ML061170008
| ML061170008 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/26/2006 |
| From: | Landis K NRC/RGN-II/DRP/RPB5 |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| References | |
| IR-06-001, IR-06-002 | |
| Download: ML061170008 (44) | |
See also: IR 07200016/2006001
Text
April 26, 2006
Virginia Electric and Power Company
ATTN.: Mr. David A. Christian
Sr. Vice President and
Chief Nuclear Officer
Innsbrook Technical Center - 2SW
5000 Dominion Boulevard
Glen Allen, VA 23060-6711
SUBJECT: NORTH ANNA POWER STATION - NRC INTEGRATED INSPECTION
REPORT NOS. 05000338/2006002, 05000339/2006002 AND
07200016/2006001
Dear Mr. Christian:
On March 31, 2006, the United States Nuclear Regulatory Commission (NRC) completed an
inspection at your North Anna Power Station, Units 1 and 2, and the North Anna Independent
Spent Fuel Storage Installation. The enclosed integrated inspection report documents the
inspection results, which were discussed on April 11, 2006 with Mr. Jack Davis and other
members of your staff.
The inspections examined activities conducted under your licenses as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your
licenses. The inspectors reviewed selected procedures and records, observed activities, and
interviewed personnel.
Based upon the results of this inspection, one self-revealing finding and one NRC-identified
finding of very low safety significance (Green) were identified. The findings were determined to
involve a violation of NRC requirements. However, because of the very low safety significance
and because they were entered into your corrective action program, the findings are treated as
non-cited violations (NCV) consistent with Section VI.A of the NRC Enforcement Policy. In
addition, one licensee- identified violation, which was determined to be of very low safety
significance, is listed in Section 4OA7 of this report. If you contest any non-cited violation in this
report, you should provide a response within 30 days of the date of this inspection report, with
the basis for your denial, to the United States Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional
Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory
Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the North
Anna Power Station.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosure, and your response, if any, will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of
VEPCO 2
NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Kerry D. Landis, Chief
Reactor Projects Branch 5
Division of Reactor Projects
Docket Nos.: 50-338, 50-339,72-016
License Nos.: NPF-4, NPF-7, SNM-2507
Enclosure: Inspection Reports 05000338/2006002, 05000339/2006002, and
07200016/2006-001
cc w/encl: (See page 3)
_________________________
OFFICE RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS RII:DRS RII:DRS
SIGNATURE JTR GJW GWL1 GWL1 for PKV PKV for HJG1
NAME JReece GWilson GLaska GJohnson KVanDoorn BMiller HGepford
DATE 04/21/2006 04/21/2006 04/24/2006 04/24/2006 04/24/2006 04/24/2006 04/24/2006
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
OFFICE RII:DRS RII:DRS RII:DRS RII:DRS
SIGNATURE RCH for NJG1 PKV for RCH
NAME RHamilton JGriffis RChou RHaag
DATE 04/24/2006 04/24/2006 04/24/2006 04/24/2006
E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO
VEPCO 3
cc w/encl:
Chris L. Funderburk, Director
Nuclear Licensing and
Operations Support
Virginia Electric and Power Company
Electronic Mail Distribution
Jack M. Davis
Site Vice President
North Anna Power Station
Electronic Mail Distribution
Executive Vice President
Old Dominion Electric Cooperative
Electronic Mail Distribution
County Administrator
Louisa County
P. O. Box 160
Louisa, VA 23093
Lillian M. Cuoco, Esq.
Senior Counsel
Dominion Resources Services, Inc.
Electronic Mail Distribution
Attorney General
Supreme Court Building
900 East Main Street
Richmond, VA 23219
Distribution w/encl: (See page 4)
VEPCO 4
Letter to David A. Christian from Kerry D. Landis dated April 26, 2006.
SUBJECT: NORTH ANNA POWER STATION - INTEGRATED INSPECTION REPORT
05000338/2006002 AND 05000339/2006002
Distribution w/encl.:
S. Monarque, NRR
L. Slack, RII
RIDSNRRDIPMLIPB
PUBLIC
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-338, 50-339,72-016
License Nos.: NPF-4, NPF-7, SNM-2507
Report Nos.: 05000338/2006002, 05000339/2006002, 07200016/2006001
Licensee: Virginia Electric and Power Company (VEPCO)
Facilities: North Anna Power Station, Units 1 & 2
North Anna Independent Spent Fuel Storage Installation
Location: 1022 Haley Drive
Mineral, Virginia 23117
Dates: January 1, 2006 - March 31, 2006
Inspectors: J. Reece, Senior Resident Inspector
G. Wilson, Resident Inspector
K. Van Doorn, Senior Reactor Inspector, Section 1R08
B. Miller, Reactor Inspector, Section 1R08
R. Chou, Reactor Inspector, Section 1R08
G. Laska, Senior Operations Examiner, Section 1R11
G. Johnson, Operations Engineer, Section 1R11
H. Gepford, Health Physicist, Sections 2OS1, 4OA1, and 4OA5
R. Hamilton, Senior Health Physicist, Sections 2PS2 and 4OA1
J. Griffis, Health Physicist, Section 2OS2
L. Garner, Senior Project Engineer, Section 1R13
Approved by: K. Landis, Chief, Reactor Projects Branch 5
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
REPORT DETAILS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R04 Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R05 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R07 Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R08 Inservice Inspection (ISI) Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
1R11 Licensed Operator Requalification Program . . . . . . . . . . . . . . . . . . . . . . . . . . 10
1R12 Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
1R13 Maintenance Risk Assessments and Emergent Work Evaluation . . . . . . . . . . 12
1R15 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
1R19 Post Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
1R20 Refueling and Other Outages . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
1R22 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
1EP6 Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
2OS1 Access Controls To Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . 18
2OS2 As Low As Reasonably Achievable (ALARA) Planning and Controls . . . . . . . 20
2PS2 Radioactive Material Processing and Transportation . . . . . . . . . . . . . . . . . . . . 21
OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
4OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4OA5 Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
4OA6 Meetings, including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
4OA7 Licensee-Identified Violation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
ATTACHMENT: SUPPLEMENTARY INFORMATION
Key Points of Contact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
List of Items Opened, Closed, and Discussed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2
List of Documents Reviewed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-3
List of Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-11
Enclosure
SUMMARY OF FINDINGS
IR 05000338/2006-002, IR 05000339/2006-002, IR07200016/2006-001; 01/01/2006 -
03/31/2006; North Anna Power Station Units 1 & 2, and North Anna Independent Spent Fuel
Storage Installation. Routine Integrated Resident and Regional Inspector Report. Event
Followup.
The report covered a three-month period of inspection by the resident inspectors and
announced inspections by a senior operations examiner, an operations engineer, a senior
health physicist, two health physicists, a senior reactor inspector, and two reactor inspectors
from the region. One self-revealing finding and one NRC-identified finding were identified. The
findings were determined to be non-cited violations (NCV). The significance of most findings is
indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply
may be Green or be assigned a severity level after NRC management review. The NRCs
program for overseeing the safe operation of commercial nuclear power reactors is described in
NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. An NRC-identified non-cited violation of 10 CFR 50 Appendix B Criterion III was
identified for failure to translate design requirements into procedures. Specifically, the
licensee failed to properly translate the Technical Specification (TS) Operable-
Operability definition into procedures which established the time the environmental
hazard barriers between the turbine building and either the main control room or the
emergency switchgear room were allowed to be inoperable during maintenance. This
issue is documented in the licensees corrective action program as Plant Issues N-
2005-1080 and N-2005-2236.
This issue is more than minor because it could become a more significant condition, in
that the unit could continue to operate at full power with main control room and
emergency switchgear equipment exposed to potentially harsh environmental conditions
(e.g. steam from a high energy line break in the turbine building) for a period of time
greater than that allowed by TS. However, the time period that the pressure boundary
door 2-BLD-STR-S54 was inoperable on March 16, 2005 did not result in a violation of
TS 3.0.3 and thus no performance deficiency existed for that specific event. After
management review, the issue was assigned a significance of Green because the
inoperability period was limited to a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by other TS. (Section 1R13)
Green. A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion III was
identified for inadequate design control resulting in a flood potential for the Units 1 and 2
safeguards instrument rack rooms. On July 9, 2005, back flush of control room chiller
service water strainers 2-HV-S-1A and 1B as directed by engineering transmittal ET
N-05-0034, Operability of 2-HV-P-22C, Service Water Pump for 2-HV-E-4C, was
performed in the Unit 2 air conditioning chiller room (ACCR). Following this work
activity, the licensee observed water around a floor drain in the adjacent air conditioning
fan rooms (ACFR) and initiated Plant Issue N-2005-2565 to evaluate the abnormal
Enclosure
condition. Subsequently, the licensee determined that back-flow preventers were not
installed in the floor drains on the ACFRs on both units. The back-flow preventers are
necessary to prevent leakage in the ACCR from bypassing the flood wall protecting the
ACFR and adjoining safeguards instrument rack room from flooding.
The inspectors determined that the finding had a credible impact on safety based on the
potential for flooding to impact the instrument rack room which contains both trains of
Solid State Protection System cabinets used for engineered safeguards. The finding, if
left uncorrected, would result in a more significant safety concern and is consequently
more than minor. A Phase III evaluation was performed for the SDP due to the loss or
degradation of equipment specifically designed to mitigate a flooding event and the
impact on two trains of a safety system. This evaluation concluded that the
performance deficiency was of very low safety significance (Green) based on the
existence of high level alarms for the associated sumps and the response time allowed
for an operator to isolate the leak (approximately 40 minutes). The inspectors also
concluded that this finding had aspects relating to the cross-cutting area of problem
identification and resolution. (Section 4OA5)
B. Licensee-Identified Violation
One violation of very low safety significance was identified by the licensee and has been
reviewed by the inspectors. Corrective actions taken or planned by the licensee have
been entered into the licensees corrective action program. This violation and corrective
action tracking numbers are listed in Section 4OA7 of this report.
Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 1 and Unit 2 began the inspection period at 100 percent power. Units 1 and 2 remained at
or near 100 percent power for the entire reporting period with the following exceptions. Unit 1
experienced a forced outage February 13 - 17, 2006, due to tube leaks in the 6B and 4B
feedwater heaters. Unit 1 entered a refueling outage on March 12, 2006, which continued
throughout the remainder of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R04 Equipment Alignment
a. Inspection Scope
The inspectors conducted three equipment alignment partial walkdowns to evaluate the
operability of selected redundant trains or backup systems, listed below, with the other
train or system inoperable or out of service. The inspectors reviewed the functional
system descriptions, Updated Final Safety Analysis Report (UFSAR), system operating
procedures, and Technical Specifications (TS) to determine correct system lineups for
the current plant conditions. The inspectors performed walkdowns of selected portions
of the systems to verify that critical components were properly aligned and to identify
any discrepancies which could affect operability of the redundant train or backup
system.
- Unit 2 train B Low Head Safety Injection (LHSI) equipment during planned
maintenance on the 2-SI-P-1A;
- Units 1 and 2 Switchyard, during planned maintenance on the #1 and #3
busses; and,
- Unit 1 train A LHSI, while 1-SI-P-1B was inoperable for motor operated valve
preventative maintenance.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
The inspectors conducted tours of the eleven areas listed below and important to
reactor safety to verify the licensees implementation of fire protection requirements as
described in Virginia Power Administrative Procedure (VPAP)-2401, Fire Protection
Program.
Enclosure
6
The inspectors evaluated, as appropriate, conditions related to: (1) licensee control of
transient combustibles and ignition sources; (2) the material condition, operational
status, and operational lineup of fire protection systems, equipment, and features; and
(3) the fire barriers used to prevent fire damage or fire propagation. Other documents
reviewed are listed in the Attachment.
- Normal Switchgear Room Unit 1 (fire zone 5-1 / NSR-1);
- Emergency Switchgear Room Unit 1 (fire zone 6-1a / ESR-1);
- Emergency Switchgear Room Unit 2 (fire zone 6-2a / ESR-2);
- Charging Pump Cubicle 1-1C (fire zone 11 Ca / CPC-1C);
- Emergency Diesel Generator 1H Unit 1 (fire zone 9A-1a / EDG-1H);
- Motor-Driven Auxiliary Feedwater Pump Room Unit 1 (fire zone 14B-1a /
MDAFW-1);
- Battery Room 1 - I Unit 1 (fire zone 7A-1 / BR1-I);
- Battery Room 1 - II Unit 1 (fire zone 7B-1 / BR1-II);
- Battery Room 1 - III Unit 1 (fire zone 7C-1 / BR1-III);
- Battery Room 1 - IV Unit 1 (fire zone 7D - 1/ BR1-IV); and
- Containment Unit 1 (fire zone 1-1a / RC-1).
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance
a. Inspection Scope
The inspectors reviewed inspection records, test results, maintenance work orders, and
other documentation to ensure that heat exchanger (Hx) deficiencies that could mask or
degrade performance were identified and corrected. The test procedures and records
were also reviewed to verify that these were consistent with Generic Letter 89-13
licensee commitments, and Electric Power Research Institute (EPRI) Heat Exchanger
Performance Monitoring Guidelines. The risk significant Hx reviewed was the Unit 1 B
Component Cooling (CC) Heat Exchanger, which was tagged out for inspection and
cleaning. The inspectors reviewed CC Hx inspection and cleaning procedures,
completed work orders, design specification sheets, and tube plugging margins to verify
that test results were consistent with design acceptance criteria, inspection methods and
performance of the Hx under the current maintenance frequency were adequate, and to
verify minimum flow requirements and Hx design bases were being maintained.
Additionally, the inspectors reviewed Plant Issue N-2006-0257, regarding CC Hx B
elevated differential pressure, for potential common cause problems and other issues
which could affect system performance to confirm that the licensee was entering
problems into the corrective action program and initiating appropriate corrective actions.
Enclosure
7
The inspectors reviewed Hx test condition reports regarding foreign material found
during recent and past CC Hx inspections. In addition, the inspectors conducted a walk
down of all four CC Hxs and the related service water piping to assess general material
condition and to identify any degraded conditions.
b. Findings
No findings of significance were identified.
1R08 Inservice Inspection (ISI) Activities
.1 Piping Systems ISI
a. Inspection Scope
On March 13-17, 2006, the inspectors reviewed the implementation of the licensees ISI
program for monitoring degradation of the reactor coolant system boundary and the risk
significant piping system boundaries for Unit 1. The inspectors selected a sample of
American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,
Section XI required examinations and a sample of risk-informed ISI Program
examinations.
The inspectors conducted an on-site review of nondestructive examination (NDE)
activities to evaluate compliance with TS, ASME Section XI and ASME Section V
requirements, 1995 Edition through 1996 Addenda, and to verify that indications and
defects (if present) were appropriately evaluated and dispositioned in accordance with
the requirements of ASME Section XI, IWB-3000 or IWC-3000 acceptance standards.
Specifically, the inspectors observed the following examinations:
Ultrasonic Testing
- 31"-RC-2-2501R-Q1, Weld #6, Reactor Coolant Elbow to Pipe on A
Steam Generator Crossover Leg;
Generator;
- 6"-WFPD-14-901, Weld #SW-44, Main Feed Line to Bypass Feed Line;
and,
Magnetic Particle
- 6"-WFPD-14-901, Weld #SW-44, Main Feed Line to Bypass Feed Line;
and,
Enclosure
8
The inspectors reviewed the following examination records in addition to the records for
the above observed examinations:
Ultrasonic Testing
- 12"-SI-14-153A-Q2, Weld #41A, Low Head Safety Injection suction
piping;
- 12"-SI-14-153A-Q2, Weld #SW-39, Low Head Safety Injection suction
piping;
- 12"-SI-14-153A-Q2, Weld #85B, Low Head Safety Injection suction
piping; and,
- 12"-SI-14-153A-Q2, Weld #26, Low Head Safety Injection suction piping.
- 12"-SI-14-153A-Q2, Weld #41A, Low Head Safety Injection suction
piping;
- 12"-SI-14-153A-Q2, Weld #SW-39, Low Head Safety Injection suction
piping;
- 12"-SI-14-153A-Q2, Weld #85B, Low Head Safety Injection suction
piping; and,
- 12"-SI-14-153A-Q2, Weld #26, Low Head Safety Injection suction piping.
Qualification and certification records for examiners, inspection equipment, and
consumables along with the applicable NDE procedures for the above ISI examination
activities were reviewed and compared to requirements stated in ASME Section V and
Section XI.
Pressure boundary welding activities associated with ASME Class 2 components were
reviewed to verify the welding process and examinations were performed in accordance
with the ASME Code Sections III, V, IX, and XI requirements. The inspectors reviewed
weld data sheets, the welding procedure specification, supporting welding procedure
qualification records, welder qualification records, weld rod material certifications, and
preservice examination results for the following welds and subsequent weld repairs:
- 10"-SI-214-153A-Q2, Weld #90, Low Head Safety Injection piping and its
associated weld repairs; and,
- 10"-RS-9-153A-Q2, Weld #10, Recirculation Spray piping and its
associated weld repairs.
The inspectors performed a review of piping system related problems that were
identified by the licensee and entered into the corrective action program. The inspectors
reviewed these corrective action documents to confirm that the licensee had
appropriately described the scope of the problems and had implemented effective
corrective actions. Specifically, the inspectors reviewed the licensees augmented
examination activities with respect to through wall leaks found on the LHSI piping during
the operating cycle.
Enclosure
9
b. Findings
No findings of significance were identified.
.2 Boric Acid Corrosion Control ISI
a. Inspection Scope
On March 13-17, 2006, the inspectors reviewed the licensees Boric Acid Corrosion
Control Program (BACCP) to ensure compliance with commitments made in response
to NRC Generic Letter 88-05 Boric Acid Corrosion of Carbon Steel Reactor Pressure
Boundary and Bulletin 2002-01 Reactor Pressure Vessel Head Degradation and
Reactor Coolant Pressure Boundary Integrity.
The inspectors conducted an on-site record review and an independent walk-down of
the reactor building to evaluate compliance with licensee BACCP requirements and 10
CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requirements. In particular,
the inspectors verified that licensee visual examinations focused on locations where
boric acid leaks can cause degradation of safety significant components and that
degraded or non-conforming conditions were properly identified in the licensees
corrective action system.
The inspectors reviewed the licensees program implementation procedures and a
sample of plant issue reports (corrective action documents) to ensure that leaks were
being identified and addressed at an appropriate threshold. A sample review of
engineering evaluations was also completed for boric acid deposits found on reactor
coolant system piping and other ASME Code Class components to verify that the
minimum design code required section thickness had been maintained for any affected
component(s). The inspectors also reviewed the licensees corrective actions
implemented in response to a Green NCV identified during the previous outage on Unit
2. Specifically, the inspectors reviewed corrective actions associated with training on
boric acid identification and reporting and actions associated with the implementation of
boric acid walkdown procedures.
b. Findings
No findings of significance were identified.
.3 Steam Generator Tube ISI
a. Inspection Scope
The inspectors reviewed activities, plans, a pre-outage degradation assessment, and
procedures for the inspection and evaluation of the 1B steam generator Inconel Alloy
690TT tubing, to determine if the activities were being conducted in accordance with TS
and applicable industry standards. Data gathering, analysis, and evaluation activities
were reviewed. The inspectors reviewed data results for tubes R18C16, R16C16,
Enclosure
10
R16C17, R29C51, R24C58, and R03C58 to verify the adequacy of the licensees
primary, secondary, and resolution analyses. The inspectors observed the licensee
perform the video/visual inspection in the lower bowl area of the steam generator to
determine if any foreign materials or debris were present. The inspectors observed the
licensees video probe inspection of the upper tube plate area around the periphery,
down the tube lane under Row 1 U-bends, and in-bundle down selected tube columns.
