ML053050395
| ML053050395 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/28/2005 |
| From: | Landis K NRC/RGN-II/DRP/RPB5 |
| To: | Christian D Virginia Electric & Power Co (VEPCO) |
| References | |
| IR-05-004 | |
| Download: ML053050395 (51) | |
See also: IR 05000338/2005004
Text
October 28, 2005
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. 05000338/2005004 AND 05000339/2005004
Dear Mr. Christian:
On September 30, 2005, the United States Nuclear Regulatory Commission (NRC) completed
an inspection at your North Anna Power Station, Units 1 and 2. The enclosed integrated
inspection report documents the inspection findings, which were discussed on September 22,
2005, 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, six self-revealing findings of very low safety
significance (Green) were identified. Five of these were determined to involve violations of
NRC requirements. However, because of their very low safety significance and because they
were entered into your corrective action program, the NRC is treating these five findings as
non-cited violations (NCVs) consistent with Section VI.A of the NRC Enforcement Policy. A
self-revealing violation whose significance determination is to be determined was also identified.
In addition, one licensee-identified violation, which was determined to be of very low safety
significance (Green), 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
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
Enclosures:
Inspection Reports 05000338/2005004 and 05000339/2005004
cc w/encls.:
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
_________________________
OFFICE
RII:DRP
RII:DRP
RII:DRS
RII:DRS
RII:DRS
RII:DRS
RII:DRS
SIGNATURE
JTR
GJW
JHW2 for
JHW2 for
JHW2 for
NAME
JReece
GWilson
WLoo
RHamilton
ANielsen
FWright
LMiller
DATE
10/28/2005
10/28/2005
10/28/2005
10/28/2005
10/28/2005
10/28/2005
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
OFFICE
RII:DRS
RII:DRS
RII:DRP
SIGNATURE
MXM3 for
MXM3
LXG1
NAME
MScott
MMaymi
LGarner
DATE
10/28/2005
10/28/2005
10/28/2005
E-MAIL COPY?
YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO YES
NO
Enclosure
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-338, 50-339
Report Nos.:
05000338/2005004, 05000339/2005004
Licensee:
Virginia Electric and Power Company (VEPCO)
Facilities:
North Anna Power Station, Units 1 & 2
Location:
1022 Haley Drive
Mineral, Virginia 23117
Dates:
July 1, 2005 - September 30, 2005
Inspectors:
J. Reece, Senior Resident Inspector
G. Wilson, Resident Inspector
W. Loo, Senior Health Physicist, Sections 2PS1, 4OA5
R. Hamilton, CHP Senior Health Physicist, Sections 2OS1, 4OA1, 4OA5
A. Nielsen, CHP Health Physicist, Section 2OS3
F. Wright, Senior Health Physicist, Section 2PS3
L. Miller, Senior Emergency Preparedness Inspector, Sections 1EP2-1EP5, and
4AO1
M. Scott, Senior Reactor Inspector, Section 1R12
M. Maymi, Reactor Inspector, Section 1R12
Approved by: K. Landis, Chief, Reactor Projects Branch 5
Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000338/2005-004, IR 05000339/2005-004; 07/01/2005 - 09/30/2005; North Anna Power
Station Units 1 & 2. Routine Integrated Resident and Regional Report. Maintenance
Effectiveness - Biennial Assessment. Emergency Preparedness Baseline. Radiation Safety.
The report covered a three-month period of inspection by the resident inspectors, health
physicists, a senior emergency preparedness inspector, and reactor inspectors from the region.
Six self-revealing Findings were identified. Five of these were determined to be Non-cited
Violations (NCVs). A self-revealing violation whose significance determination is to be
determined was also identified. 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: Initiating Events
Green. A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI,
was identified regarding a failure to promptly identify and correct deficiencies which
caused anomalies in the Unit 2 channel 1 over-temperature delta-temperature (OTDT)
instrumentation. The anormalies occurred during a lightning storm on July 29, 2003 and
the licensee took no corrective actions to correct the condition. As a result, it was not
until a Unit 2 automatic reactor trip from an OTDT signal on August 5, 2005, during a
lightning storm, that the licensee identified an installation deficiency associated with a
1989 modification. A similar Unit 2 automatic reactor trip from an OTDT signal occurred
during a lightning storm on September 17, 1998.
The finding had an impact on safety based on the deficiencies resulting in two reactor
trips and a third documented near miss event. The finding was more than minor
because it affected the Initiating Events cornerstone objective to limit the likelihood of
those events that upset plant stability and the cornerstone attribute of design control.
The finding is of very low safety significance because it did not contribute to the
likelihood of a primary or secondary system loss of coolant accident, a loss of mitigation
equipment functions or the likelihood of a fire or flood event. This finding contains
aspects relating to the cross-cutting area of problem identification and resolution.
(Section 1R14.1)
Green. A self-revealing finding was identified for untimely corrective action resulting in a
rapid reduction of power on Unit 1 due to a severe oil leak on the valve actuator for
1-EH-TV-100, main turbine auto stop oil interface valve. A similar problem on this valve
resulted in a manual reactor trip on April 19, 2003. Subsequent evaluations from a Unit
2 similar issue determined that torque values as specified by procedure for the valve
actuator diaphragm bolts were below the values as recommended by the vendor, but
untimely corrective actions resulted in a rapid Unit 1 down-power on August 5, 2005.
2
Enclosure
This finding had a credible impact on safety due to the challenge of plant control
systems from the rapid reduction of power. The finding is consequently more than
minor based on the impact to the Initiating Events cornerstone objective to limit the
likelihood of those events that upset plant stability and the cornerstone attribute of
equipment reliability. This finding contains aspects relating to the cross-cutting area of
problem identification and resolution. (Section 1R14.2)
Green. On July 22, 2005, a self-revealing non-cited violation of Technical Specification 5.4.1.a was identified for a failure to follow a surveillance procedure which resulted in
placing an incorrect bistable in a trip condition on Unit 2. Only unexpected control room
alarms occurred as a result of the performance deficiency since no other logic channels
bistables were in trip.
The inspectors determined that the finding is more than minor because it could
reasonably be viewed as a precursor to a more significant event. If another channel in
the logic had already been tripped, the plant would have been adversely affected. The
finding is of very low safety significance (Green) because it did not involve any loss of
coolant accident initiators, did not contribute to both a reactor trip or mitigating system
unavailability, nor increase the likelihood of a fire. This finding contains aspects relating
to the cross-cutting area of human performance. (Section 1R22.1)
Cornerstone: Mitigating Systems
Green. A self-revealing non-cited violation of Technical Specification 5.4.1.a was
identified for an inadequate procedure which resulted in the loss of two Unit 1 safety-
related 480V buses on May 1, 2005.
The finding had a credible impact on safety due to the loss of two safety-related 480V
buses resulting in the loss of power to multiple B train components two minutes after a
containment depressurization signal during a design basis accident. The finding is more
than minor due to the impact on two cornerstones, Mitigating Systems and Barrier
Integrity. A Phase II evaluation of the significance determination process concluded the
finding was of very low safety significance (Green) because only the B train was
affected, a two minute time delay allowed safety-related component reposition, and
emergency procedures identified appropriate operation action for manual component
operation following the fault. This finding contains aspects relating to the cross-cutting
area of human performance. (Section 1R12)
Green. A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion III, was
identified for inadequate design controls. During the development of a service water
(SW) expansion joint modification, which was implemented in December 2003, the
licensee failed to verify the design adequacy of adjacent pipe support and restraints.
The design failed to incorporate normal system pressure loads in the design. As a
result, on June 14, 2005, during inspections of the SW expansion joints, the licensee
noted severe damage on adjacent pipe support and restraints. Both the Unit 1 and Unit
2 A and B trains of SW were affected. The SW system was determined to operable
but degraded.
3
Enclosure
This finding had a credible impact on safety based on a design control error which
impacted both trains of the SW system which is a link between the transfer of reactor
decay heat to the plants ultimate heat sink. The finding is more than minor due to the
impact on the Mitigating Systems cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences (i.e. core damage) and the cornerstone attribute of design
control of plant modifications. The finding is of very low safety significance because the
design deficiency was confirmed not to result in loss of function per Generic Letter 91-
18. This finding contains aspects relating to the cross-cutting area of human
performance. (Section 1R04.2)
TBD. A self-revealing violation of 10 CFR 50, Appendix B, Criterion XVI was identified
for inadequate corrective action resulting in a flood potential for the Unit 1 and 2
safeguards instrument rack rooms. Corrective actions in October 2004, associated with
water from a capped floor drain outside the air conditioning chiller room (ACCR) failed to
identify that back-flow preventers where not installed in the floor drains between the
ACCR and the air conditioning fan room (ACFR). As a result, the lack of floor drain
back-flow preventers was not discovered until July 9, 2005, when water was
unexpectedly transferred between with the ACCR and ACFR. 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 of 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. The finding involves a Phase III evaluation for the significance
determination process 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 finding
is unresolved pending completion of the significant determination assessment and
involves aspects of the cross-cutting area of problem identification and resolution.
(Section 1R06)
Cornerstone: Barrier Integrity
Green. A self-revealing non-cited violation of Technical Specification 5.4.1.a was
identified for a failure to follow a maintenance procedure. On February 19, 2005, the
Unit 2 B quench spray pump motor breaker overload setpoints were not set in
accordance with procedures. As a result, the pump tripped while starting on August 19,
2005.
The finding had a credible impact on safety due to the starting failure of one of the
components required to reduce containment pressure following a design basis accident.
The finding was more than minor because it affected the Barrier Integrity cornerstone
objective to provide reasonable assurance that the containment physical design barriers
protect the public from radio nuclide releases caused by accidents or events, and the
respective cornerstone of human performance. The finding was determined to be of
4
Enclosure
very low safety significance because it did not impact design deficiencies, result in a loss
of system safety functions, exceed related TS outage times, nor involved a seismic,
flooding, or severe weather initiating event. This finding contains aspects relating to the
cross-cutting area of human performance. (Section 1R22.2)
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
CONTENTS
Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R04
Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R05
Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R06
Flood Protection Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1R11
Licensed Operator Requalification Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R12
Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R13
Maintenance Risk Assessments and Emergent Work Control . . . . . . . . . . . . . . . . . . . 9
1R14
Operator Performance During Non-Routine Evolutions and Events . . . . . . . . . . . . . . . 9
1R15
Operability Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12
1R17
Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .13
1R19
Post-Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1R22
Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
1R23
Temporary Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
1EP2 Alert and Notification System Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
1EP3 Emergency Response Organizational Augmentation . . . . . . . . . . . . . . . . . . . . . . . . . 17
1EP4 Emergency Action Level and Emergency Plan Changes . . . . . . . . . . . . . . . . . . . . . . 18
1EP5 Correction of Emergency Preparedness Weakness and Deficiencies . . . . . . . . . . . . 18
1EP6 Drill Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
2OS1 Access Control to Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
2OS3 Radiation Monitoring Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems . . . . . . 22
2PS3 Radiological Environmental Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
4OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
4OA4 Cross-cutting Aspects of Findings. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .27
4OA5 Other Activities
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
4OA6 Meetings, Including Exit
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
4OA7 Licensee-Identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
ATTACHMENT: SUPPLEMENTARY INFORMATION
Key Points of Contact . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
A-1
List of Items Opened, Closed, and Discussed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
A-1
List of Documents Reviewed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
A-2
Enclosure
REPORT DETAILS
Summary of Plant Status
Unit 1 and Unit 2 began the inspection period at 100 percent power, and remained at or near
100 percent power for the entire reporting period except for minor power reductions to perform
required periodic testing and the following events:
Unit 1 experienced a rapid down-power event on August 5, 2005, due to severe
oil leakage on 1-EH-TV-100, and
Unit 2 experienced an over-temperature delta-temperature (OTDT) automatic reactor trip during a lightning storm on August 5, 2005.
3.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R04
Equipment Alignment
.1
Partial System Walkdowns
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 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 1 1H Emergency Diesel Generator (EDG) during planned maintenance on
the 1J EDG;
Unit 2 Auxiliary Feedwater 2-FW-3A, during planned maintenance on 2-FW-3B;
and,
Unit 2 Quench Spray 2-QS-P-1A during emergent work on 2-QS-P-1B.
b.
Findings
No findings of significance were identified.
.2
Complete System Walkdown
a.
