Rev. 5.3
00910-C, VEGP ALARA Program, Rev. 15.3
41001-C, ALARA Job Review, Rev. 33.1
NMP-AD-035, ALARA Program, Ver. 1.0
NMP-HP-204, ALARA Planning and Job Review, Ver. 2.0
43014-C, Special Radiological Controls, Ver. 46
Records and Data
Audit No. V-HP-2011, Fleet Oversight Audit of Health Physics, 8/15/2011
NMP-GM-003-F04 Self-Assessment Final Report; ALARA Outage Planning, CR#2011100194,
12/17/10
Shutdown Chemistry Review: Vogtle Unit 2 Fuel Cycle 15, Final report October 24, 2011
ALARA Committee Meeting Minutes First Quarter 2012
2011 Annual ALARA Report, March 30, 2012
Pre-1R17 ALARA Sub-Committee Meeting Minutes, August 10, 2012
Vogtle 1R17 EPRI Standard Radiation Monitoring Points
Chemistry Crud Burst chart 9/16/12 22:00-9/20/12 22:00
ALARA Steam Generator Work Package:
ALARA Briefing Records
RWP 12-1300, R/I Primary Manway Covers & Diaphragms on S/G 1-4; ALARA Briefing
Record RWP 12-1301, Install and Remove Nozzle Dams for SGs 1-4; ALARA Briefing
Record RWP 12-1302, EDDY Current Testing on SG 1-4 & All Associated Work; ALARA
Briefing Record RWP 12-1303, Sludge Lance on S/G 1-4 & all Associated Work; ALARA
Briefing Record RWP 12-1305,
Radiological Surveys
- 152638, 9/25/12 20:58, S/G Manway Platt #1; #152279, 9/20/12 14:59, S/G Channel Head
Hot leg; #152281 9/20/12 15:30, S/G Channel Head Cold Leg; #152688 9/26/12 11:15, S/G
Attachment
8
Manway Platt 2; #152244 9/20/12 6:25 S/G Channel Head Hot Leg; #152245 9/20/12 6:35
S/G Channel Head Cold Leg; #152690 9/26/12 11:26 S/G Manyway Platt 3; #152246
9/20/12 6:43 S/G 3 Channel Head Hot Leg; #152282 9/20/12 15:57 S/G 4 Channel Head
Hot Leg; #152639 9/25/12 21:05 S/G Manway Platt 4; #152653 9/26/12 1:29 S/G #3 Hand-
Hole Plat
Work-In-Progress Reviews NMP-HP-104 (or equivalent)
RWP 12-1303, 9/27/12 80% Review; RWP 12-1302, 9/23/12 50% Review; RWP 12-1300
9/24/12 50% Review; RWP 12-1301 9/24/12 50% Review; RWP !2-1303 9/25/12 50%
Review; RWP 12-1304 9/25/12 50% Review
NMP-HP-204 Form 8 Respirator Use Evaluation Worksheets for RWP 12-1301
Radiological Surveys:
- 38622, 1/22/02, Plant Vogtle U1 Transfer Canal- top view
- 117917, 7/14/09, U1 Transfer Canal- topview
- 152208, 9/19/12, U1 Transfer Canal- Side view
ALARA Briefing Record RWP 12-1004, Install/Remove Scaffold in U1 CTMT
ALARA Briefing Record RWP 12-1104, Waste and Decon Routines in U1 CTMT
ALARA Briefing Record RWP 12-1618, Repair of U1 Transfer Canal
ALARA Post-Job Report 11-2509, RPV Cold Leg Nozzle Inspection in U2 Annulus
ALARA Post-Job Report 11-2302, Eddy Current Testing on S/G 1&4
CAP Documents
CR 167416, The accumulated dose for the U-1 Aux Bldg ECCS Flowpath Verification that was
performed
CR 339453, HP Self Assessment on High Rad Area Controls
CR 346242, Dose goal exceeded for placing old UFC skid into shipping container
Section 2RS3: In-Plant Airborne Radioactivity Control and Mitigation
Procedures, Guidance Documents, and Manuals