The inspectors observed the licensees Quality Control examiner oversee the vendor
inspection for foreign objects or debris before the closing of the lower manway in the
steam generator. The inspectors also reviewed data operators and analysts
certifications and qualifications, including medical exams.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program
.1 Biennial Review
a. Inspection Scope
The inspectors reviewed the facility operating history and associated documents in
preparation for this inspection. During the week of January 23, 2006, the inspectors
reviewed documentation, interviewed licensee personnel, and observed the
administration of simulator operating tests associated with the licensees operator
requalification program. Each of the activities performed by the inspectors was done to
assess the effectiveness of the licensee in implementing requalification requirements
identified in 10 CFR 55, Operators Licenses. The evaluations were also performed to
determine if the licensee effectively implemented operator requalification guidelines
established in NUREG-1021, Operator Licensing Examination Standards for Power
Reactors, and Inspection Procedure 71111.11, Licensed Operator Requalification
Program. The inspectors also reviewed and evaluated the licensees simulation facility
for adequacy for use in operator licensing examinations. The inspectors observed two
operator crews during the performance of the operating tests. Documentation reviewed
included written examinations, Job Performance Measures (JPMs), simulator scenarios,
licensee procedures, on-shift records, simulator modification request records and
performance test records, the feedback process, licensed operator qualification records,
remediation plans, watch standing, and medical records. The records were inspected
against the criteria listed in Inspection Procedure 71111.11. Documents reviewed
during the inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
Enclosure
11
.2 Requalifications Activities Review
a. Inspection Scope
The inspectors observed an annual licensed operator requalification simulator
examination on March 7, 2006. The scenerio, Simulator Examination Guide SXG-79,
involved a loss of first stage pressure, a loss of the main feedwater pump, a reactor
coolant pump seal failure, and a faulted steam generator.
The inspectors observed crew performance in terms of communications; ability to take
timely and proper actions; prioritizing, interpreting, and verifying alarms; correct use and
implementation of procedures, including the alarm response procedures; timely control
board operation and manipulation, including high-risk operator actions; and oversight
and direction provided by the shift supervisor, including the ability to identify and
implement appropriate TS actions. The inspectors observed the post training critique to
determine that weaknesses or improvement areas revealed by the training were
captured by the instructors and reviewed with the operators.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
For the two equipment issues listed below, the inspectors evaluated the licensees
effectiveness of the corresponding preventive and corrective maintenance. The
inspectors performed walkdowns of the accessible portions of the systems, performed
reviews of procedures and evaluations, and held discussions with system engineers.
The inspectors compared the licensees actions with the requirements of the
Maintenance Rule (10 CFR 50.65) using VPAP-0815, Maintenance Rule Program, and
Engineering Transmittal CEP-97-0018, North Anna Maintenance Rule Scoping and
Performance Criteria Matrix. Other documents reviewed are listed in Attachment.
- Elevated internal resistance readings obtained for the 1J Emergency Diesel
Generator indicating potential for damaged cells while performing Work Order
(WO) 726061-01; and,
- The maintenance rule criteria for 0-AAC-DG-0M, Station Blackout Generator,
was exceeded, Plant Issue N-2006-0357.
b. Findings
No findings of significance were identified.
Enclosure
12
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
The inspectors evaluated, as appropriate, for the four plant situations listed below: (1)
the effectiveness of the risk assessments performed before maintenance activities were
conducted; (2) the management of risk; (3) that, upon identification of an unforseen
situation, necessary steps were taken to plan and control the resulting emergent work
activities; and (4) that maintenance risk assessments and emergent work problems
were adequately identified and resolved. The inspectors verified that the licensee was
complying with the requirements of 10 CFR 50.65 (a)(4) and the data output from the
licensees safety monitor associated with the risk profile of Units 1 and 2. Other
documents reviewed are listed in Attachment.
- Unit 1 downpower with 1-CC-E-1B, 0-AAC-DG-0M, 1-PT-230.3, 1-PT-33.7
series, rack work and C Reserve Station Service Transformer (RSST) on
overhead lines on January 25, 2006;
- 1-HV-E-4B, 1-IA-C-1, 1-PT-14.2, 1-PT-213.2B.1, 1-PT-213.35B, rack work,
switchyard work and C RSST on overhead lines on February 16, 2006;
- Emergent work on the Alternate AC Diesel Generator with planned work on
instrument racks, switchyard, and C RSST energized on overhead lines on
March 3, 2006; and,
- Emergent work on 2-CC-TV-204B with planned work on instrument racks,
switchyard, 1-PT-83.12H and C RSST energized on overhead lines on March
14, 2006.
In addition, the inspectors completed an in-office review of WO 00494074-06, repair of
control room pressure barrier door 2-BLD-STR-S54-11.
b. Findings
Introduction: A Green, NRC-identified non-cited violation (NCV), involving the Mitigating
Systems Cornerstone, was identified for failure to translate design requirements into
procedures as required by 10 CFR 50 Appendix B Criterion III. Specifically, the licensee
failed to properly translate the TS Operable-Operability definition into procedures which
established the time the environmental hazard barriers between the turbine building and
either the main control room or the emergency switchgear room were allowed to be
inoperable during maintenance.
Description: On March 16, 2005, the inspectors observed that the licensee considered
themselves in a 24-hour limiting condition for operation while performing WO 00494074-
06 on control room pressure barrier door 2-BLD-STR-S54-11. Since the door functions
as a pressure barrier and also separates the harsh environment designated turbine
building area from the mild environment of the control building, this was inconsistent
with the guidance the NRC had issued in Regulatory Issue Summary (RIS) 2001-009,
Control of Hazard Barriers. The licensee initiated Plant Issue 2005-1080 to address the
concern and Plant Issue 2005-2236 to address subsequent ones identified during the
Enclosure
13
resolution of the former plant issue. The licensee determined that due to a
misapplication of the TS Operable-Operability definition regarding environmental hazard
barriers, they had failed to consider that the supported systems and components should
be considered inoperable when environmental hazard barriers become inoperable. This
problem also extended to certain hazard barriers, including flood barriers, at North Anna
and at its sister plant, the Surry Power Station.
Correct application of the definition on March 16, 2005 would have resulted in entry into
TS 3.0.3, which requires either the TS to be exited or a unit to be in Mode 3 within 8
hours. Door 2-BLD-STR-S54-11 was inoperable for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 52 minutes. Thus, no
TS time limits were exceeded.
Until a long term resolution is developed and implemented, the licensee has established
compensatory measures including appropriately specifying entry into TS 3.0.3 when
required and building temporary hazard barriers for planned evolutions.
Analysis: Not establishing TS required limiting conditions for operations into procedures
is a performance deficiency. This issue is more than minor because it could become a
more significant condition, in that the unit could continue to operate at full power with
main control room and emergency switchgear equipment exposed to potentially harsh
environmental conditions (e.g. steam from a high energy line break in the turbine
building) for a period of time greater than that allowed by TS. However, the time period
that the pressure boundary door 2-BLD-STR-S54-11 was inoperable on March 16, 2005
did not result in a violation of TS 3.0.3 and thus no performance deficiency existed for
that specific event. After management review, the issue was assigned a significance of
Green because the inoperability period was limited to a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by other
TS. Hazard barriers are associated with protecting equipment which mitigate accidents
and thus are associated with the Mitigating Systems Cornerstone.
Enforcement: 10 CFR 50 Appendix B Criterion, Design Basis, requires that design
bases be translated into instructions. Contrary to this, on March 16, 2005, design basis,
i.e., TS definition of Operable-Operability, was not translated into instructions such that
unit operation above Mode 3 would be limited to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when environment hazard
barriers were inoperable. Because this finding is of very low safety significance (Green)
and is in the licensees corrective action program as Plant Issues N-2005-1080 and N-
2005-2236, it is being treated as an NCV, consistent with Section VI.A of the NRC's
Enforcement Policy: NCV 05000338, 339/2006002-01, Failure to translate TS operable-
operability definition regarding hazard barriers into instructions as required by 10 CFR
50 Appendix B Criterion III.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed five operability evaluations affecting risk-significant mitigating
systems, listed below, to assess, as appropriate: (1) the technical adequacy of the
evaluations; (2) whether continued system operability was warranted; (3) whether other
Enclosure
14
existing degraded conditions were considered as compensating measures; (4) whether
the compensatory measures, if involved, were in place, would work as intended, and
were appropriately controlled; (5) where continued operability was considered
unjustified, the impact on TS Limiting Conditions for Operation and the risk significance
in accordance with the SDP. The inspectors review included a verification that the
operability determinations were made as specified by Procedure VPAP-1408, System
Operability.
- Plant Issue N-2006-0504, during the performance of procedure 0-PT-77.14B for
in-place testing of the Emergency Core Cooling System Pump Room Exhaust Air
Clean-up System (PREACS) Train B filter, the as found leakage for Unit 2
Safeguards Exhaust bypass dampers were out of spec high with Unit 2
Safeguards Exhaust aligned to the Charcoal Filters;
- Plant Issue N-2006-0520, during the disassembly/inspection of 1-EG-278 check
valve, it was discovered that the in-body seating area of the valve was too wide
and a proper blue check could not be obtained so the valve was declared
operable but degraded;
- Plant Issue N-2006-1175, containment breach via open containment penetration
coolers inside containment and open component cooling drain valves outside
containment;
- Plant Issue N-2006-1387, water found in safety-related conduits for
1-FW-P-3A-MOTOR, 1-SI-P-1A-MOTOR, and 1-RS-P-2A-Motor; and,
- Plant Issue N-2006-1701, for breaker 01-EE-BKR-K/J1-2, the as-found
instantaneous overload setpoints were outside acceptance criteria of Procedure
0-EPM-0302-2.
b. Findings
No findings of significance were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
The inspectors reviewed six post maintenance test procedures and/or test activities, as
appropriate, for selected risk-significant mitigating systems to assess whether: (1) the
effect of testing on the plant had been adequately addressed by control room and/or
engineering personnel; (2) testing was adequate for the maintenance performed; (3)
acceptance criteria were clear and adequately demonstrated operational readiness
consistent with design and licensing basis documents; (4) test instrumentation had
current calibrations, range, and accuracy consistent with the application; (5) tests were
performed as written with applicable prerequisites satisfied; (6) jumpers installed or
leads lifted were properly controlled; (7) test equipment was removed following testing;
and (8) equipment was returned to the status required to perform its safety function.
The inspectors verified that these activities were performed in accordance with licensee
procedure VPAP-2003, Post Maintenance Testing Program.