Inspection Scope
The inspectors performed a detailed walkdown and inspection of the Unit 2 Service
Water (SW) system outside of containment to assess properly alignment and to identify
discrepancies that could impact its availability and functional capacity. The inspectors
2
Enclosure
assessed the physical condition of the pumps, valves, pipe supports, and
instrumentation. The inspection also included a review of the alignment and the
condition of support systems including fire protection, room ventilation and emergency
lighting. Equipment deficiency tags were reviewed and the condition of the system was
discussed with engineering personnel. The operating procedures, drawings and other
documents utilized and reviewed as part of the inspection are listed in the Attachment.
b.
Findings
Inadequate Design Control Results in Degradation of SW Support/Restraints
Introduction. The inspectors identified a self-revealing non-cited violation (NCV)
associated with inadequate design control resulting in degradation of SW system
support-restraints (S/R).
Description. On June 14, 2005, the licensee was performing a preventative maintenance
(PM) inspection of SW expansion joints on plant discharge piping located in the SW tie-in
vault and noticed severely bent or degraded SW S/Rs, 1-SW-PH-3.2 on the B train
discharge header and 1-SW-PH-4.2 on the A train discharge header. Both trains are
shared between Units 1 and 2. An extent of condition walkdown was performed and one
additional S/R, 1-SW-PH-E85.2, was identified with structural damage and documented
in Plant Issue N-2005-2225. The inspectors reviewed and verified the resultant
functional evaluation which concluded that a generic letter (GL) 91-18 (operable but
degraded) condition existed for A and B SW trains. The licensee performed a root cause
evaluation which determined that a design analysis failure occurred during a modification
(Design Change 02-006) which was implemented in December, 2003, and converted the
metal expansion joints to a design using rubber as the flexible component. The
additional pressure component of piping loads associated with the new rubber design
was not translated by engineering personnel to the modification process resulting in
S/Rs too weak to handle the loads associated with normal system operation. The
human performance aspects of the design failure analysis is a noncompliance with 10 CFR 50, Appendix B, Criterion III, which states in part that measures shall provide for
verifying or checking the adequacy of design.
Analysis. This finding had a credible impact on safety based on a design control error
which impacted both trains of the SW system which is a link between transfer of reactor
decay heat to the plants ultimate heat sink. The inspectors reviewed Inspection Manual
Chapter (IMC) 0612 and determined the finding is more than minor due to the impact on
the Mitigating Systems cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable
consequences (i.e., core damage) and the cornerstone attribute of design control of plant
modifications. The inspectors referenced IMC 0609 for the Significant Determination
Process (SDP) and determined that the finding is Green or very low safety significance
because the design deficiency was confirmed not to result in loss of function per GL 91-18. This finding contains aspects relating to the cross-cutting area of human
performance.
3
Enclosure
Enforcement. 10 CFR 50, Appendix B, Criterion III, requires in part that measures shall
provide for verifying or checking the adequacy of design. Contrary to the above,
inadequate verification of a modification, Design Change 02-006, implemented in
December, 2003, to replace SW metal expansion joints with a rubber design resulted in
an operable but degraded condition due to damaged SW system S/Rs discovered on
June 14, 2005. This finding is of very low safety significance or Green, is in the
licensees corrective action program (CAP) as Plant Issue N-2005-2229, and is
characterized as a NCV, consistent with Section VI.A of the NRC's Enforcement Policy:
NCV 05000338, 339/2005004-01, Inadequate Design Control Resulting in Degraded
Service Water Support-Restraints.
1R05
Fire Protection
.1
Fire Drill
a.
Inspection Scope
During a fire protection drill on August 31, 2005, at the Service Water Pump House, the
inspectors assessed the timeliness of the fire brigade in arriving at the scene, the fire
fighting equipment brought to the scene, the donning of fire protective clothing, the
effectiveness of communications, and the exercise of command and control by the scene
leader. The inspectors also assessed the acceptance criteria for the drill objectives and
reviewed the licensees CAP for recent fire protection issues. Documents reviewed are
listed in the Attachment.
b.
Findings
No findings of significance were identified.
.2
Fire Area Tours
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. 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.
Auxiliary Building (includes Z-18 and Z-20) (fire zone 11a / AB);
Quench Spray Pump House and Safeguards Area Unit 2 (includes Z-16-2) (fire
zone 15-2a / QSPH-2);
Fuel Building (fire zone Z-18 / FB);
Main Control Room (fire zone 2a / CR);
4
Enclosure
Cable Vault and Tunnel Unit 2 (includes Control Rod Drive Room and Z-27-1)
(fire zones 3-2a / CV & T-2);
Cable Vault and Tunnel Unit 1 (includes Control Rod Drive Room and Z-27-1)
(fire zone 3-1a / CV & T-1);
Service Water Pump House (fire zone 12a / SWPH);
Safeguards Area Unit 2 (fire zone Z-16-2 / SA-2);
Safeguards Area Unit 1 (fire zone Z-16-1 / SA-1);
Casing Cooling Tank & Pump House Unit 1 (fire zone Z-41-1 / CCT & PH-1); and,
Casing Cooling Tank & Pump House Unit 2 (fire zone Z-41-2 / CCT&PH-2).
b.
Findings
No findings of significance were identified.
1R06
Flood Protection Measures
a.
Inspection Scope
The inspectors reviewed internal flood protection measures for the Unit 1 and 2 air
conditioning chiller rooms (ACCRs) and adjacent air conditioning fan rooms (ACFRs).
Flooding in the ACCRs and ACFRs could impact risk-significant components in the
instrument rack rooms adjacent to the ACFRs if flood mitigation features were degraded.
ACCR and ACFR protection features were observed to verify that they were installed and
maintained consistent with the plant design basis. The inspectors reviewed the
instrumentation and associated alarms for the rooms above to verify that the
instrumentation was periodically calibrated and that the respective alarms were
appropriately integrated into plant procedures. The inspectors also reviewed licensee
instructions in the event of severe flooding and evaluated the availability of systems,
structures and components (SSCs) for safe shutdown under worst case water levels.
Documents reviewed are listed in the Attachment.
b.
Findings
Inadequate Corrective Action Results in Safeguards Instrument Rack Room Flood
Problem
Introduction. The inspectors identified a self-revealing violation associated with
inadequate corrective action. Back-flow preventers were not installed in floor drains that
resulted in a flood potential for the Unit 1 and 2 Safeguards Instrument Rack Rooms.
The safety significance is under evaluation and thus the item is classified as an
unresolved item (URI).
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 ACCR.
During this work activity, the licensee observed water discharging from the floor drains in
the adjacent ACFR, and initiated Plant Issue N-2005-2565 to evaluate the absence of
5
Enclosure
back-flow preventers in the floor drains. The licensee initiated a flood watch, declared
the flood walls between the ACCR and adjacent ACFR on Units 1 and 2 inoperable, and
entered a Yellow 6 day maintenance rule risk condition based on the unavailability of the
flood walls to perform their function. The respective ACFR 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 2 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. A subsequent 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 ACCRs.
The calculation concluded that the peak flow rate from the Units 1 and 2 ACCRs to
adjacent ACFRs via the floor drain piping was 182.9 gpm and 169.4 gpm respectively.
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 that Plant Issue N-1999-3405, which documented operational experience
from Three Mile Island regarding check valves missing from floor drains and the impact
on flood protection, did not result in the identification and correction of this problem. The
inspectors concluded that the inadequate corrective actions for Plant Issue N-2004-4554
is contrary to the requirements of 10 CFR 50, Appendix B, Criterion XVI, which requires
that the establishment of measures to assure conditions adverse to quality are promptly
identified and corrected.
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 finding is an URI pending completion of the
significance determination assessment and contains aspects relating to the cross-cutting
area of problem identification and resolution.
Enforcement. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires the
establishment of measures to assure conditions adverse to quality are promptly and
identified and corrected. Contrary to the above, prompt identification and correction of
deficiencies relating to Plant Issue N-2004-4554 failed to identify and correct the absence
of back-flow preventers in the Unit 1 and 2 ACFRs. This violation is characterized as an
URI pending significance determination, and is identified as URI 05000338,
6
Enclosure
339/2005004-02, Inadequate Corrective Action Results in Safeguards Instrument Rack
Room Flood Problem. This finding is in the licensee's CAP as Plant Issue N-2005-2565.
1R11
Licensed Operator Requalification Program
a.
Inspection Scope
The inspectors observed an annual licensed operator requalification simulator
examination on September 13, 2005. The scenerio, Simulator Examination Guide
SXG-56, involved a loss of instrument air, followed by increased primary plant leakage, a
loss of bearing cooling pumps with subsequent reactor trip, and a small break loss of
cooling accident (LOCA).
The scenario required classifications and notifications that were counted for NRC
performance indicator input. 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
.1
Periodic Evaluation (Biennial)
a.
Inspection Scope
The inspectors reviewed the licensees Maintenance Rule periodic assessments, 2003
Maintenance Rule Periodic Assessment Report [NAPS-SA-03-03, dated 6/11/04] and
2005 Maintenance Rule Periodic Assessment Report [NAPS-SA-03-37, dated 8/15/05]
while on-site the week of August 15, 2005. These reports were issued to satisfy
paragraph (a)(3) of 10 CFR 50.65, and covered the 18 month periods ending August 31,
2003, and ending February 28, 2005, respectively, for Units 1 and 2. The inspection was
to determine the effectiveness of the assessment and that it was issued in accordance
with the time requirement of the Maintenance Rule (MR) and included evaluation of:
balancing reliability and unavailability, (a)(1) activities, (a)(2) activities, and use of
industry operating experience. To verify compliance with 10 CFR 50.65, the inspectors
reviewed selected MR activities covered by the assessment period for the following
maintenance rule component and attendant systems: Control Room Bottled Air, Control
Room Chilled Service Water Motors, High Head Safety Injection pump seals; Service
7
Enclosure
Water Spray Arrays, Reactor Water Storage Tank Chillers. Specific procedures and
documents reviewed are listed in the Attachment to this report.
During the inspection, the inspectors reviewed selected plant work order data,
assessments, modifications, the site guidance implementing procedures, discussed and
reviewed relevant corrective action [plant] issues, reviewed generic operations event
data, attendant MR related meeting minutes, probabilistic risk reports, and discussed
issues with system engineers. Operational event information was evaluated by the
inspectors in its use in MR functions. The inspectors selected work orders and other
corrective action documents on systems recently removed from 10 CFR 50.65 a(1)
status and those in a(2) status for some period to assess the justification for their status.
The inspectors toured and inspected repaired components. The documents were
compared to the sites MR program criteria, and the MR a(1) evaluations and rule related
data bases.
b.
Findings
No findings of significance were identified.
.2
Quarterly Sample
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
in-office 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. The inspectors also completed review of unresolved item
(URI) URI 05000338/20050003-01 which is documented in NRC Integrated Inspection
Report Nos. 05000338/2005003. Other documents reviewed are listed in Attachment.
The mechanical seals on pump 2-CH-P-1C were recently replaced with new seals
associated with Work Order (WO) 523899 for 20 ml/min outboard end bell leak on
2-CH-P-1C; and,
The Refueling Water Storage Tanks (RWST) mechanical chillers have had
multiple issues associated with the reliability of these chillers.
b.
Findings
(Closed) URI 05000338/20050003-01, Inadequate Maintenance of a Procedure Results
in Loss of Safety Related 480V Buses.
Introduction. A Green, self-revealing NCV was identified for failure to comply with TS 5.4.1 which resulted in the loss of two safety-related 480V buses on Unit 1.
8
Enclosure
Description. URI 05000338/2005003-01 documented a noncompliance with TS 5.4.1
which involved an inadequate maintenance procedure that resulted in the loss of two
safety-related 480 volt buses, 1J1-2N and 1J1-2S on May 1, 2005. The lack of adequate
instructions for breaker wiring resulted in a termination screw for the B phase field cable
connection to a thermal overload relay penetrating the adjacent insulation on the C
phase field cable connection. The resulting fault caused a flashover event within the
breaker cubicle and resulted in the upstream feeder breaker tripping on overcurrent with
the subsequent loss of the 480V buses.
Analysis. The inspectors referenced IMC 0612 and determined that the finding is more
than minor because it affected the reactor safety Mitigating Systems cornerstone
objective to ensure availability, reliability and capability of systems that respond to
initiating events to prevent core damage and the Barrier Integrity cornerstone objective to
provide reasonable assurance that physical design barriers such as containment protect
the public from radio nuclide releases caused by accidents or events. The attribute of
procedure quality was affected for each aforementioned cornerstone. The inspectors
referenced IMC 0609, for the SDP and determined that a Phase II analysis was required
because the finding affected two cornerstones. This analysis reviewed accidents
resulting in high containment pressure which would initiate a Containment
Depressurization Actuation (CDA) signal which, after a two minute time delay, would
close the affected breaker (to start a radiation monitor sample pump) resulting in the
fault. The analysis also reviewed the emergency procedures (EP) involving those
components which reposition prior to the fault due to the time delay as well as the
components which must be locally, manually controlled after the fault. Completion of the
applicable SDP worksheets of the Risk-Informed Inspection Notebook for North Anna
Power Station resulted in a risk of very low significance (Green) because only the B train
was affected, a two minute time delay allowed safety-related component reposition, and
emergency procedures identified appropriate operation action for manual component
operation following the fault. This finding contains aspects relating to the cross-cutting
area of human performance.