43635-C, Operation and Calibration of the AMS-4 Continuous Air Monitor, Rev. 18.1
47013-C, Inspection, Repair, and Storage of Self Contained Breathing Apparatus, Ver. 37
NMP-HP-301, Airborne Radioactivity Sampling and Evaluation, Ver. 1.2
00970-C, Respiratory Protection Program, Rev. 11.4
NMP-GM-002, Corrective Action Program, Ver. 12.1
Records and Data
Compressed Air/Gas Quality Testing, SCBA Compressor (Demin Plant), 2/5/10, 5/10/10,
8/11/10, 11/9/10, 2/14/11, 5/11/11, 8/25/11, 11/7/11, 2/9/12, 5/7/12, 8/14/12
00012-C, Shift Manning Requirements, Rev. 17.1, Data Sheet 1, Minimum Shift Manning
(Either Unit in Mode 1-4), Day Shift, 9/21/12
Qualification List, ID V-GE-099H, MSA High Pressure Firehawk MMR SCBA, Maintenance,
Health Physics, and Operations
Qualification List, ID V-HP-999 MSA C.A.R.E. Tech
Control Room Monthly Emergency SCBA Inspection, 8/7/12
TSC Monthly Emergency SCBA Inspection, 8/10/12
SCBA PosiChek3 Test Results, SCBA Unit HP-82, 11/26/10, 12/30/10, 11/23/11
SCBA Posichek3 Test Results, SCBA Unit HP-61, 12/29/10, 12/28/11
SCBA Posichek3 Test Results, SCBA Unit HP-58, 8/24/10, 8/25/11, 8/27/12
Maintenance Record Report, SCBA Units HP-58, HP-61, HP-82, 8/1/10 - 9/16/12
Attachment
9
CAP Documents
V-HP-2011, Fleet Oversight Audit of Health Physics, 8/15/11
CR 167804
Section 2RS4: Occupational Dose Assessment
Procedures, Guidance Documents, and Manuals
NMP-HP-107-001, Instructions for Retrieving, Printing and Updating Individual Radiation
Exposure Records, Version 1.0
NMP-HP-105, Comparisons of OSLD and ED Dosimetry Results, Ver. 1.0
NMP-HP-106, Investigating of Exposures Exceeding Fleet Administrative Limits, Ver. 1.0
NMP-HP-103, Skin Dose Assessment, Ver. 1.1
NMP-HP-100, Bioassay Program, Ver. 1.0
NMP-HP-101, In-Vivo Bioassay and Internal Dose Assessment, Ver. 1.0
45013-C, Issuance, Use and Collection of Personnel Dosimetry, Ver. 26.3
Records and Data
Canberra Report of Performance Testing Results for Nuclear Enterprises (NE) Model SPM
904B/906 Personnel Portal Monitor, May 18, 2012
Personnel Contamination Events/Personnel Contamination Reports (PCE/PCR) Logs, 9/2011-
9/2012
EDE & NRC Form 5 Calculations for Steam Generator Multibadging Jobs entry made on
9/20/12; Multibadge RCA Authorization/Worksheets
NMP-HP-109 Data Sheets, Investigation of Lost, Damaged or Malfunctioning Personnel
Dosimetry, for occurrence on 9/26/12
NMP-HP-109 Data Sheet 2, Investigation of Lost, Damaged or Malfunctioning Personnel
Dosimetry, for occurrence on 9/28/12
Varskin 3 Calculations 10/5/2011
CAP Documents
CR 149580, On 4-16-10 there were (62) EPD's and (1) EPD "Vibe Mesh Transmitter" found at
the RPF control room
CR 152467, A vendor arrived on site on June 22 at 08:45 am. and asked for a source that was
shipped to Plant Vogtle.
CR 466572, HP Tier 1 outage readiness Health Physics Tier 1 outage review was performed
with the HP Manager.