Enclosure
15
- Procedure 0-MCM-0803-01, Periodic Disassembly, Inspection, and Repair of
the Control Room Chiller Condenser (1/2-HV-E-4A, B) and the Front Office
Chiller Condenser (1-HV-3A, B, and C), Revision 18, per WO 526179-01 for
2-HV-E-4B;
- Procedure 0-MCM-0103-04, Disassembly, Inspection and Repair of
Westinghouse/Nuttall Type SU High Speed Gear Drives (Charging Pump Speed
Increase), Revision 13, and Procedure 0-MCM-0103-01, Repair of the
Charging and High Head Safety Injection Pump, Revision 37, per WO 443561;
- Furmanite leak seal injection of the leaking hinge pin of 2-FW-134, per WO 727585-04;
- Procedure 0-MPM-0102-02, Motor Driven Auxiliary Feed Pumps Preventive
Maintenance, Revision 1, per WO 720465 for the 1-FW-P-3B lube oil cooler
cleaning;
- Procedure 1-PT-30.4.2, NIS Source Range channel 11 (N-32) Calibration,
Revision 5, and 1-ICP-NI-32, MS Source Range Channel 11 (N-32) Calibration,
Revision 0, per WO 730503; and,
- Procedure 82J 1J Emergency Diesel Generator Slow Start Test, Revision 35,
per WOs 734272, 734273, and 726222.
b. Findings
No findings of significance were identified.
1R20 Refueling and Other Outages
.1 Unit 1 Unscheduled Outage
a. Inspection Scope
Unit 1 began an unscheduled outage on February 13, 2006, due to tube leaks in the 6B
and 4B feedwater heaters. The unit cooled down to Mode 4 (approximately 330
degrees F reactor coolant system (RCS) temperature) in order to secure main
condenser vacuum for repairs to 6B feedwater heater. During the forced outage, the
inspectors evaluated the licensees outage activities to verify that appropriate risk
consideration was given in developing schedules and that the licensee adhered to
administrative risk reduction methodologies. The inspectors also monitored the
licensees risk management of off-normal plant conditions and ensured mitigation
strategies were developed for any loss of key safety functions. The unit was
synchronized to the grid on February 17, 2006, and 98% power was obtained on
February 21, 2006. The licensee subsequently performed a coast down in preparation
for a refueling outage.
b. Findings
No findings of significance were identified.
Enclosure
16
.2 Unit 1 Refueling Outage
a. Inspection Scope
The inspectors performed the inspection activities described below for the Unit 1
refueling outage that began on March 12, 2006 and ended April 10, 2006. The
inspectors used inspection procedure 71111.20, Refueling and Outage Activities, to
observe portions of the shutdown, cooldown, refueling, maintenance activities, and
startup activities to verify that the licensee maintained defense-in-depth commensurate
with the outage risk plan and applicable TS.
The inspectors monitored licensee controls over the outage activities listed below.
Documents reviewed during the inspection are listed in the Attachment.
- Licensee configuration management, including daily outage reports, to evaluate
defense-in-depth commensurate with the outage safety plan and compliance
with the applicable TS when taking equipment out of service;
- Installation and configuration of reactor coolant instruments to provide accurate
indication and an accounting for instrument error;
- Controls over the status and configuration of electrical systems and switchyard to
ensure that TS and outage safety plan requirements were met;
- Licensee implementation of clearance activities to ensure equipment was
appropriately configured to safely support the work or testing;
- Decay heat removal processes to verify proper operation and that steam
generators, when relied upon, were a viable means of backup cooling;
- Controls to ensure that outage work was not impacting the ability to operate the
spent fuel pool cooling system during and after-core offload;
- Reactor water inventory controls including flow paths, configurations, and
alternative means for inventory addition, and controls to prevent inventory loss;
- Reactivity controls to verify compliance with TS and that activities which could
affect reactivity were reviewed for proper control within the outage risk plan;
- 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; and,
- While the unit did not enter reduced inventory or mid-loop conditions, procedures
were reviewed for commitments to Generic Letter 88-17 to verify that these
commitments were in place, and distractions from unexpected conditions or
emergent work did not affect operator ability to maintain the required reactor
vessel level.
b. Findings
No findings of significance were identified.
Enclosure
17
1R22 Surveillance Testing
a. Inspection Scope
For the six surveillance tests listed below, the inspectors examined the test procedure,
witnessed testing, and reviewed test records and data packages, to determine whether
the scope of testing adequately demonstrated that the affected equipment was
functional and operable and that the surveillance requirements of the TS were met.
In-Service Tests:
- 2-PT-57.1A, Emergency Core Cooling Subsystem Low Head Safety Injection
Pump (2-SI-P-1A), Revision 48
- 1-PT-64.4A.2, Casing Cooling Pump (1-RS-P-3A) Biennial Test First
Comprehensive Pump Test, Revision 0
Other Surveillance Tests:
- 2-PT-82.2, 2J Diesel Generator Test, Simulated Loss of Offsite Power,
Revision 56
- 2-PT-71.3Q, Unit 2 Motor Driven Auxiliary Feedwater (2-FW-P-3B) Pump and
Valve Test, Revision 29
- 2-PT-33.7, Reactor Trip System Operational Test for Reactor Coolant Pump
(RCP) Bus 2A Undervoltage Test, Revision 9
- 2-PT-82H, 2H Emergency Diesel Generator Slow Start Test, Revision 39
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation
a. Inspection Scope
On February 21, 2006, the inspectors reviewed and observed the performance of an
Emergency Planning Drill that involved a simulation of an earthquake, major break Loss
of Coolant Accident (LOCA), and equipment malfunctions, resulting in a site area
emergency and subsequent general emergency. The inspectors assessed emergency
procedure usage, emergency plan classification, notifications, and the licensees
identification and entrance of any drill problems into their corrective action program.
This inspection evaluated the adequacy of the licensees conduct of the drill and critique
performance. Drill issues were captured by the licensee in their corrective action
program and were reviewed by the inspectors.
Enclosure
18
b. Findings
No findings of significance were identified.
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Controls To Radiologically Significant Areas
a. Inspection Scope
Access Controls. The inspectors reviewed and evaluated licensee guidance and its
implementation for controlling and monitoring worker access to radiologically significant
areas and tasks associated with the 2006 Unit 1 Refueling Outage (RFO). The
inspectors evaluated changes to, and adequacy of, procedural guidance; directly
observed implementation of established administrative and physical radiological
controls; appraised radiation worker and technician knowledge of, and proficiency in
implementing, radiation protection activities; and assessed radiation worker exposures
to radiation and radioactive material.
The inspectors directly observed controls established for radiation workers and Health
Physics Technician (HPT) staff in potential airborne radioactivity area, radiation area,
high radiation area (HRA), locked high radiation area (LHRA), and very high radiation
area (VHRA) locations. Controls and their implementation for LHRA, LHRA > 15 rem/hr,
and VHRA keys, and for storage of irradiated material within the spent fuel pool were
reviewed and discussed in detail. Established radiological controls were evaluated for
selected RFO tasks including transfer canal blind flange removal, upper internals set,
head set, refueling operations, valve maintenance, radioactive waste (radwaste)
processing and storage, and radioactive material/waste shipping activities. In addition,
licensee controls for areas where dose rates could change significantly as a result of
plant shutdown and refueling operations were reviewed and discussed.
For selected tasks, the inspectors reviewed Radiation Work Permit (RWP) details and
attended pre-job briefings to assess communication of radiological control requirements
to workers. Occupational worker adherence to selected RWPs and HPT proficiency in
providing job coverage were evaluated through direct observations, remote
observations, and interviews with licensee staff. Electronic dosimeter (ED) alarm
set-points and worker stay times were evaluated against applicable radiation survey
results. Worker exposure as measured by ED and by licensee evaluations of internal
doses during current refueling outage activities were reviewed and assessed
independently. For HRA tasks involving significant dose gradients, e.g., radiological
surveys of the steam generator (S/G) bowl and reactor cavity entry, the inspectors
evaluated the use and placement of whole body and extremity dosimetry to monitor
worker exposure.
Enclosure
19
Postings and physical controls established within the radiologically controlled area
(RCA) for access to the Unit 1 reactor containment building (RCB), the Unit 1 and Unit 2
reactor auxiliary building (RAB) locations, radioactive material storage locations,
decontamination building, and Independent Spent Fuel Storage Installation (ISFSI) were
evaluated during facility tours. The inspectors independently measured radiation dose
rates or directly observed conduct of licensee radiation surveys and results for the
transfer canal, Unit 1 B S/G bowl, reactor cavity, Unit 1 primary filter, posted LHRAs
within the Unit 1 RCB, and select dose significant areas in the RAB. Results were
compared to current licensee surveys and assessed against established postings and
radiation controls. Licensee controls were observed for selected Unit 1 and Unit 2 RAB
The inspectors evaluated implementation and effectiveness of licensee controls for both
airborne and external radiation exposure. The inspectors reviewed and discussed
selected whole body count analyses conducted between September 2005 and March
2006 to evaluate implementation and effectiveness of personnel monitoring. The
inspectors directly observed processes used for externally contaminated individuals,
including those with potential uptakes of radioactive material. The inspectors reviewed
administrative and physical controls including air sampling, barrier integrity, engineering
controls, and postings for tasks having the potential for individual worker internal
exposures to exceed 30 millirem committed effective dose equivalent.
Radiation protection activities were evaluated against UFSAR, TS, and 10 CFR Parts 19
and 20 requirements. Specific assessment criteria included UFSAR Section 12,
Radiation Protection, TS Section 5.4.1, Procedures, and Section 5.7, High Radiation
Area. Detailed procedural guidance and records reviewed for this inspection area are
listed in Sections 2OS1 and 4OA5 of the Attachment.
Problem Identification and Resolution. An audit, a self-assessment, and licensee
Corrective Action Program (CAP) documents associated with access controls to
radiologically significant areas were reviewed and assessed. The inspectors evaluated
the licensees ability to identify, characterize, prioritize, and resolve the identified issues
in accordance with VPAP-1501, Deviations, Revision 17 and VPAP-1601, Corrective
Action, Revision 21. Licensee CAP documents associated with access control issues,
personnel radiation monitoring, and personnel exposure events which were reviewed
and evaluated in detail during inspection of this program area are identified in Sections
2OS1, 4OA1, and 4OA5 of the Attachment.