Enforcement. 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 (RG) 1.33, Revision 2, Appendix A, February 1978,
of which part 9.e. specifies general procedures for the control of maintenance work.
Contrary to the above, on December 28, 2004, maintenance procedure 0-EPM-0304-01
was not adequate, in that, it failed to provide sufficient instructions to preclude faulty
retermination of wiring in breaker 1-EE-BKR-1J1-2N-B5. This led to an electrical fault
and the loss of 1J1-2N and 1J1-2S MCCs on May 1, 2005. This violation is considered a
Non-cited Violation (NCV), consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000338/2005004-03, Inadequate Maintenance of a Procedure Results in Loss of
Safety Related 480V Buses. This issue is in the licensee's CAP as Plant Issue
N-2005-1615.
9
Enclosure
1R13
Maintenance Risk Assessments and Emergent Work Evaluation
a.
Inspection Scope
The inspectors evaluated, as appropriate, for the six activities 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.
Yellow maintenance rule 6-day window entered twice due to Control Room Chiller
area, Fan area, and SSPS Rack Room area flood concerns documented by Plant
Issue N-2005-2565;
Maintenance rule risk evaluation for unplanned Unit 1 down power with
concurrent Unit 2 reactor trip on August 8, 2005;
Maintenance rule risk evaluation for unplanned work on 2-QS-P-11B concurrent
with the components 2-CW-P-2A, 2-SW-MOV-221A, 2-HV-E-4A, 1-EE-BKR-
15J11 and 15D1 and 15D3, including rack work, switchyard and RSSTs;
Maintenance rule risk evaluation for planned restoration of A RSST to
underground line concurrent with the maintenance on 2-CW-P-2A, 2-SW-MOV-
221A, 1-EP-BKR-15A1, 1-FP-P-1, 2-CC-P-1B, 2-EP-BKR-25A1, SWYD, rack
work, B RSSTs on overhead lines, 0-EPM-1805-02, 0-PT-100.2, and 2-PT-44.7;
Maintenance rule risk evaluation for unplanned work on 2-EE-E6-2H concurrent
with the components 1-SW-P-4, 2-SW-MOV-221A, 1-CC-P-1A, 1-EE-BKR-15H12
and 2-MS-PCV-201A; and,
Maintenance rule risk evaluation for unplanned work on 2-EE-EG-2J, concurrent
with the components 2-CW-P-2A, 2-SW-MOV-221A, rack work, switchyard work,
and 1-PT-32.1.1.
b.
Findings
No findings of significance were identified.
1R14
Operator Performance During Non-Routine Evolutions and Events
a.
Inspection Scope
The inspectors reviewed operator logs and plant computer data for the two events listed
below to determine if plant and operator responses were in accordance with plant design,
procedures, and training. The inspectors also evaluated performance and equipment
problems to ensure that they were entered the licensees CAP.
10
Enclosure
The inspectors evaluated the response of the Unit 1 and 2 control room operators
on August 5 and 6, 2005, during an unplanned down power of Unit 1 for
diaphragm replacement on 1-EH-TV-100, and,
The inspectors evaluated the response of the Unit 2 control room operators on
August 5 and 6, 2005, following an automatic reactor trip which occurred during
the Unit 1 down power event above.
b.
Findings
.1
Inadequate Corrective Actions Results in a Reactor Trip
Introduction. A Green, self-revealing NCV was identified for a failure to identify and
correct deficiencies associated with reactor coolant instrumentation resulting in a reactor
trip.
Description. On August 5, 2005, a Unit 2 automatic reactor trip occurred due to actuation
of an OTDT reactor protection signal. A subsequent evaluation determined that a
lightning strike during a storm in progress at the time of the trip caused a transient in
Channel 1 and 2 reactor coolant temperature circuitry resulting in the automatic OTDT
trip signal. The inspectors verified that a reactor trip due to the same actuation signal
occurred during a lightning storm on September 17,1998. The subsequent root cause
evaluation concluded that the event was attributed to an external casual factor since it
was an event (lightning storm) outside the control of the company. Additionally, on July
29, 2003, during a lightning storm a transient was observed on Unit 2 channel 1 OTDT
instrumentation. However, a request for engineering assistance to investigate the
transient was not approved based on the conclusion of extremely dry soil conditions
affecting the grounding grid that year. Following the August 5, 2005, event the licensee
performed a more rigorous root cause evaluation and investigation that identified
ungrounded spare T-hot and T-cold narrow range resistance temperature detector (RTD)
shields that share the same thermowell in the reactor coolant system and same
containment electrical penetration as the active narrow range RTDs. Therefore, an
electrical transient induced by lightning in the spare, unshielded RTD elements was
consequently introduced into the active RTD elements, entered the narrow range
temperature reactor protection circuitry and resulted in the OTDT reactor trip. The
inspectors verified that these RTD shields were required to be grounded to the terminal
boards associated with protection channels 1 & 2 per a modification, DCP 89-41,
implemented in 1989, and properly completed on Unit 1. The inspectors concluded that
the failure to identify and correct the deficiencies associated with the July 29, 2003, event
was contrary to the requirements of 10 CFR 50, Appendix B, Criterion XVI, which
requires the establishment of measures to assure conditions adverse to quality are
promptly identified and corrected.
Analysis. The inspectors determined that the finding had a credible impact on safety
based on the deficiencies resulting in a reactor trip and a documented near-miss event.
The inspectors reviewed IMC 0612 and concluded the finding was more than minor
because it affected the Initiating Events cornerstone objective to limit the likelihood of
11
Enclosure
those events that upset plant stability and the cornerstone attribute of design control.
The inspectors referenced IMC 0609 for the SDP and concluded the finding is of very low
safety significance (Green) because it did not contribute to the likelihood of a primary or
secondary system LOCA, a loss of mitigation equipment functions, or the likelihood of a
fire or flood event. This finding contains aspects related to the cross-cutting area of
problem identification and resolution.
Enforcement. 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires the
establishment of measures to assure conditions adverse to quality are promptly and
identified and corrected. Contrary to the above, prompt identification and correction of
deficiencies relating to modification, DCP 89-41 was not performed following the
aforementioned Unit 2 reactor coolant instrumentation transient occurring on July 29,
2003, which resulted in a Unit 2 automatic reactor trip from actuation of an OTDT reactor
protection signal on August 5, 2005. This finding is of very low safety significance or
Green, is in the licensees CAP as Plant Issue N-2005-3016, and thus is characterized as
an NCV, consistent with Section VI.A of the NRC's Enforcement Policy: NCV 05000339/2005004-03, Failure to Identify and Correct Deficiencies in Instrumentation
Results In Reactor Trip.
.2
Unit 1 Rapid Power Reduction Due to Loss of Turbine Auto Stop Oil Pressure
Introduction: A Green, self-revealing finding was identified for not performing Unit 2
corrective actions in a timely manner on Unit 1. This resulted in the Unit 1 rapid
reduction of power from 100% to ~8% (main turbine off-line) on August 5, 2005.
Description: On August 5, 2005, the licensee rapidly reduced power on Unit 1 due to
severe oil leakage on the actuator for valve, 1-EH-TV-100 (Main Turbine Auto Stop Oil
Interface Valve). Subsequent evaluations determined that the torque specifications of
12-13 ft-lbs as specified in maintenance procedure 0-MCM-1412-01,Main Turbine
Interface Valve Diaphragm Replacement, did not provide adequate clamping force
between the diaphragm and actuator cover flange faces which resulted in diaphragm
movement and oil leakage from the actuator. The inspectors determined that an actuator
oil leak from the same valve resulted in a manual reactor trip due to low electro-hydraulic
or auto stop oil pressure on April 19, 2003. The inspectors reviewed the root cause
evaluation from that event and concluded that the licensee did not contact the vendor for
specific torque values. The inspectors also reviewed a December 2004, event involving
similar leakage on the Unit 2 equivalent valve. In this case, the resultant evaluation
concluded that the interface valve diaphragm torque values should have been 20 ft-lbs
per vendor technical manual 59-264-00006, Fisher Instruction Manual, Types 655 and
655R Actuators for Self-Operated Control. However, the inspectors determined that
associated corrective actions for Unit 1 had not been implemented prior to the August 5,
2005, rapid down-power event.
Analysis: This finding had a credible impact on safety due to the challenge of plant
control systems from the rapid reduction of power. The inspectors referenced IMC 0612
and determined that the finding was more than minor based on the impact to the Initiating
Events cornerstone objective to limit the likelihood of those events that upset plant
12
Enclosure
stability and the cornerstone attribute of equipment reliability. The inspectors referenced
IMC 0609 for the SDP and determined that the finding is Green (very low safety
significance) because it did not contribute to the likelihood of a primary or secondary
system LOCA initiator or a loss of mitigation equipment functions, and did not increase
the likelihood of a fire or internal/external flood. This issue is in the licensees CAP as
Plant Issue N-2005-2984. This finding contains aspects relating to the cross-cutting area
of problem identification and resolution.
Enforcement: Since this finding is associated with nonsafety-related secondary plant
equipment, no violation of regulatory requirements occurred. Therefore, this finding is
identified as a Green finding FIN 05000338/2005004-04, Untimely Corrective Actions for
Actuator Oil Leakage on Turbine Interface Valve Results in Rapid Down Power.
1R15
Operability Evaluations
a.
Inspection Scope
The inspectors reviewed six 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
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-2005-2751, licensee identified problem with oil leaking from 1J
EDG exhaust manifold with the diesel in a stand by condition;
Plant Issue N-2005-2866, during the overspeed test of Station Blackout Diesel
per 0-MCM-0710-03 the BIMBA fuel rack air cylinder did not fully extend after the
diesel tripped as required by acceptance criteria;
Plant Issue N-2005-2927, NRC identified problem with loose control rods on SW
expansion joints 2-SW-REJ-24A through 2-SW-REF-24H;
Plant Issue N-2005-3240, Quench Spray pump operable but degraded due to out
of tolerance A&C phase instantaneous overcurrent settings on breakers;
Plant Issue N-2005-2937, Recirculation Spray seal accumulator high level alarms;
and,
Plant Issue N-2005-3527, ESGR HVAC units 2-HV-AC-7, 2-HV-AV-6 and 1-HV-
AC-7 have access cover latches that are very loose and can be pulled off with
little effort.
b.
Findings
No findings of significance were identified.
13
Enclosure
1R17
Permanent Plant Modifications
a.
Inspection Scope
The inspectors reviewed the completed permanent plant modification DCP 04-019,
Replacing RSST Underground Cables - Unit 1. The inspectors conducted a walkdown of
the installation, discussed the desired improvement with system engineers, and reviewed
the 10 CFR 50.59 Safety Review/Regulatory Screening, technical drawings, test plans
and the modification package to assess TS implications.
b.
Findings
No findings of significance were identified.
1R19
Post Maintenance Testing
a.
Inspection Scope
The inspectors reviewed seven 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.
Procedure 2-PT-14.2, Charging Pump 2-CH-P-1B per WO 487833 and Plant
Issue N-2005-2472;
Procedure 0-MCM-0701-20, Repair of EDG Pre-lube and Standby Lube Oil
Pumps per WO 604135;
Procedure 2-PT-64.4A, Casing Cooling Pump (2-RS-P-3A) Test per WO 602353;
Procedure 0-ICM-XX-AOV-001, AOV Inspection and Diagnostic Testing per WO 606826 for work on 2-FW-FCV-2499;
Procedure 0-MCM-0701-34, Removal and Installation of EDG Exhaust Manifold,
and 2-PT-82H, 2H EDG Slow Start Test per WO 722151;
Procedure 0-EPM-03202-02 and 0-EPM-302-4, BBC / ITE 480 Volt K-Line
Breaker and Associated Switchgear Cubicle Maintenance per WOs 528002-05,
515205-01, and 528002-03; and,
Procedure 1-PT-74.2A, Component Cooling Pump 1-CC-P-1A Test per WO 722256
14
Enclosure
b.