Section 2RS5: Radiation Monitoring Instrumentation
Procedures, Guidance Documents, and Manuals
17005-1, Annunciator Response Procedures for ALB 05 on Panel 1A2 on MCB, Ver. 33
17100-2, Annunciator Response Procedure for the Process and Effluent Radiation Monitoring
System (RMS), Rev. 20.1
24625-1, Containment High Range (2RE-0005) Area Monitor 2RX-0005 Channel Calibration,
Rev. 29.1
24652-1, Plant Vent Wide Range Radiogas Monitor 1RX-12444 Channel Operational Test and
Channel Calibration, Rev. 19
34313-C, Operation of the DRMS Plant Vent Effluent Wide Range Monitor 1(2) RE-12444,
Rev 17
43802-C, Calibration of Gamma Standards, Rev. 12.1
Attachment
10
43500-C, Health Physics Instrument Calibration and Control Program, Rev. 53.10
43685-C, Calibration and Operation of the Asp-1, Rev 20.3
43693-C, Operation and Use of the Jl Shepherd Model 89-400 Shielded Calibrator, Rev. 2.1
NMP-HP-703, RO-2, RO-2A and RO-20 Operation and Calibration, Ver.1.0
NMP-HP-703, Daily Instrumentation Source Checks, Ver 1.1
NMP-EP-110-GL03, VEGP EALs - ICs, Threshold Values and Basis, Ver. 3.0
Vogtle Electric Generating Plant (VEGP), ODCM, Ver. 28
Records and Data
10 CFR Part 61 Analysis, DAW, Dated 12/21/2011
10 CFR Part 61 Analysis, DAW, Dated 12/28/2010
Chemistry Issues Turnover Log, Dated 09/17/2012
In Situ Calibration of High Range Monitors, VEGP Units 1&2, Dated 02/18/1997
LS 6500 SN 7069842 Source Check Response Charts, Dated 02/11/2011 - 03/28/2012
HPGE Detector #1 Control Charts, Dated 12/25/2011 - 04/03/2012
Radiation Monitor RE-005 Monthly Tech Spec Response Surveillance, Dated May 2010 -
September 2012
Radiation Monitor System Health Reports for 1st and 2nd Quarters 2012
LCO/TR Status Sheet, 1RE006 Failure, initiated 08/14/2010
Special Report 2011-001-00, Inoperable Radiation Monitor 1RE-006, Dated 03/29/2011
RadCal Electrometer Model 2025, SN4676 Calibration Record, Dated 10/31/2011
RadCal Probe Model 20X5-3, SN14166 Calibration Record, Dated 10/31/2011
RadCal Probe Model 20X5-180, SN16121 Calibration Record, Dated 10/31/2011
RadCal Probe Model 20X5-1800, SN21721 Calibration Record, Dated 10/31/2011
Instruments Missing Operation Checks Report, Dated 09/19/2012
Online Instruments Issued Report, Dated 09/19/2012
IPM-7A/8/9, VEGP-HP 0637, Calibration Records, Dated 06/14/2011 and 06/12/2012
Model 28-5 Calibrator, Source VEGP 0292, Calibration Record, Dated 05/18/12
Model 89-400 Calibrator, Source VEGP 0413, Calibration Record, Dated 08/12/2012
Model 878-10 Calibrator, Source VEGP 1049, Calibration Record, Dated 08/12/2012
SAM-11, VEGP-HP 1151, Calibration Records, Dated 05/26/2011 and 05/23/2012
SPM-904B, VEGP-HP 0754, Calibration Records, Dated 02/8/2011 and 02/10/2012
WO 1090817001, Plant Vent Wide Range Radiogas Flow 1F-12444 Channel Operational Test
and Channel Calibration, Dated 03/03/11
WO SNC332418, Plant Vent Wide Range Radiogas Flow 1F-12444 Channel Operational Test
and Channel Calibration, Dated 03/28/2012
WO 1090816001, Plant Vent Wide Range Radiogas Radiation Monitor 1RX12444 Channel
Operational Test and Channel Calibration, Dated 10/15/10
WO 1100068601, 2RE0005 Containment High Range Monitor Channel Calibration,
Dated 08/09/2010
WO 11000408201, 2RE0005 Containment High Range Monitor Channel Calibration,
Dated 03/29/2011