The inspectors completed the 21 specified line-item samples detailed in Inspection
Procedure 71121.01.
b. Findings
No findings of significance were identified.
Enclosure
20
2OS2 As Low As Reasonably Achievable (ALARA) Planning and Controls
a. Inspection Scope
Implementation of the licensee's ALARA program during the 2006 Unit 1 RFO was
observed and evaluated by the inspectors. The inspectors reviewed ALARA planning,
dose estimates, and prescribed ALARA controls for outage work tasks expected to incur
the maximum collective exposures. Reviewed activities included containment
scaffolding, head disassembly, manual valve maintenance, air operated valve
maintenance, and routine HP coverage. Incorporation of planning, established work
controls, expected dose rates, and dose expenditure into the ALARA pre-job briefings
and RWPs for those activities were also reviewed. Work in progress reviews were
inspected for three RWPs in which the actual dose was approaching the estimated dose
for the job. Selected elements of the licensee's source term reduction and control
program were examined to evaluate the effectiveness of the program in supporting
implementation of the ALARA program goals. Shutdown chemistry program
implementation and the resultant effect on RCB and RAB dose rate trending data were
reviewed and discussed with cognizant licensee representatives. Small areas with
abnormally high dose rates forming in the steam generator channel heads were
discussed with HP and Chemistry Supervision.
Trends in individual and collective personnel exposures at the facility were reviewed.
The inspectors examined the dose records of all declared pregnant workers during 2005
and first quarter of 2006 to evaluate total or current gestation doses. Applicable
procedures were reviewed to assess licensee controls for declared pregnant workers.
Trends in the plants three-year rolling average collective exposure history, outage,
non-outage, and total annual doses for selected years were reviewed and discussed
with licensee representatives.
The licensee's ALARA program implementation and practices were evaluated for
consistency with UFSAR Chapter 12, Radiation Protection; 10 CFR Part 20
requirements; Regulatory Guide 8.29, Instruction Concerning Risks from Occupational
Radiation Exposure, February 1996; and licensee procedures. Documents reviewed
during the inspection of this program area are listed in Section 2OS2 of the Attachment.
Problem Identification and Resolution. The inspectors reviewed the CAP documents
listed in Section 2OS2 of the report Attachment that were related to the licensees
ALARA program. The inspectors assessed the licensees ability to identify,
characterize, prioritize, and resolve the identified issues in accordance with VPAP-1501,
Deviations, Revision 17 and VPAP-1601, Corrective Action, Revision 21.
The inspectors completed 22 of the specified line-item samples detailed in Inspection
Procedure 71121.02.
b. Findings
No findings of significance were identified.
Enclosure
21
Cornerstone: Public Radiation Safety
2PS2 Radioactive Material Processing and Transportation
a. Inspection Scope
Waste Processing and Characterization. During system walk-downs, the inspectors
observed selected liquid and solid radwaste processing system components for material
condition and system configuration agreement with the UFSAR and Process Control
Program (PCP). Inspected equipment included the high level and low level waste drain
tanks, waste evaporator tank, evaporator test tanks, spent resin hold-up and dewatering
tanks, and associated piping, valves, and pumps. The inspectors discussed component
function, processing system changes, and radwaste program implementation with
licensee staff.
The 2004 Effluent Report and radionuclide characterizations for each major waste
stream were reviewed and discussed with the radwaste staff. For Unit 2 Dry Active
Waste and liquid waste treatment resin, the inspectors evaluated analyses for
hard-to-detect nuclides, reviewed the use of scaling factors, and examined comparison
results between licensee waste stream characterizations and outside laboratory data.
The licensees waste stream mixing and concentration averaging methodology was
evaluated and discussed with radwaste personnel. The inspectors also discussed the
licensees guidance for monitoring changes in waste stream isotopic mixtures with
knowledgeable personnel.
Radwaste processing activities were reviewed for compliance with the licensees PCP
and UFSAR, Chapter 11. Waste stream characterization analyses were reviewed
against regulations detailed in 10 CFR Part 20, 10 CFR Part 61, and guidance provided
in the Branch Technical Position on Waste Classification and Waste Form. Reviewed
documents are listed in Section 2PS2 of the Attachment.
Transportation. The inspectors directly observed preparation activities for the shipment
of pressurizer safety relief valves. The inspectors noted appropriateness of package
markings and placarding and interviewed shipping technicians regarding Department of
Transportation (DOT) regulations. The inspectors observed radiation surveys of the
transport vehicle prior to shipment.
Six shipping records were reviewed for consistency with licensee procedures and
compliance with NRC and DOT regulations. The inspectors reviewed emergency
response information, DOT shipping package classification, radiation survey results, and
evaluated whether receiving licensees were authorized to accept radioactive materials.
Licensee procedures for opening and closing Type B shipping casks were compared to
recommended vendor protocols and Certificate of Compliance requirements. In
addition, training records for selected individuals currently qualified to prepare
radioactive material shipments were reviewed.
Enclosure
22
Transportation program implementation was reviewed against regulations detailed in
10 CFR Part 20, 10 CFR Part 71, 49 CFR Parts 172-178, and the guidance provided in
NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.
Documents reviewed during the inspection are listed in Section 2PS2 of the Attachment.
Problem Identification and Resolution. Two audits and licensee CAP documents were
reviewed and assessed. The inspectors evaluated the licensees ability to characterize,
prioritize, and resolve the identified issues in accordance with procedures VPAP-1501,
Deviations, Revision 17 and VPAP-1601, Corrective Action, Revision 21. Documents
reviewed for problem identification and resolution are listed in Section 2PS2 of the
Attachment.
The inspectors completed the six specified line-item samples detailed in Inspection
Procedure 71122.02.
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification
a. Inspection Scope
Cornerstones: Initiating Events and Barrier Integrity
The inspectors sampled licensee submittals for the three Performance Indicators listed
below for U1 and U2. The inspectors reviewed data from the licensees corrective action
program, maintenance rule records, operating logs and maintenance work orders for the
period covering the first quarter 2005 through the fourth quarter 2005. Discussions with
licensee personnel were held by the inspectors regarding the data reviewed. The data
was compared with that displayed on the NRCs public web site. The performance
indicator method of assessment was compared with the guidelines contained in Nuclear
Energy Institute NEI 99-02, "Regulatory Assessment Indicator Guideline," Revision 3,
and the Performance Indicator Frequently Asked Questions (FAQ) list.
- Unplanned Scrams;
- Scrams with Loss of Normal Hear Removal; and,
- Reactor Coolant System Activity.
Cornerstone: Public Radiation Safety
The inspectors reviewed the Radiological Control Effluent Release Occurrences
Performance Indicator results for the period of January 2005 through December 2005.
For the assessment period, the inspectors reviewed cumulative and projected doses to
the public. The inspectors also reviewed licensee procedural guidance for collecting
Enclosure
23
and documenting Performance Indicator data. Documents reviewed are listed in
Section 4OA1 of the Attachment.
Cornerstone: Occupational Radiation Safety
The inspectors reviewed the Occupational Exposure Control Effectiveness Performance
Indicator results from July 2005 through March 2006. For the assessment period, the
inspectors reviewed documented electronic dosimeter alarms and CAP documents
related to controls for exposure significant areas. The inspectors also reviewed licensee
procedural guidance for collecting and documenting Performance Indicator data.
Report Section 2OS1 contains additional details regarding the inspection of controls for
exposure significant areas. Documents reviewed are listed in Sections 2OS1 and 4OA1
of the Attachment.
During plant tours the inspectors also periodically assessed the Occupational Exposure
Control Effectiveness and the RETS/ODCM Radiological Effluent Occurrence
Performance Indicators by determining if high radiation areas (>1R/hr) were properly
secured and looking for unmonitored radiation release pathways.
b. Findings
No findings of significance were identified. The performance indicators all remained in
the licensee response band (Green).
4OA2 Identification and Resolution of Problems
.1 Daily 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 reviewing daily
Plant Issues summary reports and periodically attending daily Plant Issue Review Team
meetings.
.2 Annual Sample Review
a. Inspection Scope
The inspectors reviewed the licensees assessments and corrective actions for Plant
Issue N-2005-3462, 2H emergency diesel generator developed a coolant leak on the
control side between the #1 and #3 cylinders, during performance of 2-PT-82H. EDG
was manually unloaded and shutdown prior to completion of PT. The Plant Issue was
reviewed to ensure that the full extent of the issue was identified, an appropriate
evaluation was performed, and appropriate corrective actions were specified and
prioritized. The inspectors also evaluated the Plant Issue against the requirements of
Enclosure
24
the licensees corrective action program as specified in VPAP-1601, Corrective Action
Program, VPAP-1501, Deviations, and 10 CFR 50, Appendix B.
b. Findings and Observations
There were no findings of significance identified. On September 6, 2005, the licensee
initiated Plant Issue N-2005-3462 in response to coolant leaks on the Unit 2H
Emergency Diesel Generator (EDG) identified during the monthly surveillance test. The
licensee subsequently completed a functional evaluation and declared a Generic Letter 91-18 condition (operable but degraded) for the component. The inspectors verified the
licensees functional evaluation which considered that the leakrate of the emergency
diesel generator was within the tanks makeup capability and makeup could be
completed without any extraordinary efforts. The inspectors reviewed the history of
coolant leaks for the EDGs which included Plant Issue N-2005-0101 and work order 526277-01, for coolant leaks on the Unit 1H EDG, which was completed on January 9,
2005. During that time, the licensee identified that the water by-pass fitting gaskets that
were manufactured by Cogemica should be replaced with the new model 3000 gaskets
that were manufactured by Garlock. The licensee subsequently scheduled the
replacement of the Unit 1 EDGs gaskets during the 2005 diesel inspections and Unit 2
EDGs gaskets during the 2006 diesel inspections. During the fall of 2005, the
licensees 2H and 2J EDGs incurred failures of their water by-pass fitting gaskets (Plant
Issue N-2005-3462/3633), which resulted in more coolant leaks. The licensee
appropriately determined to review their previous corrective actions and concluded that
their corrective actions were untimely; the inspectors agreed with their conclusion. The
replacement of the by-pass gaskets on the Unit 2 EDGs was subsequently rescheduled
to September 2005 to correct the failed gaskets. The inspectors used IMC 0612 to
review the licensees actions and determined that because the failure of the water
by-pass fitting gaskets did not render the EDGs inoperable, the licenses untimely
corrective action for a condition adverse to quality would be considered a minor violation
of 10 CFR 50, Appendix B Criterion XVI.