Findings
No findings of significance were identified.
1R22
Surveillance Testing
a.
Inspection Scope
For the nine 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:
1-PT-63.1A, Quench Spray System A Subsystem (1-QS-P-1A), an inservice
test,
2-PT-71.2Q, Unit 2 Motor Driven Auxiliary Feedwater (2-FW-P-3A) Pump Test;
1-PT-52.2, Reactor Coolant System Leak Rate (Hand Calculation) VPAP-0502 -
Procedure Process Control;
2-PT-82J, 2J Diesel Generator Test Slow Start Test;
2-PT-63.1B, Quench Spray System - B Subsystem;
2-PT-213.8B, Valve Inservice Inspection (B Train of Safety Injection System);
2-PT-31.7, Pressurizer Level Channel (2-RC-L-2459) Channel Operational Test;
1-PT-75.2B, Unit 1 Service Water Pump (1-SW-P-1B); and,
2-PT-57.1B, Emergency Core Cooling Subsystem - Low Head Safety Injection
Pump (2-SI-P-1B).
b.
Findings
.1
Failure to Follow Procedures During SSPS Testing
Introduction. A Green, self-revealing NCV of TS 5.4.1.a was identified for failure to
implement a surveillance procedure which resulted in placing an incorrect bistable in a
trip condition.
Description. On July 22, 2005, during the performance of SSPS testing on Unit 2 in
accordance with procedure 2-PT-31.7, Pressurizer Level Channel I (2-RC-L-2459)
Channel Operational Test, of which step 6.1.5 requires placement of trip switches BS1
and BS2 on card C1-442 in the trip position, instrument technicians incorrectly placed
switches BS1 and BS2 on card C1-422 (same switch designation but a different card) in
the test position, which initiated an unexpected alarm (LO LO Tave Interlock Loop 1
A-B-C) in the control room. This caused Unit 2, Loop 1 T cold inputs to the SSPS
Relays K148 (Lo Lo Tave)(BS1) and K140 (Lo Tave)(BS2) to fail safe and show a trip
condition. A subsequent review by the inspectors of I/C drawings revealed that these
relays were Channel I inputs for P-12 (Lo Lo Tave Steam Dump Interlock) and feedwater
isolation permissives. The inspectors concluded that since loops two and three were not
in a trip condition, the two out of three logic was not satisfied, and the plant was not
affected.
15
Enclosure
Analysis. The inspectors reviewed IMC 0612 and determined that the finding was more
than minor because it could reasonably be viewed as a precursor to a more significant
event. If another channel in the logic had already been tripped, the plant would have
been adversely affected by the performance deficiency. The inspectors consulted IMC 0609 for the SDP and determined that the finding is Green (very low safety significance)
because it did not involve any LOCA initiators, did not contribute to both a reactor trip or
mitigating system unavailability, nor increase the likelihood of a fire. This finding contains
aspects relating to the cross-cutting area of human performance.
Enforcement. TS 5.4.1.a, requires that written procedures shall be established,
implemented, and maintained per RG 1.33, Appendix A, of which Part 8 stipulates
procedures for surveillance tests. Procedure, 2-PT-31.7.1, step 6.1.5. states, Place the
following comparator trip switches in TEST: On card C1-442, BS1 and BS2. Contrary to
the above on July 22, 2005, step 6.1.5 was improperly implemented in that comparator
switches, BS1 and BS2, on card C1-422 were placed in trip as opposed to the switches
on the correct card, C1-442. This finding is of very low safety significance or Green, is in
the licensees CAP as Plant Issue N-2005-2755, and thus is characterized as an NCV,
consistent with Section VI.A of the NRC's Enforcement Policy: NCV 05000339/2005004-04, Failure to Follow Procedure During Solid State Protection System
Testing.
.2
Failure to Follow Procedures Affecting Safety-Related Breakers
Introduction. A Green, self-revealing NCV of TS 5.4.1.a was identified for a failure to
follow procedures resulting in a trip of the Unit 2 Quench Spray Pump, 2-QS-P-1B.
Description. On August 19, 2005, during performance testing of 2-QS-P-1B per
2-PT-63.1B, Quench Spray System - B Subsystem, the respective motor breaker,
2-EE-BKR-24J1-4, closed and then immediately tripped open. The licensee
subsequently determined that two of the three as-found phase values of the breaker
overload device instantaneous pickup were low when compared to the North Anna
Setpoint Document (NASD) procedure which contains the setpoints, trip times and test
currents for all overload trip devices for 480-volt BBC/ITE K-Line Breakers. Therefore,
the motor starting current of approximately 3028 amps compared to the overload
instantaneous setpoints of 2268 amps and 2912 amps for B and C phases respectively
resulted in a premature trip of the breaker. The licensee previously performed
maintenance on this breaker on February 19, 2005, when the overload devices were set
and tested in accordance with electrical maintenance procedure, 0-EPM-302-02,
BBC/ITE 480-volt K-Line Breaker & Associated Switchgear Cubicle Maintenance,
which references the NASD. Procedure 0-EPM-302-02, step 6.19.4.a.2 states, If the trip
setpoint is within tolerance (80-120 percent) that was recorded in step 6.19.1, then go to
substep 6.19.4.b, and if not, then make adjustments using Attachment 5, Instantaneous
And Short-Time Pickup Adjustment, and repeat steps 6.19.4.a.1 and 6.19.4.a.2.
Contrary to the above, the technician performing the maintenance left the B and C
phase instantaneous overload setpoints low outside of the allowable procedural tolerance
at 3030 & 3002 amps respectively instead of within the allowable procedural tolerance of
3080 to 4620 amps. The licensee determined that a contributing cause was setpoint drift
16
Enclosure
on the associated overload device. However, the inspectors determined that given the
worst case drift, B phase at 812 amps, and an initial setpoint of 3850 amps (middle of
the established ban), the resulting drift would have resulted in a value above the motor
starting current.
Analysis. The inspectors referenced IMC 0612 and determined that the finding was more
than minor because it affected the Barrier Integrity cornerstone objective to provide
reasonable assurance that the containment physical design barriers protect the public
from radio nuclide releases caused by accidents or events and the cornerstone attribute
of human performance. The inspectors referenced IMC 0609 for the SDP and
determined that the finding is Green (very low safety significance) because it did not
impact design deficiencies, result in a loss of system safety functions, exceed related TS
outage times, nor involve a seismic, flooding, or severe weather initiating event. This
finding contains aspects relating to the cross-cutting area of human performance.
Enforcement. TS 5.4.1.a, requires that written procedures shall be established,
implemented, and maintained as documented in RG 1.33, Appendix A, of which Part 9
stipulates procedures for maintenance. Procedure 0-EPM-302-02, step 6.19.4.a.2
stated, If the trip setpoint is within tolerance (80-120 percent) that was recorded in step
6.19.1, then go to substep 6.19.4.b, and if not, then make adjustments using Attachment
5, Instantaneous And Short-Time Pickup Adjustment, and repeat steps 6.19.4.a.1 and
6.19.4.a.2. Contrary to the above, on February 19, 2005, this step was not properly
implemented or followed resulting in improper instantaneous overload setpoints on B
and C phases and a subsequent trip of 2-QS-P-1B. This finding is of very low safety
significance or Green, is in the licensees CAP as Plant Issue N-2005-3225, and thus is
characterized as an NCV, consistent with Section VI.A of the NRC's Enforcement Policy:
NCV 05000339/2005004-05, Failure to Follow Procedures Affecting Safety-Related
Breakers.
1R23
Temporary Plant Modifications
a.
Inspection Scope
The inspectors reviewed two temporary plant modifications to verify that the modifications
did not affect system operability or availability as described by the TS and UFSAR. In
addition, the inspectors verified that the installation of the temporary modifications was in
accordance with the work package, that adequate controls were in place, procedures and
drawings were updated, and post-installation tests verified the operability of the affected
systems.
The temporary plant modifications reviewed were:
Temporary Modification 2005-1759, Install 3" Float Stop Backflow Preventer in
Floor Drains Located in Emergency Switchgear Fan Rooms; and,
Temporary Modification 2005-1761, Construction of a Temporary Dam (approx
1/2" tall) on top of Tandem Seal Package for 1-RS-P-2A to help trouble shoot
numerous seal head tank HI-LO level alarms (IT-C4).
17
Enclosure
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP2 Alert and Notification System Testing
a.
Inspection Scope
The inspectors evaluated the adequacy of licensee methods for testing the alert and
notification system in accordance with NRC Inspection Procedure 71114, Attachment 02,
Alert and Notification System (ANS) Testing. The applicable planning standard 10 CFR Part 50.47(b)(5) and its related 10 CFR Part 50, Appendix E, Section IV.D requirements
were used as reference criteria. The criteria contained in NUREG-0654, Criteria for
Preparation and Evaluation of Radiological Emergency Response Plans and
Preparedness in Support of Nuclear Power Plants, Revision 1, was also used as a
reference.
The inspectors reviewed various documents which are listed in the Attachment to this
report.
b.
Findings
No findings of significance were identified.
1EP3 Emergency Response Organization Augmentation
a.
Inspection Scope
The inspectors reviewed the Emergency Response Organization (ERO) augmentation
staffing requirements and the process for notifying the ERO to ensure the readiness of
key staff for responding to an event and timely facility activation. The inspectors
reviewed the results of the February 22, 2005, unannounced off-hours augmentation drill
and reviewed the backup notification systems. The qualification records of key position
ERO personnel was reviewed to ensure all ERO qualifications were current. A sample of
problems identified from augmentation drills or system tests performed since the last
inspection were reviewed to assess the effectiveness of corrective actions.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 03, Emergency Response Organization (ERO) Augmentation Testing. The
applicable planning standard, 10 CFR 50.47(b)(2) and its related 10 CFR 50, Appendix E
requirements were used as reference criteria.
The inspectors reviewed various documents which are listed in the Attachment to this
report.
18
Enclosure
b.
Findings
No findings of significance were identified.
1EP4 Emergency Action Level and Emergency Plan Changes
a.
Inspection Scope
The inspectors evaluated the associated 10 CFR 50.54(q) reviews associated with non-
administrative emergency plan, implementing procedures and Emergency Action Level
(EAL) changes. The inspectors reviewed Emergency Plan revisions 29 and 30 and
reviewed 10 CFR 50.47(q) evaluations for the period covering July 2004 to July 2005.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 01, Emergency Action Level and Emergency Plan Changes. The
applicable planning standard, 10 CFR 50.47(b)(4) and its related 10 CFR 50, Appendix E
requirements were used as reference criteria. The criteria contained in NUREG-0654,
Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and
Preparedness in Support of Nuclear Power Plants, Revision 1 and RG 1.101 were also
used as references.
The inspectors reviewed various documents which are listed in the Attachment to this
report.
b.
Findings
No findings of significance were identified.
1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies
a
Inspection Scope
The inspectors reviewed the corrective actions identified through the Emergency
Preparedness (EP) program to determine the significance of the issues and to determine
if repeat problems were occurring. The facilitys self-assessments and audits were
reviewed to assess the licensees ability to be self-critical, thus avoiding complacency
and degradation of their EP program. In addition, inspectors reviewed licensees self-
assessments and audits to assess the completeness and effectiveness of all EP-related
corrective actions.
The inspection was conducted in accordance with NRC Inspection Procedure 71114,
Attachment 05, Correction of Emergency Preparedness Weaknesses and Deficiencies.
The applicable planning standard, 10 CFR 50.47(b)(14) and its related 10 CFR 50,
Appendix E requirements were used as reference criteria.
The inspectors reviewed various documents which are listed in the Attachment to this
report.
19
Enclosure
b.
Findings
No findings of significance were identified.
1EP6 Drill Evaluation
a.
Inspection Scope
On September 13, 2005, the inspectors reviewed and observed the performance of an
simulator drill that involved a loss of Bearing Cooling Pumps with a subsequent reactor
trip, followed by increased primary plant leakage, a small block LOCA and a loss of
instrument air. The inspectors assessed emergency procedure usage, emergency plan
classification, notifications, and the licensees identification and entrance of any problems
into their CAP. This inspection evaluated the adequacy of the licensees conduct of the
drill and critique performance. Drill issues were captured by the licensee in their CAP
and were reviewed by the inspectors.
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas
a.
Inspection Scope
Access Control. Licensee activities for monitoring workers and controlling access to
radiologically significant areas were inspected. The inspectors evaluated procedural
guidance and directly observed implementation of administrative and physical controls;
appraised radiation worker and technician knowledge of, and proficiency in implementing,
Radiation Protection (RP) program activities; and assessed worker exposures to
radiation and radioactive material.