WO 1080853101, 1RE0011 In-Core Instrumentation Room Area Monitor Calibration, Dated
09/26/2009
WO 1100216601, 1RE0011 In-Core Instrumentation Room Area Monitor Calibration, Dated
03/14/2011
WO SNC328096, 1RE0018 Waste Liquid Effluent Channel Operational Test and Calibration,
Dated 03/09/2012
Attachment
11
WO 1090816201, 1RE0018 Waste Liquid Effluent Channel Operational Test and Calibration,
Dated 01/28/2011
WO 1090816201, 1RE0018 Waste Liquid Effluent Isotopic Channel Calibration, Dated
02/23/2012
WBC Calibration Report, Calibration of the Canberra Fastscan-1 WBC System at the Vogtle
Electric Generating Plant, Dated 06/08/2011 and 06/13/2012
WBC Calibration Report, Calibration of the Canberra Fastscan-2 WBC System at the Vogtle
Electric Generating Plant, Dated 06/09/2011 and 06/15/2012
CAP Documents
Fleet Oversight Audit of Health Physics, V-HP-2011, Dated 08/15/2011
CR 285751
CR 302822
CR 410359
CR 321427
CR 472385
CR 483434
CR 481949
CR 490231
CR 521213
Section 4OA1: Performance Indicator Verification
Procedures, Guidance Documents, and Manuals
00163-C, NRC Performance Indicator and Monthly Operating Report Preparation and
Submittal, Ver. 14.3
Records and Data
Liquid Permit Post-Release Data Permit # L-20120725-087-B
Gas Permit Post-Release Data Permit # G-20120813-171-C
Gas Permit Post-Release Data Permit # G-20120808-165-C
Liquid Permit Post-Release Data Permit # G-20120803-091-B
ED Alarm Log Sept 2011-Sept 2012
CAP Documents
CR 349797, Trouble shoot plan for U2 CVCS Demin Results in Resin in Drain Line
CR 474221, Dose equivalent Iodine definition in Tech Specs and ODCM does not match
CR 480703, U-1 RMWST degassifier dissolved oxygen anomaly
CR 354479, Increased dose rates in 2AB-A-72
Section 4OA2: Identification and Resolution of Problems
Procedures/Calculations/Engineering Documents
CAR 196113 - Root Cause and Corrective Actions Evaluation for the Unit 1 Steam Generator
Loop T Mismatch
12001-C, Unit Heat Up to Hot Shutdown (Mode 5 to Mode 4)
14850-1/2, Cold Shutdown Valve In-Service Test
26854-C, MSIV Actuator Maintenance
Attachment
12
Condition Reports
530916 - Unit 1 steam generators 2&3 do not indicate steam flow
Section 4OA7: Licensee Identified Violations
Radiological Survey 151559, Valve Gallery SFP Cooling Demin (1AXA46)
CR 519818
LIST OF ACRONYMS
1R17 Unit 1 Refueling Outage 17
ALARA As Low As Reasonably Achievable
ARM Area Radiation Monitor
CAP Corrective Action Program
CFR Code of Federal Regulations
CR Condition Report
DAW Dry Active Waste
DPW Declared Pregnant Worker
ED Electronic Dosimeter
FSAR Final Safety Analysis Report
HP Health Physics
HPT HP Technician
IP Inspection Procedure
LHRA Locked High Radiation Area
NCV Non-cited Violation
NPR Negative Pressure Respirator
NVLAP National Voluntary Laboratory Accreditation Program
OA Other Activities
OSLD Optically Stimulated Luminescent Dosimeter
PCM Personnel Contamination Monitor
PI Performance Indicator
Radwaste Radioactive Waste
RCA Radiologically Controlled Area
RS Radiation Safety
RWP Radiation Work Permit
S/G Steam Generator
SAM Small Article Monitor
SCBA Self-contained Breathing Apparatus
SDP Significance Determination Process
SFP Spent Fuel Pool
TS Technical Specification
U1 Unit 1
VHRA Very High Radiation Area
WBC Whole Body Counter
Attachment