4OA5 Other Activities
.1 Independent Spent Fuel Storage Installation (ISFSI) Radiological Controls
a. Inspection Scope
The inspectors conducted independent gamma and neutron surveys of the ISFSI facility
and compared the results to previous surveys. The inspectors also observed and
evaluated implementation of radiological controls, including RWPs and postings, and
discussed the controls with an HPT and HP supervisory staff. Radiological controls for
loading the ISFSI casks were also reviewed and discussed. The inspectors reviewed
environmental thermoluminescent dosimeter records and discussed the use of the
dosimeters and resultant neutron/gamma data with cognizant HP supervisory staff.
Radiological control activities for ISFSI areas were evaluated against 10 CFR Part 20,
10 CFR Part 72, and applicable licensee procedures. Documents reviewed are listed in
Section 4OA5 of the Attachment
Enclosure
25
b. Findings
No findings of significance were identified.
.2 (Closed) Unresolved Item (URI) 05000338, 339/2005004-02: Inadequate Design
Control Results in Safeguards Instrument Rack Room Flood Problem
Introduction. A Green self-revealing non-cited violation was identified for inadequate
design control. Specifically, back-flow preventers were not installed in floor drains that
resulted in a flood potential for the Units 1 and 2 Safeguards Instrument Rack Rooms.
This finding was initially characterized as contrary to the requirements of 10 CFR 50,
Appendix B, Criterion XVI, Corrective Action. Subsequent review resulted in a finding
which was contrary to the requirements of Criterion III, Design Control.
Discussion. On July 9, 2005, back flush of control room chiller service water strainers
2-HV-S-1A and 1B as directed by engineering transmittal, ET N-05-0034, Operability of
2-HV-P-22C, Service Water Pump for 2-HV-E-4C, was performed in the Unit 2 air
conditioning chiller room (ACCR). Following this work activity, the licensee observed
water around a floor drain in the adjacent air conditioning fan rooms (ACFR), and
initiated Plant Issue N-2005-2565 to evaluate the abnormal condition. Subsequently,
the licensee determined that back-flow preventers were not installed in the floor drains
on the ACFRs on both units. This design requirement is necessary to ensure the
functionality of the flood walls between the ACCR and adjacent ACFR on Units 1 and 2.
Therefore, the licensee initiated a flood watch, declared the flood walls inoperable, and
entered a Yellow six-day maintenance rule risk condition based on the unavailability of
the flood walls to perform their function. The ACFRs on both units are adjacent and
open to the safeguards instrument rack rooms, which contain the solid state protection
system (SSPS) and process instrumentation and are at a two feet lower elevation. Each
instrument rack room has a sump with two pumps rated at 40 gpm each. On Unit 2, the
sump pumps discharge line is hard-piped directly to the ACCR sump. However, on Unit
1 the sump pumps discharge line is routed to a drain funnel interconnected to the floor
drain system of the adjacent ACFR. The licensee determined that this funnel did not
have a back-flow preventer installed and initiated Plant Issue N-2005-2597. Calculation
ME-0782 was performed by the licensee to evaluate the consequences of a service
water line break in either the Unit 1 or 2 ACCR. The calculation concluded that the peak
flow rate from the Units 1 and 2 ACCR to the adjacent ACFR via the floor drain piping
was 182.9 gpm and 169.4 gpm respectively, which exceeds the capacity of the sump
pumps.
The inspectors reviewed the licensees corrective action database and determined that
on October 15, 2004, Plant Issue N-2004-4554 was initiated due to water discharge
from a capped floor drain outside of the ACCR. An other evaluation was assigned to
engineering to review this condition for impact on the flood protection assumed for the
ACCR and connecting areas as applicable. This evaluation did not identify and correct
the absence of back-flow preventers in the adjacent ACFR floor drains. The inspectors
also identified the following related Plant Issues that did not result in the identification
and correction of this problem:
Enclosure
26
- N-1999-3405, which documented operational experience from Three Mile Island
regarding check valves missing from floor drains and the impact on flood
protection; and,
- N-1990-0020, IN 83-44-S1: Potential damage to redundant safety equipment as
a result of backflow through the equipment and floor drain system.
The inspectors concluded that the failure to install the back-flow preventers is contrary
to the requirements of 10 CFR 50, Appendix B, Criterion III, which requires in part that
measures shall be established to assure the design basis for those structures, systems
and components (SSCs) to which this appendix applies are correctly translated into
specifications, drawings, and procedures.
Analysis. The inspectors determined that the finding had a credible impact on safety
based on the potential for flooding to impact both trains of SSPS cabinets used for
engineered safeguards. The inspectors referenced IMC 0612 and determined that if left
uncorrected this finding would result in a more significant safety concern and is
consequently more than minor. Based on a review of IMC 0609 for the SDP, the
inspectors determined the finding would require a Phase III evaluation due to the loss or
degradation of equipment specifically designed to mitigate a flooding event and the
impact on two trains of a safety system. This evaluation concluded that the
performance deficiency was of very low safety significance (Green) based on the
existence of high level alarms for the associated sumps and the response time allowed
for an operator to isolate the leak (approximately 40 minutes). The inspectors also
concluded that this finding has aspects relating to the cross-cutting area of problem
identification and resolution in that the licensee had multiple opportunities to identify the
condition in their corrective action program (i.e. plant issues N-1990-0020,
N-1999-3405, and N-2004-4554).
Enforcement. 10 CFR 50, Appendix B, Criterion III, Design Control, requires in part that
measures shall be established to assure the design basis for those SSCs to which this
appendix applies are correctly translated into specifications, drawings, and procedures.
Contrary to the above, the licensee failed to ensure that back-flow preventers were
installed in the Unit 1 and 2 ACFRs to ensure the functionality of the flood walls and
safety-related instrumentation in the SSPS rack rooms. This violation is characterized
as a Green NCV and is identified as NCV 05000338, 339/2006002-02, Inadequate
Design Control Results in Safeguards Instrument Rack Room Flood Problem. This
finding is in the licensee's corrective action program as Plant Issue N-2005-2565. URI
05000338, 339/2005004-02, Inadequate Corrective Action Results in Safeguards
Instrument Rack Room Flood Problem, is closed.
Enclosure
27
4OA6 Meetings, including Exit
.1 Exit Meeting Summary
An exit meeting was conducted on January 27, 2006 to discuss the findings of the
biennial Licensed Operator Requalification Program inspection. The inspectors
confirmed that proprietary information was reviewed but is not contained in this report.
Two interim exit meetings were conducted on March 17 and March 24, 2006 with the
site vice-president and ISI/Engineering managers. Additionally, an exit was conducted
on March 30, 2006 to discuss the findings of the Radiation Protection Baseline
Inspection.
On April 11, 2006, the Senior Resident Inspector and the Chief of Reactor Projects
Branch 5 presented the inspection results for the routine integrated quarterly report to
Mr. Jack Davis and other members of the staff. The licensee acknowledged the
findings. The inspectors confirmed that proprietary information was not provided or
examined during the inspection.
.2 Annual Assessment Meeting Summary
On April 18, 2006, the NRC Chief of Reactor Projects Branch 5 met with Virginia Electric
and Power Company to discuss the NRCs Reactor Oversight Process (ROP) and the
North Anna Power Station annual assessment of safety performance for the period of
January 1, 2005 - December 31, 2005. The major topics addressed were the NRCs
assessment program and the results of the North Anna Power Station assessment.
Attendees included North Anna Power Station site management and members of site
staff.
This meeting was open to the public. The presentation material used for the discussion
is available from the NRCs document system (ADAMS) as accession number
ML061140047. ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
4OA7 Licensee-Identified Violation
The following finding of very low safety significance was identified by the licensee and is
a violation of NRC requirements which meets the criteria of Section VI of the NRC
Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
TS 5.4.1 requires that written procedures shall be established, implemented, and
maintained covering the activities in the applicable procedures recommended by
Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, of which part 9.a.
requires procedures for performing maintenance. Contrary to the above, on January 6,
2006, the licensee failed to establish adequate procedure steps in maintenance
procedures 0-MCM-0103-01, "Repair of the Charging and High Head Safety Injection
Pump, and 0-MCM-0103-04, Disassembly, Inspection and Repair of
Westinghouse/Nuttall Type SU High Speed Gear Drives (Charging Pump Speed
Enclosure
28
Increaser). This resulted in tight clearances between the bearing/shaft clearances and
a failure to drain and clean the oil side of the lube oil cooler.