Radiological postings and material labeling were directly observed during tours of the
auxiliary building, external buildings and the independent spent fuel storage installation
(ISFSI). Inspectors conducted independent surveys in the auxiliary building and the
ISFSI to verify posted radiation levels and to compare with current licensee survey
records. During plant tours, control of High Radiation Area (HRA), HRA with dose rates
greater than 15 rem/hr and very HRA keys and the physical status of HRA doors were
examined. In addition, the inspectors observed radiological controls for non-fuel items
stored in the spent fuel pools. The inspectors also reviewed selected RP procedures and
radiation work permits (RWPs), and discussed current access control program
implementation with RP supervisors.
20
Enclosure
During the inspection, radiological controls for work activities in HRAs were observed and
discussed. The inspectors observed workers adherence to RWP guidance and Health
Physics Technician (HPT) proficiency in providing job coverage. Controls for limiting
exposure to airborne radioactive material were reviewed and operation of ventilation units
and positioning of air samplers were also observed. The inspectors evaluated electronic
dosimeter alarm set points for consistency with radiological conditions in auxiliary
building, decontamination building and the ISFSI. In addition, the inspectors interviewed
workers to assess knowledge of RWP requirements.
The inspectors evaluated worker exposures through review of data associated with
discrete radioactive particle and dispersed skin contamination events. Controls used for
monitoring extremity doses and the placement of dosimetry when work involved
significant dose gradients were reviewed. The inspectors discussed the processes that
would be used if an individual were to have an uptake of radioactive materials.
RP program activities were evaluated against 10 CFR Part 20; RG 8.38, Control of
Access to High and Very High Radiation Areas in Nuclear Power Plants; and approved
licensee procedures. Licensee guidance documents, records, and data reviewed are
listed in the Attachment.
Problem Identification and Resolution. Five plant issues and two audits associated with
radiological controls, personnel monitoring, and exposure assessments were reviewed
and discussed with RP supervisors. The inspectors assessed the licensees ability to
identify, characterize, prioritize, and resolve the identified issues in accordance with
licensee procedures VPAP-1501, Deviations, and VPAP-1601, Corrective Action.
Specific documents reviewed are listed in the Attachment.
b.
Findings
No findings of significance were identified.
2OS3 Radiation Monitoring Instrumentation
a.
Inspection Scope
Radiation Monitoring Instrumentation and Post-Accident Sampling. During tours of the
auxiliary building and Spent Fuel Pool building, the inspectors observed installed
radiation detection equipment including the following instrument types: Area Radiation
Monitors (ARMs), Continuous Air Monitors (CAMs), Personnel Contamination Monitors
(PCMs), and components of the Post-Accident Sampling System (PASS). The
inspectors observed the physical location of the components, noted the material
condition, and compared sensitivity ranges with the UFSAR. The inspectors also
observed HPT selection and use of portable instruments during a survey of the ISFSI
perimeter fence and support of work in a decontamination building.
In addition to equipment walk-downs, the inspectors observed functional checks and
alarm setpoint testing of various fixed and portable detection instruments. These
21
Enclosure
observations included response checks of portable ion chambers and teletectors, PCMs,
Small Article Monitors (SAMs), Portal Monitors, and a Whole Body Counter (WBC). The
10 CFR Part 61 analysis for Dry Active Waste was reviewed to determine if calibration
and response check sources are representative of the plant source term.
The inspectors reviewed calibration records for a selected PCM, portal monitors, SAM,
and WBC, ARM channel RM-153, Fuel Pit Bridge ARM, and for all Unit 1 containment
high-range ARMs (channels RM-165 and 166). The records were evaluated to determine
frequency and adequacy of the calibrations. Calibration stickers on portable survey
instruments were noted during inspection of storage areas for ready-to-use equipment.
In addition, the inspectors discussed in-place radiation detection system reliability with
the responsible engineer.
Operability and reliability of selected radiation detection instruments were reviewed
against details documented in the following: 10 CFR Part 20; NUREG-0737, Clarification
of TMI Action Plan Requirements; TS Section 3; UFSAR Chapter 12; and applicable
licensee procedures. Documents reviewed during the inspection are listed in the
Attachment.
Self-Contained Breathing Apparatus (SCBA) and Protective Equipment. Selected SCBA
units staged for emergency use in the Control Room and other locations were inspected
for material condition, air pressure, and number of units available. The inspectors also
reviewed maintenance records for components of selected SCBA units for the past five
years and certification records associated with supplied air quality.
Qualifications for licensee staff responsible for testing and repairing SCBA equipment
were evaluated through review of manufacturer training certificates. In addition, selected
Control Room operators were interviewed to determine their knowledge of available
SCBA equipment locations, including corrective lens inserts if needed, and their training
on bottle change-out during periods of extended SCBA use. Respirator qualification
records were reviewed for several Control Room operators and Maintenance department
personnel assigned emergency response duties.
Licensee activities associated with maintenance and use of respiratory protection
equipment were reviewed against 10 CFR Part 20; RG 8.15, Acceptable Programs for
Respiratory Protection; ANSI-Z88.2-1992, American National Standard for Respiratory
Protection; and applicable licensee procedures. Documents reviewed during the
inspection are listed in the Attachment.
Problem Identification and Resolution. Five plant issues and one audit associated with
instrumentation and protective equipment were reviewed and assessed. The inspectors
evaluated the licensees ability to identify, characterize, prioritize, and resolve the
identified issues in accordance with procedure VPAP-1601, Corrective Action.
Documents reviewed are listed in the Attachment.
22
Enclosure
b.
Findings
No findings of significance were identified.
Cornerstone: Public Radiation Safety
2PS1 Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems
a.
Inspection Scope
Effluent Processing Equipment. The inspectors reviewed the operability and reliability of
selected radioactive effluent process sampling and detection equipment used for routine
and accident monitoring activities. Inspection activities included review of the most
recent calibration records and direct observation of select monitors. The inspectors
observed the material condition of the effluent monitoring equipment and assessed the
installed configurations, where accessible. The inspectors also reviewed applicable parts
of licensee procedures related to effluent monitoring equipment calibration.
Selected parts of the liquid radioactive waste (radwaste) system were examined and
reviewed with cognizant count room staff. The inspectors discussed with cognizant count
room staff liquid waste release permits. In addition, the inspectors directly observed the
collection and analysis of liquid effluent samples taken from the clarifier tank.
Major waste gas system components were inspected and discussed with cognizant count
room staff. Also, cognizant count room staff were interviewed regarding the gaseous
radwaste system configuration and effluent monitor operation. Inspectors also observed
Instrumentation and Calibration staff performing a calibration of the service water
discharge radiation monitor (RM-SW-108).
Installed configuration, material condition, operability, and reliability for selected effluent
sampling and monitoring equipment were reviewed against details documented in
10 CFR Part 20; UFSAR Section 11, Off-Site Dose Calculation Manual (ODCM); and RG 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases
of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled
Nuclear Power Plants." Procedures and records reviewed during the inspection are
listed in the Attachment.
Effluent Release Processing and Quality Control (QC) Activities. The inspectors directly
observed and evaluated licensee proficiency in effluent release processing during
preparation of a containment purge weekly release permit. The inspectors also reviewed
effluent release procedural guidance.
QC activities regarding gamma spectroscopy and liquid scintillation counting
instrumentation were discussed with cognizant count room staff. The inspectors
reviewed records of daily QC checks and trending data for selected gamma
spectroscopy detectors. In addition, results of the radiochemistry cross-check program
23
Enclosure
were discussed for years 2003 and 2004. The inspectors also reviewed the 2003 and
2004 Annual Effluent Reports to identify any anomalous releases.
Observed task evolutions, offsite dose results, and count room activities were evaluated
against RG 1.21 guidance, 10 CFR Part 20 requirements, Appendix I to 10 CFR Part 50
design criteria, UFSAR details, and ODCM requirements. Documents reviewed are listed
in the Attachment.
Problem Identification and Resolution. Select plant issues associated with effluent
release activities were reviewed and assessed. The inspectors evaluated the licensees
ability to identify, characterize, prioritize, and resolve the identified issues in accordance
with procedure VPAP-1601, Corrective Action, and associated guideline documents.
Documents reviewed are listed in the Attachment.
b.
Findings
No findings of significance were identified.
2PS3 Radiological Environmental Monitoring Program (REMP) and Radioactive Material
Control Program
a.
Inspection Scope
REMP Implementation. The inspectors reviewed the licensees most recent Annual
Radiological Environmental Operating Reports for 2003 and 2004 which described
implementation of the REMP and provided an assessment of the program results.
Information regarding surveillance results, analysis of data, land use census, the
interlaboratory comparison program, and permitted program deviations were evaluated.
The inspector also reviewed and discussed implementation of the REMP with respect to
sampling locations, monitoring and measurement frequencies.
The inspectors observed collection of air particulate filters and charcoal cartridges at five
air sampling stations and assessed sample collection methodology and techniques.
Calibration procedures and records for the air sampling stations were reviewed. The
inspectors also observed thermoluminescent dosimeters (TLDs) placement at eight
locations as described in the ODCM.
Through the above reviews and observations, the licensees practices and
implementation of their radiological monitoring program were evaluated by the inspectors
for consistency with the ODCM, UFSAR, TS, and 10 CFR Part 20 requirements.
Meteorological Monitoring Program. The inspectors reviewed the operability of the
meteorological monitoring equipment and operator access to meteorological data.
Current meteorological monitoring equipment performance was reviewed with the system
engineer. Licensee technicians primarily responsible for equipment maintenance and
surveillance were interviewed by the inspectors concerning equipment performance,
reliability, and routine inspections.
24
Enclosure
Calibration procedures and records for the two most recent calibrations of the
meteorological monitoring instruments for air temperature and for wind speed and
direction were also reviewed. The inspectors evaluated the operability of instruments
and determined the availability of current meteorological conditions displayed in the
Control Room for the primary tower.
Meteorological monitoring program implementation and results were reviewed against
TS, ODCM guidance, and procedures listed in the Attachment.
Unrestricted Release of Materials from the Radiologically Controlled Area (RCA). The
inspectors reviewed and evaluated radiation protection program activities associated with
the unconditional release of licensed materials from RCA locations. Licensee guidance
and implementation of RCA exit monitoring activities were evaluated against 10 CFR Part 20 requirements and applicable procedures documented in the Attachment.
Problem Identification and Resolution. The inspectors reviewed audits, and selected
Plant Issues associated with REMP operations and the program for unrestricted release
of materials from the RCA. The inspectors assessed the licensees ability to identify,
characterize, prioritize, and resolve the identified issues in accordance with licensee
procedures VPAP-1601, Corrective Action. Specific Plant Issues reviewed and evaluated
in detail for these program areas are identified in the Attachment.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4AO1 Performance Indicator (PI) Verification
Emergency Preparedness PI Verification
a.
Inspection Scope
The inspectors reviewed the licensees procedure for developing the data for the
Emergency Preparedness PI which are: (1) Drill and Exercise Performance (DEP); (2)
ERO Drill Participation; and (3) ANS Reliability. The inspectors examined data reported
to the NRC for the period June, 2004, to June, 2005. Procedural guidance for reporting
PI information and records used by the licensee to identify potential PI occurrences were
also reviewed. The inspectors verified the accuracy of the PI for ERO drill and exercise
performance through review of a sample of drill and event records. The inspectors
reviewed selected training records to verify the accuracy of the PI for ERO drill
participation for personnel assigned to key positions in the ERO. The inspectors verified
the accuracy of the PI for alert and notification system reliability through review of a
sample of the licensees records of periodic system tests.
25
Enclosure
The inspection was conducted in accordance with NRC Inspection Procedure 71151,
Performance Indicator Verification. The applicable planning standards, 10 CFR 50.9
and NEI 99-02,Regulatory Assessment Performance Indicator Guidelines, Revision 3,
were used as reference criteria.
The inspectors reviewed various documents which are listed in the Attachment to this
report.
b.
Findings
No findings of significance were identified.
Radiation Safety PI Verification
a.
Inspection Scope
The inspectors sampled licensee records to verify the accuracy of reported PI data for
the periods listed below. To verify the accuracy of the reported PI elements, the
reviewed data were assessed against guidance contained in NEI 99-02, "Regulatory
Assessment Indicator Guideline," Rev. 3, and the Performance Indicator Frequently
Asked Questions (FAQ) list.