Debris, later found in the lube oil cooler, in conjunction with the tight clearances
subsequently led to a forced shutdown of 2-CH-P-1A (A Charging Pump, Unit 2) from
high vibration due to impending bearing failure. The inspectors reviewed IMCs 0612
and 0609 and determined that the finding was of very low safety significance given the
availability of the other charging pumps. The licensee has this finding documented in
their corrective action program as Plant Issue N-2006-0154.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
W. Anthes, Assistant Manager, Maintenance
G. Bischof, Director, Nuclear Safety and Licensing
J. Breeden, Supervisor, Radioactive Analysis and Material Control
W. Corbin, Director, Nuclear Engineering
J. Crossman, Assistant Manager, Nuclear Operations
J. Davis, Site Vice President
E. Dryer, Health Physicist
J. Eastwood, Corporate ISI Coordinator
R. Evans, Manager, Radiological Protection
R. Foster, Supply Chain Manager
E. Holloway, ISI
S. Hughes, Manager, Nuclear Operations
P. Kemp, Supervisor, Nuclear Safety & Licensing
J. Kirkpatrick, Manager, Maintenance
S. Kotowski, Engineering Supervisor
A. Kozak, Simulator Support Operations
L. Lane, Director, Operations and Maintenance
M. Lane, Supervisor Health Physics Operations
J. Leberstien, Licensing Technical Advisor
T. Maddy, Manager, Nuclear Protection Services
M. Main, Component Engineer
T. Mayer, Corporate Eddy Current Level III Examiner
C. McClain, Manager, Organizational Effectiveness
F. Mladen, Manager, Nuclear Site Services
B. Morrison, Assistant Engineering Manager
J. Rayman, Emergency Planning Supervisor
H. Royal, Manager, Nuclear Training
G. Salomone, Licensing
M. Sartain, Manager, Nuclear Engineering
J. Scott, Supervisor, Nuclear Training (operations)
W. Shura, Nuclear Training Supervisor
R. Wesley, Supervisor Shift Operations
M. Whalen, Licensing
R. Williams, Component Engineer
NRC personnel
K. Landis, Chief, Branch 5, Division of Reactor Projects, Region II
Attachment
A-2
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Closed
05000338, 339/2005004-02 URI Inadequate Corrective Action Results in
Safeguards Instrument Rack Room Flood Problem
(Section 4OA5)
Opened and Closed
05000338, 339/2006002-01 NCV Failure to translate TS operable-operability
definition regarding hazard barriers into instructions
as required by 10 CFR 50 Appendix B Criterion III
(Section 1R13)
05000338, 339/2006002-02 NCV Inadequate Design Control Results in Safeguards
Instrument Rack Room Flood Problem (Section
4OA5)
Attachment
A-3
LIST OF DOCUMENTS REVIEWED
Section 1R05: Fire Protection
Plant Issues
- N-2006-1588, NRC identified problem with oil and solvent staged in containment
beyond the specified amount identified on the transient combustible permit
Section 1R08: Inservice Inspection (ISI) Activities
Documents for Nondestructive Examination
- NDE-UT-812, Ultrasonic Examination of Austenitic Piping Welds in Accordance
with ASME Section XI, Appendix VIII, Revision 0
- NDE-UT-811, Ultrasonic Examination of Ferritic Piping Welds in Accordance with
ASME Section XI, Appendix VIII, Revision 0
- NDE-UT-705, Manual Ultrasonic Examination of Reactor Coolant Piping Welds,
Revision 3
- NDE-MT-701, Magnetic Particle Examination, Revision 4
- NASES-6.23, Boric Acid Corrosion Control Program (BACCP), Revision 2
- DNAP-1004, Boric Acid Corrosion Control Program (BACCP), Revision 3
- 1-PT-48.5, Leakage Inspection Above Reactor Vessel Head, Revision 1
- 1-PT-48.4, Bare Metal Inspection of Vessel BMI Nozzles, Revision 1
- 1-PT-48.3, Visual Inspection Borated Systems in Containment, Revision 1
1-PT-48.1, Visual Inspection of ASME XI Class 2 Pressure Boundary
Components Inside Reactor Containment, Revision 2
- 1-PT-48, Visual Inspection of Reactor Coolant Pressure Boundary Components,
Revision 13
- 1-PT-46.21, RCS Pressure Boundary Components Affected by Boric Acid
Accumulation, Revision 15
- Boric Acid Corrosion Control Program Health Report, 2005-Q4
- Areva Document No. 54-ISI-400, Multi-Frequency Eddy Current Examination of
Tubing, Revision 14
- Areva Document No. 1275114, Eddy Current Data Management Guidelines,
Revision 07
- Station Administrative Procedure VPAP-1302, Foreign Material Exclusion
Program, Revision 21
- Maintenance Vendor Procedure 03-6033813A, Field Procedure for Removal and
Installation of Primary Steam Generator Manway Insert for Dominion Generators,
Revision 2
- Areva Document Identifier 51-9014652, North Anna Unit 1 1R18 - EPRI
Appendix H Eddy Current Technique Review
- Work Order 733421-01 to 07, B S/G Steam Dome Inspection
- Confined Space Evaluation and Entry Permit - Hot and Cold Legs
- Foreign Material Control Log
- Eddy Current Analyses Calibration No. 11 for tubes R18C16, R16C16, and
R16C17; 47 for tube R29C51; and 48 for tubes R24C58 and R03C58
Attachment
A-4
- Data Aquisition and Analysis Personnel Qualification for Level II Data Operators,
Level II A and Level III A Analysts
Plant Issues
- N-2005-5615, Initiate evaluation of the effectiveness of the boric acid corrosion
control program
- N-2006-1030, Listing of boric acid leaks identified during 1-PT-46.21 walkdown
- N-2006-1082, Boron crystals on Unit 1 reactor head thermocouples
- N-2006-1090, Valves discovered with boric acid during 1-PT-48 walkdown
- N-2006-1204, High number of boric acid leaks identified
- N-2005-5399, Boric acid leakage through canopy seal weld on SI sampling valve
- N-2006-1357, Deviation in the standard for the signal amplitude between
standards ADVB-013-96 and ADVB-016-96
Section 1R11: Licensed Operator Requalification - Biennial Review
Documents
- Functional Implementation Guideline (FIG) -07 Implementing LORP Instructional
/Evaluation Components, Revision 5
- FIG - 07, Program Administration and Documentation, Revision 11
- FIG - 09, Administering the LORP Examination Banks (test items and task
performance evaluations), Revision 16
- DNAP-0509, Dominion Nuclear Procedure Adherence and Usage Revision 4
- TRCP-3002, Simulator Modification Record Process (SMR), Revision 9
- TRCP-3006, Simulator Configuration Management, Revision 6
- TRCP 3007, Simulator Performance Testing, Revision 1
- Executive Summary-Simulator Performance Test Procedure, January 2006
- NAS-RC-01 System Test Procedure, Revision 0
- Memo Dated January 13, Simulator Review Board Minutes
- Various Simulator Modification Requests (SMR)
- Scenario SXG #44, Revision 3
- Scenario SXG #31, Revision 4
- JPM - N534 Align the Service Water System to the Containment Air
Recirculation Fans on the backboards watchstation. (10/12/05)
- JPM - R514 Restore Residual Heat Removal Flow (1-AP-11) (10/12/05)
- JPM - N1473/14045 Rack in a 4160 Volt Breaker (10/12/05)
- JPM - N10 Isolate Reactor Coolant Pump Seals Locally (10/12/05)
- JPM - N1585 Align the turbine-driven auxiliary feedwater pump to feed the steam
generators by way of the hand control valve header (10/12/05)
- Badge Access Transaction Reports for Reactivation of Licenses (3)
- Licensed Operator Medical Records (20)
- Feedback Summaries
Attachment
A-5
Section 1R12: Maintenance Effectiveness
Plant Issues
- N-2005-1050, identification by engineering that the process used in the
manufacture of EXIDE type 3CC-7 batteries may result in premature degradation
of the battery
- N-2005-1165,internal resistance reading obtained on the 1J EDG battery indicate
the cell may be degraded
- N-2006-0357, maintenance rule criteria exceeded for SBO diesel generator
- N-2006-0387, internal resistance readings obtained for the 1JEDG batteries
were higher than those obtained previously
Section 1R13: Maintenance Risk Assessments and Emergent Work Evaluation
Plant Issues
- N-2006-1174, NRC identified problem with a failure to assess risk for
2-CC-TV-204B; subsequent calculations resulted in a Green risk condition.
Section 1R20: Refueling and Other Outages
Documents
- Procedure 1-GOP-13.0, Alternate Core Cooling Method Assessment
- Dominion Memorandum dated 3/6/06, 2006 Outage Plan Safety Review North
Anna Unit 1
- VPAP-2805, Shutdown Risk Program
- 1-PT-93, Reactor Vessel Water Level Determination
- 1-OP-5.4, Draining the Reactor Coolant System
NRC identified Plant Issues from Containment Closeout Inspection
- N-2006-2076, snubber 1-RC-HSS-838 was bound in place by a support on one
side and insulation on the other side
- N-2006-2077, dry boric acid around packing gland for 1-RC-HCV-1556A
- N-2006-2079, insulation issues were identified on several components: 1-BD-1,
1-HC-343, 1-CH-366, 1-RC-199, 1-CH-365, 1-CC-716, 1-CC-944, 1-CC-45,
sample system line in mechanical penetration area, 1-RH-E-1B, A SG
blowdown line on RHR flat
- N-2006-2080, boric acid identified on packing gland for 1-CH-TV-1204A
- N-2006-2081, BACC flag found on 1-IA-38
- N-2006-2083, miscellaneous material and debris was identified and removed
from containment including: several large screwdrivers, pieces of tape &
tie-wraps, 5 large washers, piece of cloth, loose insulation, cup of paint chips
- N-2006-2084, metal filings and dirt was identified on the 4160V electrical lead
insulators in the containment penetration area for 1-RC-P-1B, 1-RC-P-1C, and
1-RC-P-1A
- N-2006-2085, 2 openings in the screen installed to prevent paint on containment
air recirc fan duct from reaching sump was identified near 1-HV-F-1A
- N-2006-2087, the inspector informed the licensee that the UFSAR required
encapsulation of all insulation, yet licensees procedure, NAI-029, Specification
for Installation of Thermal Insulation, allows for insulation in SG cubicles to be
left unjacketed if certain criteria are met
Attachment
A-6
Section 2OS1: Access Controls to Radiologically Significant Areas
Procedures, Manuals, and Guidance Documents
- Health Physics Administrative Procedure HPAP-1081, Radiation Work Permit
Program, Revision 4
- C-HP-1032.061, High Radiation Area Key Control, Revision 3
- C-HP-1032.060, Radiological Posting and Access Control, Revision 1
- C-HP-1061.020, Personnel Contamination Monitoring and Decontamination,
Revision 8
- C-HP-1081.010, Radiation Work Permits: Preparing and Approving, Revision 7
- C-HP-1081.020, Radiation Work Permits: RWP Briefing and Controlling Work,
Revision 5
- C-HP-1081.040, Radiation Work Permits: Providing HP Coverage During Work,
Revision 1
- HP-1071.020, Controlling Contaminated Material, Revision 6
- Standing Order #125, Compensatory measures implemented for the use of
RMS, 8/15/05
Radiation Work Permits
- 05-2-1102, Load, transport, and store spent fuel dry storage casks
- 06-2-3503, Replace transfer canal blank flange and transfer cart inspection
- 06-2-3507, Vacuum debris from cavity/transfer canal during refueling operations,
perform decon of transfer canal
- 06-2-3510, Perform eddy current testing and tube plugging of S/G
- 06-2-3230, Walkdowns, inspections, and observations
- 06-2-3220, Perform ultrasonic, magnetic particle, and liquid penetrant
examination for inservice inspection program
- 06-2-1502, Containment entry during subatmospheric conditions
- 06-2-1210, Spent resin transfer
Surveys, Data, Records
- Plant Status Board Radiological Surveys (3/13/06-3/17/06, 3/27/06-3/29/06)
- High Radiation Key Locker Log, printed 3/8/06
- Whole Body Count and Inhalation Intake Data Sheets, TLD 3020, 10/14/05
- Special Dosimetry Issue Worksheets, TLDs 6021 and 7625, 3/20/06
Audits and Self-Assessments
- ITC-SA-04-02, Assessment of Nuclear Business Unit for Adverse Trends in
Radiological Protection Events, 4/29/04
- Nuclear Oversight Audit Report 05-06, Radiological Protection, Process Control
Program, and Chemistry Program at Millstone, North Anna, and Surry,
05/09/2005 - 05/27/2005
Attachment
A-7
CAP Documents (Plant Issues)
- N-2005-3958, Worker received dose rate alarm while working under Unit 2
reactor vessel, 10/3/05
- N-2005-4038, Lock for B motor cube door on 262' of Unit 2 containment is
broken. Currently this door is being controlled as a locked high radiation area.,
10/5/05
- N-2005-4099, Supplemental information to document events propagated from a
Level 1 PCE; 8,000 dpm in facial area, 10/6/05
- N-2005-4389, Individual received a DAD dose rate alarm while working on
scaffold platform in B cube 241', 10/13/05
- N-2005-4657, Worker received dose rate alarm while performing
deconatmination activities in Unit 2 reactor cavity, 10/21/05
- N-2006-1001, Decon Building basement high radiation gate locks have been
found to be defective 8 times in the last 24 months, 3/9/06
- N-2006-1336, An HP technician received a dose alarm while performing surveys
in Unit 1 reactor containment, 3/18/05
- N-2006-1378, Individual received dose rate alarm while performing observations
on the scaffold platform in C cube 241', 3/19/06
- N-2006-1462, Individual received dose rate alarm while performing valve
maintenance activities on scaffold platform in C cube 241', 3/21/06
- N-2006-1544, Worker performing inspections in the secondary side of Unit 1 B
S/G received a dose alarm, 3/22/06
Attachment
A-8
- 06-023, Perform Health Physics zone coverage, Surveys, and walkdowns during
the Unit 1 2006 Refueling outage.