Occupational Radiation Safety Cornerstone
Occupational Exposure Control Effectiveness
The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for
the period of January 2004 through June 2005. For the assessment period, the
inspectors reviewed HP shift log entries, electronic dosimeter alarm logs, and licensee
procedural guidance for collecting and documenting Performance Indicator data. Plant
Issues were reviewed for uptakes and abnormal TLD results. Report section 2OS1
contains additional details regarding the inspection of controls for high dose areas and
review of related Plant Issues. Documents reviewed are listed in the Attachment.
Public Radiation Safety Cornerstone
Radiological Control Effluent Release Occurrences
The inspectors reviewed the Radiological Control Effluent Release Occurrences PI
results for the period of January 2004 through June 2005. For the assessment period,
the inspectors reviewed cumulative and projected doses to the public. The inspectors
also reviewed licensee procedural guidance for collecting and documenting PI data.
Documents reviewed are listed in the Attachment.
b.
Findings
No findings of significance were identified.
26
Enclosure
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 CAP. 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-2005-2320, during the performance of 1-PT-71.1Q (1-FW-P-2, Turbine Driven
Auxilliary Feedwater (TDAFW) pump), noted the outboard bearing slinger ring leaking oil
at approximately 3-4 drops per second. 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 the licensees CAP as specified in
VPAP-1601, Corrective Action Program, VPAP-1501, Deviations and 10 CFR 50,
Appendix B. Additional documents reviewed are listed in the Attachment.
b.
Findings and Observations
No findings of significance were identified. On June 21, 2005, the licensee initiated Plant
Issue N-2005-2320 in response to an oil leak on the Unit 1 TDAFW pump outboard
bearing identified during the quarterly surveillance test. The licensee completed a
functional evaluation and declared a GL 91-18 condition (operable but degraded) for the
component. During subsequent testing, the licensee better quantified the leak at 1.58
gallons per day as opposed to the original estimate of 8.5 gallons per day. The
inspectors verified the licensee functional evaluation which considered the following facts
that the design basis accident mission time for TDAFW operation is 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and that the
pump oil reservoir is maintained at 12 - 18 gallons of which 8 gallons are below pump
suction. This would result in a leakage of .53 gallons during the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> mission time
resulting in the maintenance of pump operability. The inspectors reviewed the history of
bearing oil leaks for the Unit 1 and 2 TDAFW pumps which included work order,
00505761-01, for an oil leak on the Unit 1 TDAFW pump outboard bearing which was
completed on September 18, 2004. The licensee subsequently identified this corrective
action as rework. The inspectors also found for the Unit 2 TDAFW pump an Item
Equivalency Evaluation Review (IEER) report, N95-5022-000, which installed new seals
of a different design due to similar problems of oil leakage. The licensee could not
explain why this same design had not been considered for the Unit 1 TDAFW pump. The
inspectors reviewed the IEER process as implemented by VPAP-0708, Item Equivalency
Evaluation, and the corrective action process as implemented by VPAP-1601 and
VPAP-1501. The inspectors determined that VPAP-0708 did not perform an extent of
27
Enclosure
condition review nor reference, consider or require a plant issue. The inspectors also
determined that neither VPAP-1601 or VPAP-1501 discussed the IEER process as part
of the CAP. The inspectors concluded the failure to implement adequate corrective
action for the Unit 1 TDAFW pump constituted a minor violation. This finding is not yet
captured in the licensees corrective action program.
4OA4 Cross-cutting Aspects of Findings
Section 1R04 describes a finding associated with human performance involving
inadequate design control relating to verification of design adequacy. The inspectors
determined that information relating to the SW system pressure loads was known within
the engineering organization. However, this information was neither transferred into the
modification design, nor was the design verified to ensure the S/Rs were adequate for
the replacement expansion joints.
Section 1R06 describes a finding for inadequate corrective action resulting a flood
problem for the safeguards instrument rack room. The inspectors determined that
previous corrective actions and plant area flood reviews failed to identify the
ACCR/ACFR floor drain flood path.
Section 1R14 documents two findings associated with corrective action problems:
The first finding involves two circumstances in which lightning impacts the same
Unit 2 instrumentation in each case with one involving a reactor trip. For both
cases, the licensee took no corrective action and instead attributed the cause to
either outside the control of the company, or an isolated event from the
extremely dry soil conditions affecting the grounding grid that year; and,
The second finding for inadequate corrective action concerns a failure to involve
the valve vendor to obtain important information relative to the problem. Once the
vendor was involved to obtain the correct torque information for the valve
actuator, actions were untimely and allowed a subsequent leak forcing a unit
shutdown.
Section 1R22 describes two findings associated with human performance relating to a
failure to follow procedure:
The first finding concerning SSPS testing involved two maintenance technicians
of which one incorrectly identified a card on which the trip switches would be
manipulated and the second incorrectly performed independent verification of the
card contrary to procedure requirements; and,
The second finding concerning breaker maintenance involved a supplemental
employee who failed to adhere to procedure requirements to ensure as left
overload setpoints were within the specified band.
28
Enclosure
4OA5 Other Activities
.1
(Closed) Temporary Instruction (TI) 2515/161 Transportation of Reactor Control Rod
Drives in Type A Packages
a.
Inspection Scope
The inspectors reviewed shipping logs and discussed shipment of Reactor Control Rod
Drives (CRD) in Type A packages with shipping staff. The inspectors noted that no
shipments of Reactor CRDs in Type A packages have been made since January 1, 2002.
b.
Findings
No findings of significance were identified.
.2
(Discussed) Temporary Instruction (TI) 2515/163, Operational Readiness of Offsite
Power
Completion of this TI was documented in NRC Inspection Report Nos. 05000338,
339/2005003. However, after an NRC headquarters review of the data provided,
additional information related to the TI was requested. The inspectors collected this
information from licensee discussions, site procedures and licensee documentation. The
information was subsequently provided to the headquarters staff for further analysis.
.3
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 a HPT and RP supervisory staff. Radiological controls for
loading the ISFSI casks were also reviewed and discussed.
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
b.
Findings
No findings of significance were identified.
4OA6 Meetings, including Exit
On September 22, 2005, the senior resident inspector and the reactor projects branch
chief presented the inspection results to Mr. Jack Davis and other members of the staff.
29
Enclosure
The licensee acknowledged the findings. The inspectors confirmed that proprietary
information was not provided or examined during the inspection.
4OA7 Licensee-Identified Violation
The following finding of very low 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 RG 1.33, Revision 2, Appendix A, February 1978, of which Part 2 requires general plant
operating procedures. Contrary to the above, on August 6, 2005, the licensee failed to
implement step 5.36 of operating procedure 1-OP-2.2, Unit Power Operation From
Mode 1 to Mode 2, which requires the performance of power range low setpoint channel
operational tests to comply with TS surveillance requirement 3.3.1.8. The licensee
discovered the procedure noncompliance during plant startup requirements, entered TS
SR 3.0.3 and successfully completed the required testing. The inspectors reviewed IMCs
0612 and 0609, and determined that the finding was of very low safety significance given
the successful completion of the surveillance tests. The licensee has this finding
documented in their CAP as Plant Issue N-2005-2980.
Attachment
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. Costello, Supervisor, Nuclear Emergency Preparedness (Virginia)
J. Davis, Site Vice President
R. Evans, Manager, Radiological Protection
R. Foster, Supply Chain Manager
S. Hughes, Manager, Nuclear Operations
P. Kemp, Supervisor, Nuclear Safety & Licensing
J. Kirkpatrick, Manager, Maintenance
L. Lane, Director, Operations and Maintenance
J. Leberstien, Licensing Technical Advisor
T. Maddy, Manager, Nuclear Protection Services
M. Main, Component Engineer
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
M. Sartain, Manager, Nuclear Engineering
J. Scott, Supervisor, Nuclear Training (operations)
G. Salomone, Licensing
R. Williams, Component Engineer
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
05000338, 339/2005004-02 URI
Inadequate Corrective Action Results in Safeguards
Instrument Rack Room Flood Problem (Section 1R06)
Opened and Closed
05000338, 339/2005004-01 NCV
Inadequate Design Control Results in Degradation of SW
Supports/Restraints (Section 1R04.2)05000338/2005004-03
Inadequate Maintenance of a Procedure Results in Loss of
Safety Related 480V Buses (Section 1R12)05000339/2005004-03
Failure to Identify and Correct Deficiencies in
Instrumentation Results In Reactor Trip (Section 1R14.1)
A-2
Attachment
Untimely Corrective Actions for Actuator Oil Leakage on
Turbine Interface Valve Results in Rapid Down power
(Section 1R14.2)05000339/2005004-04
Failure to Follow Procedures During Solid State Protection
System Testing (Section 1R22.1)05000339/2005004-05
Failure to Follow Procedures Affecting Safety-Related
Breakers (Section 1R22.2)
Closed
Inadequate Maintenance of a Procedure Results in Loss of
Safety Related 480V Buses (Section 1R12)
2515/161
TI
Transportation of Reactor Control Rod Drives In Type A
Packages (Section 4OA5)
Discussed
2515/163
TI
Operational Readiness of Offsite Power (Section 4OA5.1)
LIST OF DOCUMENTS REVIEWED
Section 1R04: Equipment Alignment
Documents
List of open work orders for Unit 2 SW components
List of plant issues since 2004 for Unit 2 SW components
TS 3.7.8, "Service Water (SW) System"
Plant Issue N-2005-2927, NRC identified issue with loose tie rods on SW
expansion joints associated with RS heat exchanger supply and return piping.
Plant Issue N-2005-3376, NRC identified issue with loose tie rods on SW
expansion joints associated with the SBO diesel generator.
0-OP-49.1, Service Water System Normal Operation
Module, NCRODP-13-NA, Service Water System
Root Cause Evaluation N-2005-2229, Damaged SW Supports
Engineering Transmittal, ET-CEM-05-0009, Documentation of the Results of the
Structural Review for As-Found Condition of Service Water supports in the Tie-in
Vault and Valve House Expansion Joint Vault, NAPS Units 1 & 2"
Calculation Number, CE-1799, Structural Operability Evaluation for Service
Water Lines in the Tie-In Vault and Valve House Expansion Joint Vault, NAPS 1
& 2"
Drawings
11715-FM-078A, B, C, series of flow diagrams for SW system
11715-PSSK-105AN.01, Sheets 1, 2, Pipe Support 1-WS-PH-E85.1 for
321/4"-WS-E85-151-Q3"
A-3
Attachment
11715-PSSK-105AN.02, Sheets 1, 2, 3, Pipe Support 1-WS-PH-E85.2 for
321/4"-WS-E85-151-Q3"
11715-PSSK-105AK.10, Sheets 1, 2, 3, Pipe Support 1-WS-PH-3.2 for
36"-WS-3-151-Q3"
11715-PSSK-105AK.06, Sheets 1, 2, 3, Pipe Support 1-WS-PH-4.2 for
36"-WS-4-151-Q3"
11715-WMKS-0105AMA, Sheet 1, Inservice Inspection Isometric WS Sys:36",
24", 18" Valve HSE Pipe#1"
11715-FP-5AN, Sheet 1, Plan & Sections Service Water Valve House Piping
11715-FP-5AK, Sheet 1, Service Water Buried Piping Tie-In
11715-WMKS-0105AK, Sheet 1, Inservice Inspection Isometric WS Sys:36"
Tie-In Vault
Section 1R05: Fire Protection
Documents
PI N-2005-3733, NRC identified issue regarding the lack of a corrective action
process to resolve deficiencies identified during fire drills.
UFSAR Section 9.5.1, Fire Protection System
0-FPMP-10.0, "Conduct of Fire Drills"
VPAP-2401, "Fire Protection Program"
Appendix A to Branch Technical Position APCSB 9.5-1, "Guidelines for Fire
Protection for Nuclear Power Plants Docketed Prior to July 1, 1976"
NFPA 27, "Private Fire Brigades, 1981"
Section 1R06: Flood Protection Measures
Documents
ET N-04-0043, Rev.0, Evaluation of the Potential for Flooding of the U2
Emergency Switchgear from the Turbine Building Through U2 Cable Vault Floor
Drain Check Valve, 1-DB-424"
Plant Issue N-2005-2605, Floodwalls, 1-BLD-FLW-7 and 2-BLD-FLW-5, between
the ACCR and ACFR on both units have exceeded their maintenance rule
performance criteria of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per year.
ET NAF 00-0069, Rev. 0, Summary of Components Considered in the IPE
Internal Flooding Analysis for Surry and North Anna Power Stations, Units 1&2
Calculation Number, ME-0782, Maximum Backflow Flowrate Through Floor Drain
Between Chiller Room and Fan Room - Elevation 252'-0" and 254'-0"
Plant Issue N-2005-2597, Licensee identified issue with air gap in Unit 1
instrument rack room sump pump discharge piping (i.e., piping discharges to a
funnel at elevation of adjacent fan room)
Plant Issue 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
Plant Issue N-2005-2251, licensee identified issue of modification, DCP
59-92-161 that installed a backflow preventer in the charging pump cubicle drain
as part of the internal flood protection program but station drawing, 11715-FB-9A,
sheet 1 was not revised to add mark numbers to the drawing
A-4
Attachment
Engineering transmittal, ET-CEP-00-0006, Rev. 0, Evaluation of The Potential
For Flooding In The Emergency Switchgear Rooms North Anna Power Station,
Units 1 & 2"
Engineering Work Request,90-131, NP-1971, Outside containment Flooding
Protection, recommended the modification of the instrument rack room sump
pumps for Unit 1 & 2. These pumps are to be automatic with level alarm
indication.
UFSAR Section 9.3.3.2, System Description
Drawings
11715-FB-26A, Plumbing Service Building - Sheet 1 of 1, Revision 18
Section 1R12: Maintenance Effectiveness
Documents
SDBD- NAPS-QS, Revision 06
WO 00487833, HHSI seal replacement (TYPICAL)
WO 52002302, SW hand torque valve (TYPICAL)
WO 51440701, Replace SI valve stem (TYPICAL)
Response to Adverse Trend Plant Issue N-2005-0478, electrical maintenance
August 2005
Procedures
VPAP-0815, Maintenance Rule Program, Revision 14
STD-GN-0044, Supplemental Maintenance Rule Guidelines, Revision 4
Plant Issues
N-2000-2600, bottled air flow problems
N-2001-2479, HHSI motors (TYPICAL)
N-2002-1125, 1-HV-P-22C motor problems
N-2002-1875, Service water spray arrays
N-2002-2951, refrigerant leak control room chiller
N-2002-3065, HHSI pump seals
N-2003-1801, SW stainless steel MIC problems
N-2004-0053, while repairing compressor terminal plate under WO 489188-01,
found internal compressor motor protection device disabled
N-2004-2193, 1H EDG
N-2004-2368, RWST chillers
N-2004-2382, transform deluge system heat detectors
N-2004-5064, service and instrument air compressor timers
N-2005-0605, 1-HV-P-22A, high vibration control room chiller
N-2005-1615, H/J 480 VAC buses
N-2004-2195, 2-QS-MR-1A #2 fan motor bad and locking up
N-2004-2368, found 01-QS-MR-1B breaker tripped, compressor failed due to
starter contacts welded together
N-2004-4434, RWST temp is 47 degrees with no chiller running
A-5
Attachment
N-2004-4844, observed 14 day trend of U2 RWST temperature shows continual
increase from 42 degrees to 46 degrees
N-2005-1177, DCP and parts delays may require re-evaluation f (a)1 corrective
action dates associated with replacement of 1/2 QS-MR-1A/B
N-2005-2494, 2-CH-P-1C had a 20 ml/minute outboard endbell leak
N-2005-2536, 2-CH-P-1C was disassembled for OB mechanical seal leakage and
high vibes on the OB bearing
N-2005-2584, 2-QS-MR-1A Unit 2 A RWST chiller tripped on oil failure relay
N-2005-2610, 2-QS-MR-1B, Unit 2 B RWST chiller tripped with RWST
temperature at 45 degrees
Section 1R17: Permanent Plant Modifications
Documents
Design Change Package 04-019, RSST 34.5 kV Cable Replacement / North
Anna / Units 1 & 2"
Procedure 1-MOP-26.77, A RSS Transformer and D Transfer Bus, Revision
18
Procedure 1-MOP-26.78, B RSS Transformer and E Transfer Bus, Revision
18-P2
Procedure 1-MOP-26.79, C RSS Transformer and F Transfer Bus, Revision 17
Drawings
11715-FE-1BB
11715-FE-1BD
11715-FE-1A
Section 1R19: Post Maintenance Testing
Documents
Plant Issue N-2005-1061, Valves 1-BC-MOV-127 and 2-BC-MOV-227 are being
installed to support new BC tower returning to service
Procedure 0-ECM-1401-03, General Maintenance of Electrical Motors, Revision
31
Procedure 0-ECM-0206-01, Installation of Lugs, Revision 6
Section 1EP2: Alert and Notification System Testing
Procedures
0-EPM-0501-01, Early Warning System Preventive Maintenance, Revision 14
0-PT-172.3, Early Warning System Polling Function Test, Revision 0
0-PT-172.2, Early Warning System Sirens Activation Monitoring, Revision 2
Records and Data
Siren Problem Tracking Report
Various Plant Issues written against NAPS Sirens
Six Quarterly Data packages from 3/1/04to 4/17/2005 for 0-PT-172.2, Early
Warning System Sirens Activation Monitoring
A-6
Attachment
Miscellaneous
Sterling Siren Technical manual
WPS-2800 Series High Power Voice & Siren System - Installation & Instruction
Manual
Section 1EP3: Emergency Response Organization Augmentation
Procedures
EPIP-3.05, Augmentation of Emergency Response Organization, Revision 2
Records and Data
VPAP-2601, Attachment 3, Augmentation Capability Assessment of Emergency
Response Organization, 02/22/2005 at 1800
VPAP-2601, Attachment 3, Augmentation Capability Assessment of Emergency
Response Organization, 03/18/2004 at 2000
VPAP-2601, Attachment 3, Augmentation Capability Assessment of Emergency
Response Organization, 04/07/2004 at 1900
Section 1EP4: Emergency Action Level and Plan Changes
Records and Data
10 CFR50.54(q) Review for North Anna Power Station Emergency Plan Revision
29
10 CFR50.54(q) Review for North Anna Power Station Emergency Plan Revision
30
North Anna Power Station Emergency Plan Revision 29
North Anna Power Station Emergency Plan Revision 30
Procedures
EPIP-1.01, Emergency Manager Controlling Procedure, Revision 40
EPIP-2.01, Notification of State and Local Governments, Revision 27
EPIP-4.07, Protective Measures, Revision 16
EPIP-1.06, Protective Action Recommendations, Revision 6
Section 1EP5: Correction of Emergency Preparedness Weakness and Deficiencies
Records and Data
Report of Declaration: Notification of Unusual Event Declared at North Anna
Power Station on October 8, 2004
North Anna Power Station June 7, 2005 Training Exercise/Medical Drill Critique
Results, Resolution Report and Ongoing Self Assessment
North Anna Power Station June 7, 2005 Training Exercise/Medical Drill Exercise
Manual
North Anna Power Station May 5, 2005 Training Exercise Critique Results,
Resolution Report and Ongoing Self Assessment
North Anna Power Station May 5, 2005 Training Exercise Manual
North Anna Power Station March 1, 2005 Training Exercise Critique Results,
Resolution Report and Ongoing Self Assessment
North Anna Power Station March 1, 2005 Training Exercise Manual
A-7
Attachment
Section 2OS1: Access Control to Radiologically Significant Areas
Procedures, Manuals, and Guides
Health Physics Procedure Number C-HP-1020.011, Radiological Protection
Action Plan During Diving Activities, Revision 3
Health Physics Procedure Number C-HP-1031.021, Dosimetry Requirements for
Site Restricted Areas, Revision 6
Health Physics Procedure Number C-HP-1031.022, RWP Dosimetry: Exposure
Control Support, Revision 9
Health Physics Procedure Number C-HP-1032.020, Radiological Survey Criteria
and Scheduling, Revision 5
Health Physics Procedure Number Dominion, NAPS, C-HP-1032.060,
Radiological Posting and Access Control, Revision 1
Health Physics Procedure Number C-HP-1032.061, High Radiation Area Key
Control, Revision 2
Health Physics Procedure Number C-HP-1081.010, Radiation Work Permits:
Preparing and Approving, Revision 7
Health Physics Procedure Number C-HP-1081.020, Radiation Work Permits:
RWP Briefing and Controlling Work, Revision 4
Health Physics Procedure Number C-HP-1081.040, Radiation Work Permits:
Providing HP Coverage During Work, Revision 1.14
Station Administrative Procedure (SAP), No. VPAP-1501, Deviations, Revision 17
SAP, No. VPAP-1601, Corrective Action, Revision 20
Radiation Work Permits
Radiation Work Permit 05-2-1212, Obtain a sample from Spent Resin Hold-up
Tank (1-LW-TK-1) in decontamination building basement. [LHRA]
Radiation Work Permit 05-2-1502, General entry during sub-atmospheric
conditions for the purpose of walkdowns, inspections, radiological surveys, minor
maintenance and adjustments [LHRA>15 rem/hr]
Radiation Work Permit 05-2-1503, General entry by Operations, Health Physics,
Security and assorted craft personnel for the performance of routine PTs,
surveys, inspections and corrective maintenance as required. [LHRA > 15 rem/hr]
Radiation Work Permit 05-2-1504, Survey, lifting and transferring radioactive
waste liners to include associated support and placing of material into liner [LHRA
>15 rem/hr]
Corrective Action Program (CAP) Documents/Audits
Audit 03-06: Radiological Protection/ Chemistry, 9/22/2003
Audit 04-08: Radiation Protection & Process Control Programs, 9/20/2004
Plant Issue N-2005-0149-R1, A worker entered a posted Radiation Area in the
TSC without the proper dosimetry (Digital Alarming Dosimetry). The area was
posted as a Radiation Area and Radiation Work Permit required for entry.
Plant Issue N-2005-1467-R1, Observed an increase (3X) in dose rates on the
remote monitoring dosimeter by 1-LW-491 located in the Unit 1 side of demin
alley.
Plant Issue N-2005-1898, Two TLDs with abnormal readings for which a TLD re-
evaluation was requested, were confirmed as having normal response by the
A-8
Attachment
vendor on 05/20/2005. The two individuals TLD readings for the first quarter
2005 were 147 and 163 mrem, while the DAD readings totaled 0 mrem.
Plant Issue N-2005-2010, The HP lock for the Fuel Building Basement to Decon
building basement jail bar door is sticking and will not allow the door to be
opened.