and inspection of AOV's during Unit 1 2006 Refueling outage.
refueling outage, 3/27/06
Includes A Seals, and annual PM's, 3/27/06
2006 refueling outage, 3/28/06
Section 2PS2: Radioactive Material Processing and Transportation
Procedures, Guidance Documents, and Manuals
- HP-1071.021, Storing Radioactive Material Outside the Protected Area,
Revision 14
- HP-1071.030, Receiving Radioactive Material, Revision 11
- HP-1072.010, Packaging Radioactive Waste, Revision 9
- HP-1072.020, Sampling, Analyzing, and Classifying Solid Radioactive Waste,
Revision 7
- 0-HSP-INST-002, Health Physics Assessment of Radioactive Waste Stream
Changes, Revision 0
- C-HP-1071.040, Packaging and Shipment of Radioactive Material, Revision 1
Attachment
A-9
- C-HP-1072.040, Radioactive Waste Disposal Using the Barnwell Disposal
Facility, Revision 3
- C-HP-1072.050, Radioactive Waste Transfer to Licensed Waste Processors,
Revision 1
- VPAP-2104, Radioactive Waste Process Control Program, Revision 5
Records and Data
- 2004 and 2005 Radioactive Waste Shipment Summary Forms
- Radioactive Material Shipment Log, 2005 and 2006 (Year-to-Date)
- HAZMAT/Transportation Safety Qualification List, Undated
- Annual Radioactive Effluent Release Report, North Anna Power Station (January
01, 2004 to December 31, 2004)
- Flow Valve Operating Numbers Diagram (FVOND), Waste Disposal System,
North Anna Power Station - Unit 1 (NAPSU1), Drawing No. 11715-FM-87A,
Sheet 1, Revisions 26 and 27
- FVOND, Waste Disposal System, NAPSU1, Drawing No. 11715-FM-87A, Sheet
2, Revision 30
- FVOND, Waste Disposal System, NAPSU1, Drawing No. 11715-FM-87A, Sheet
3, Revision 33
- FVOND, Waste Disposal System, NAPSU1, Drawing No. 11715-FM-087D,
Sheet3, Revision 33
- New Waste Stream Data Reports, Dated 11/16/2002, 04/25/2003, 12/08/2004,
02/27/2005, 08/05/2005
- Radwaste Shipping Package 05-DUR-05, 8/31/05
- Radwaste Shipping Package 05-DUR-06, 12/16/05
- Radwaste Shipping Package 05-CNS-02, 6/7/05
- SCO Shipping Packages 05-1032 (8/17/05), 05-1046 (10/6/05), and 06-1021
(3/19/06)
Audits and Self-Assessments
- Nuclear Oversight Audit Report 04-08: Radiation Protection & Process Control
Program, 05/17/2004 - 05/28/2004
- Nuclear Oversight Audit Report 05-06: Radiation Protection/Process
Control/Chemistry Programs, 05/09/2005 - 05/27/2005
CAP Documents (Plant Issues)
- N-2004-1182, Repetitive failures are occurring on the discharge line for
2-DA-P-7B, Unit 2 inside Mat Sump pump, 04/15/2004
- N-2004-1788, Spent OLCMS secondary resin (which was counted in two
separate marinelli containers) was released for disposal to clean trash,
05/16/2004
- N-2005-2366, An upward trend in dose rates around the Fluid Waste Treatment
Tank has been noticed by HP during routine radiological surveys.
- N-2005-2586, contrary to the requirements of 49 CFR Subpart H Code of
Federal Regulations, Transportation, a Facilities and Support personnel has
been transporting hazardous material (gasoline and diesel fuel) offsite without
completing employer HAZMAT training.
Attachment
A-10
- N-2005-3345, Upon completion of the final pump down of the Fluid Waste
Treatment Tank, dose rates in the area increased to 300-500 mr/hr around the
tank.
Section 4OA1: Performance Indicator Verification
Procedures, Manuals, and Guidance Documents
- VPAP 1501, Deviations, Revision 17
Records and Data
- EPD Dose/Dose Rate Alarms: July 2005 - March 2006
- NAPS NRC Performance Indicator Data: January 2005 - January 2006
- North Anna Power Station Units 1 and 2 and ISFSI Annual Radioactive Effluent
Release Report for Calendar Year 2004, dated 4/14/05
- Quarterly printouts for Liquid and Gaseous Effluent doses to members of the
public, January - December 2005
CAP Documents (Plant Issues)
- Review of all Plant Issues from July 1, 2005 to March 18, 2006 with search terms
DAD Dose Rate Alarm, Dose Rate Alarm, Dose Rate, DAD, and
dosimeter. Detailed review and discussion with licensee of Plant Issues listed
in Section 2OS1 of the Attachment.
Section 4OA5: Other Activities
ISFSI Radiological Controls
Procedures
- HP-1020.012, Radiological Protection Action Plan During Dry Storage Cask
Activities, Revision 014
- 0-Health Physics Surveillance (HPS)-ISFSI-001, Independent Spent Fuel
Storage Installation (ISFSI) Health Physics TLD Survey Surveillance, Revision 3
Surveys, Data, and Records
- ALARA Package 05-003, Load, transport, and store spent fuel dry storage cask
- Cask TN-32 #47 load, transport, and placement surveys (4/5/05-4/14/05)
- ISFSI Pad surveys (2/17/04, 5/21/04, 8/22/04, 11/19/04, 8/24/05 11/23/05,
2/21/06)
- Cask TN-32 #42 Grid Survey, 2/7/04
- Cask TN-32 #43 Grid Survey, 3/13/04
- Cask TN-32 #45 Grid Survey, 6/21/04
2005)
CAP Documents (Plant Issues)
- N-2004-2910, Some discrepancies between TLD readouts and calculated
neutron exposures were noted on the 1st quarter 2004 TLDs, 8/4/04
- N-2004-4993,While performing ISFSI perimeter fence survey as part of the
quarterly Controlled Area Survey, noted that a jersey barrier was located directly
between the northernmost ISFSI TLD and the SF casks, 11/19/04
- N-2005-1326, During spent fuel cask activities, 4 bubble dosimeters used to
estimate neutron exposure, did not respond as expected, 4/6/05
Attachment
A-11
LIST OF ACRONYMS
ALARA As Low As Reasonably Achievable
ASME American Society of Mechanical Engineers
BACCP Boric Acid Corrosion Control Program
CAP corrective action program
CC component cooling
DOT Department of Transportation
ED electronic dosimeter
EDG emergency diesel generator
EPRI Electric Power Research Institute
HPT health physics technician
Hx heat exchanger
IMC inspection manual chapter
ISFSI Independent Spent Fuel Storage Installation
ISI inservice inspection
LHRA locked high radiation area
LHSI Low Head Safety Injection
LOCA Loss of Coolant Accident
NCV non-cited violation
NDE nondestructive examination
PREACS Emergency Core Cooling System Pump Room Exhaust Air Clean-up System
RAB reactor auxiliary building
RCA radiologically controlled area
RCB reactor containment building
RCP reactor coolant pump
RFO Refueling Outage
RSST Reserve Station Service Transformer
RWP Radiation Work Permit
SDP Significance Determination Process
S/G steam generator
SSC structures, systems, and components
SSPS solid state protection system
TS Technical Specifications
UFSAR Updated Final Safety Analysis Report
VHRA very high radiation area
VPAP Virginia Power Administrative Procedure
WO Work Order
Attachment