Plant Issue N-2005-2184, Employee issued a DAD against the wrong RWP. The
employee should have issued a DAD against Radiation Work Permit 05-2-1505;
instead the DAD was issued against Radiation Work Permit 05-2-1105
Self Assessment: ITC-SA-04-02, Assessment of NBU for Adverse Trends in
Radiological Protection Events, 04/29/04
Section 2OS3: Radiation Monitoring Instrumentation
Procedures
Health Physics Procedure Number C-HP-1042.450, Self-Contained Breathing
Apparatus Maintenance, Revision 10
Health Physics Procedure Number C-HP-1042.520, Respiratory Protection
Program Equipment Criteria and Verification, Revision 4
Procedure No. 0-FPMP-3, SCBA Operability Test, Revision 2
Procedure No. ICP-RM-1-RMS-165, Containment High Range Radiation
Monitoring System (RMS-165), Revision 14
Procedure No. ICP-RM-1-RMS-166, Containment High Range Radiation
Monitoring System (RMS-166), Revision 15
Calibrations, Surveillance Tests, and Licensee Records
10 CFR Part 61 Analysis, Dry Active Waste (U1, U2, and Common), 12/8/04,
8/30/04, and 9/17/03
FASTSCAN WBC Calibration, 3/16/05
MSA Factory Training Certificates for Individuals Qualified to Repair SCBA Vital
Components
PCM-1B Serial No. 176, Calibrations, 6/10/04 and 6/13/05
PM-7 Serial No. 372, Calibrations, 12/1/04 and 4/18/05
RM-153, Fuel Pit Bridge ARM Calibrations, 11/1/01 and 7/9/03
RM-165 and 166, U1 Containment High Range ARM Calibrations, 165: (9/25/04,
1/15/03, 3/10/03) and 166: (9/24/01, 8/9/01, 1/15/03, 3/10/03)
SAM-11 Serial No. 177A, Calibrations, 12/1/04 and 6/7/05
SCBA Air Regulator Number ND263131, Maintenance History, 7/18/00 - 8/25/04
SCBA Qualification Records, Selected Operations and Maintenance Department
Staff
Service Air Breathing Air Quality Analyses, 11/18/03, 3/24/04, 10/21/04, 3/30/05
Source Certificate Number 98CS5001061, Cs-137 SAM-11 Calibration Source
CAP Documents/Audits
Audit 04-08, Radiation Protection & Process Control Programs, 9/20/04
SAP, No. VPAP-1601, Corrective Action, Revision 20
Plant Issue N-2004-0182, Filter not in motion alarms are occurring frequently on
2-RM-RMS-259
A-9
Attachment
Plant Issue N-2004-0991, 1-RM-RMS-163 spiking and causing numerous Hi-Hi
alarms
Plant Issue N-2004-1384, Teletector failed performance check after being used to
survey HRA
Plant Issue N-2005-0575, Electronic dosimeter not turned on prior to attempted
RCA entry
Plant Issue N-2005-2714, Respiratory qualification report showed incorrect
expiration dates
Section 2PS1: Radioactive Gaseous and Liquid Effluent Treatment and Monitoring
Systems
Procedures, Guidance Documents, and Operating Manuals
Health Physics Procedure Number HP-3010.020, Radioactive Liquid Waste
Release Permits, Revision 9
Health Physics Procedure Number HP-3010.021, Radioactive Liquid Waste
Sampling and Analysis, Revision 17
Health Physics Procedure Number HP-3010.022, Radioactive Liquid Waste
Accountability and Dose Calculations, Revision 6
Health Physics Procedure Number HP-3010.023, Abnormal Liquid Release,
Revision 1
Health Physics Procedure Number HP-3010.030, Radioactive Gaseous Waste
Release Permits, Revision 9
Health Physics Procedure Number HP-3010.031, Radioactive Gaseous Waste
Sampling and Analysis, Revision 21
Health Physics Procedure Number HP-3010.032, Radioactive Gaseous Waste
Accountability and Dose Calculations, Revision 11
Health Physics Procedure Number HP-3010.033, Abnormal Gaseous Release,
Revision 15
Health Physics Procedure Number HP-3010.040, Radiation Monitoring System
Setpoint Determination, Revision 17
SAP, Offsite Dose Calculation Manual (North Anna), Procedure Number VPAP-
2103N, Revision 7
Instrument Calibration Procedure, Number 0-1CP-SW-RM-108, Service Water
Discharge Radiation Monitor (RM-SW-108) Calibration, Revision 4
Records, Data, and Drawings
Calibration Certificates - Beckman LS-6000SC Dated 06/23/03 and Gamma
Products G-5020 Dated 06/02/04
Condenser Air Ejector In-Line Radio Gas Radiation Monitor (RM-SV-121 and 221)
Channel Calibrations, Test Results Dated 10/22/04 and 04/09/05
Discharge Tunnel Effluent Radiation Monitors (RM-SW-130 and 230) Channel
Calibrations, Test Results Dated 05/05/05 and 08/22/04
Radiological Environmental Monitoring Program, 2003
ECCS PREACS Train A and B Filter In-Place Tests (1-HV-FL-3A and 3B), Test
Results Dated 11/14/03 and 04/30/04
A-10
Attachment
Effluent Radiation Monitor Setpoint Records for 01-GW-RM-178-1, 1-SS-RM-125,
1-SV-RM-121, 1-SW-RM-108, 1-SW-RM-130, 1-VG-RM-179-1, 1-VG-RM-180-1,
2-SS-RM-225, 2-SV-RM-221, 2-SW-RM-230, RM-LW-111
Gaseous Effluents Cumulative Dose Summary for 2004 Through May 2005
Heating and Ventilation Flow A (F-HV-1212A) and B (F-HV-1212B) Channel
Calibrations, Test Results Dated 04/26/05
High Capacity Steam Generator Blowdown Radiation Monitors (RM-SS-125 and
225) Calibrations, Test Results Dated 12/15/04 and 03/24/05
Liquid Effluents Cumulative Dose Summary for 2004 Through May 2005
Liquid Waste Batch Release Permit and Record, Permit No. 04-LBATCH-01
Dated 05/14/05
Liquid Waste Clarifier Radiation Monitor (RM-LW-111) Channel Calibration, Test
Results Dated 09/03/04
Miscellaneous Gaseous Release Records, Permit Numbers 04-MGR-54 Dated
05/06/04, 04-MGR-125 Dated 09/13/04, 04-MGR-128 Dated 09/15/04, and 04-
MGR-130 Dated 09/16/04
NAPS, First Quarter 2004 Count Room Confirmatory Measurements
Process Vent Blowers Discharge Flow (1-GW-F-108) Calibration, Test Results
Dated 04/14/04
Process Vent Normal and High Range Effluent Radiation Monitor (GW-RM-178)
Channel Calibration, Test Results Dated 02/01/05
Reactor Containment Release Records, Permit Nos. 04-RXC-01 Dated 05/02/04
and 04-RXC-12 Dated 10/02/04
Results of Radiochemistry Cross Check Program, North Anna Power Station,
Third Quarter 2003
Service Water Discharge Radiation Monitor (RM-SW-108) Calibration, Test
Results Dated 07/20/05
Unplanned Gaseous Release Record, ID No. 05-AGR-01 Dated 04/04/05
Vent Stack A and B Normal and High Range Effluent Radiation Monitors (VG-RM-
179 and 180)
Channel Calibrations, Test Results Dated 02/24/05 and 06/04/04
CAP Documents/Audits
Nuclear Oversight Audit Report, No. 03-11, Offsite Dose Calculation Manual
Radiological Environmental Monitoring Program and Environmental Protection
Program, Dated 02/25/04
SAP, Number VPAP-1601, Corrective Action, Revision 20
Plant Issue N-2003-3417, Operations Failed to Notify the Health Physics Count
Room Prior to Performing Make-Up to the Unit #2 RWST
Plant Issue N-2004-3756, 1-HV-MOD-102A, A Fuel Building Exhaust Fan
Discharge Damper, Was Noted to Not Come Open When Restoring Fuel Building
Ventilation
Plant Issue N-2004-4089, A Service Water Sample From the Catch Basin
244'Auxiliary Building Used for Draining Service Water Prior to Pumping to Storm
Drains indicated the Presence of Licensed Material
Plant Issue N-2005-1121, Received Alert and Hi Alarms on 1-RI-VG-180, B Vent
Stack Gaseous RM
A-11
Attachment
Plant Issue N-2005-2031, While Performing 1-PT-38.1.11 (Liquid Waste
Radiation Monitor), Technicians Noted 1-LW-LCV-111 Repositioned to the Closed
Position on Receiving the Hi-Hi Radiation Alarm on 1-LW-RM-111 During PT
Radioactive Effluent Control Program Evaluation, 4th Quarter 2003 to 2nd Quarter
2005
Section 2PS3: Radiological Environmental Monitoring Program (REMP) and Radioactive
Material Control Program
Procedures and Guidance Documents
Health Physics Procedure Number C-HP-1033.440, NE Technology Sam-9/SAM-
11 Calibration and Operation, Revision 1
Health Physics Procedure Number C-HP-1033.620, Portable Air samplers
Calibration and Operation, Revision 4
Health Physics Procedure Number HP-3051-010, Radiological environmental
monitoring Program, Revision 15
Procedure No. 0-HPS-ISFSI-001, Independent Spent fuel Storage Installation
(ISFSI), Health Physics TLD Survey Surveillance, Revision 3
Procedure Number 0-ICP-MM-DP-1, Primary Meteorological Tower Dew Point
Measuring System Calibration, Revision 6
Procedure Number 0-ICP-MM-RG-1, Primary Meteorological Tower Precipitation
Monitor Calibration, Revision 5
Procedure Number 0-ICP-MM-S-101A, Weather Tower 48 Meter Wind Speed
Calibration, Revision 8
Procedure Number 0-ICP-MM-S-101B, Weather Tower 10 Meter Wind Speed
Calibration, Revision 10
Procedure Number 0-ICP-MM-T-100A, Weather Tower 10 Meter Temperature
Calibration, Revision 9
Procedure Number 0-ICP-MM-T-100B, Weather Tower 10/48 Meter Delta
Temperature Calibration Revision 10
Procedure Number 0-ICP-MM-Temp-1, Primary Meteorological Tower Ambient
Temperature and Differential Temperature Calibration, Revision 11
Procedure Number 0-ICP-MM-Z-101A, Weather Tower 48 Meter Wind Direction
Calibration, Revision 8
Procedure Number 0-ICP-MM-Z-101B, Weather Tower 10 Meter Wind Direction
Calibration, Revision 8
Procedure Number 0-ICP-MM-ZR-1A, Primary Meteorological Tower 10 Meter
Wind Speed and Wind Direction Calibration, Revision 7
Procedure Number 0-ICP-MM-ZR-1B, Primary Meteorological Tower 48 Meter
Wind Speed and Wind Direction Calibration, Revision 8
Procedure Number 0-PT-487.10, Radiological Environmental Monitoring
Program, Land Use Census, Revision 8
Procedure Number 0-PT-487.21, Annual Radiological Environmental Operating
Report, Draft, Revision 5
Procedure Number 0-PT-487.22, Annual Radiological Environmental Operating
Report, Final, Revision 5
Memorandum, North Anna Meteorological Data, 01/28/05
VPAP-1601, Corrective Action, Revision 20
A-12
Attachment
Instrument Calibration and Environmental Data Records
2003 Annual Radiological Environmental Operating Report
2004 Annual Radiological Environmental Operating Report
Calibration Certificate Portable Environmental Air Sampler HiQ, Kit 1, 04/20/05
Calibration Certificate Portable Environmental Air Sampler HiQ, Kit 2, 04/20/05
Calibration Certificate Portable Environmental Air Sampler HiQ, Kit 3, 04/20/05
Calibration Certificate Portable Environmental Air Sampler HiQ, Kit 5, 04/28/05
Calibration Certificate Portable Environmental Air Sampler HiQ, Kit 6, 04/20/05
Framatone ANP Environmental laboratory Analytical Service Semi-Annual Quality
Status Report (January - June 2004)
CAP Documents
Procedure VPAP-1601, Corrective Action, Rev. 20
Plant Issue N-2001-3454, the Interior of the Trailers (Primary and Backup) Are In
Need of Upgrading/Refurbishment, 12/04/2001
Plant Issue N-2003-2304, the Current Assumption Regarding The Charcoal
Cartridge Collection Efficiency for Iodine Is Incorrect, 06/09/03
Plant Issue N-2003-2852-R, Change In Most Limiting Exposure Pathway Grass-
Cow-Milch to Vegetable/Broadleaf Vegetation, 07/23/03
Plant Issue N-2003-2986, Extension Cord Alarmed Sam -11 at Service Building,
08/04/03
Plant Issue N-2003-3342, Sam 9 Contamination Monitors Failed Performance
Check, 09/04/03
Plant Issue N-2004-0129-E1, Worker Alarms PM-7s Located at Protection Area
Exit, 01/14/04
Plant Issue N-2004-0435, Eye Bolt Found Near the Unit 1 Boron Recovery Tank
in the Yard, 02/10/04
Plant Issue N-2004-1389, Individual Alarmed PM-7 at PA When Exiting, 05/02/04
Plant Issue N-2004-1441-E1, Worker Alarmed PM-7 Upon Trying to Exit the PA,
05/04/04
Plant Issue N-2005-1571-E1, Individual Alarmed PM 7 While Attempting to Exit
Protected Area, 05/09/04
Plant Issue N-2004-1654, Individual Alarms PM-7 at Security, 05/11/04
Plant Issue N-2004-1788-E1, Spent Secondary Resin Was Released for Disposal
to Clean Trash, 05/16/04
Plant Issue N-2004-2811-E1, In-coming Positive Whole Body Count, 07/28/04
Plant Issue N-2004-4198, Worker Inappropriately Removed Radioactive Material
From An RCA, 09/30/04
Plant Issue N-2004-5094, Individual Alarmed PM-7 At Security, 11/03/04
Plant Issue N-2004-5126, Shackle Alarmed SAM-11 At Service Building Tool
Crib, 12/02/04
Plant Issue N-2005-1116, Two Acetylene Hoses With Yellow Paint In
Maintenance Shop, 03/22/05
Plant Issue N-2005-1152, Small Sledge Hammer With Yellow and Magenta Paint
Found In Personal Tool Box, 03/23/05
Plant Issue N-2005-2738, NRC Walk-down of Primary Met Tower Express
Concerns of Tree Height, 07/20/05
A-13
Attachment
Radiological Incident Investigation, Contaminated Hammer Found In Mechanics
Tool Box
Radiological Incident Investigation, Contaminated Shackle Found Outside RCA
Radiological Incident Investigation, Contaminated Wire Rope Rigging Found
Outside RCA
Radiological Incident Investigation, Plant Issue N-2003-2986, Contaminated
Extension Cord In Protected Area
Radiological Incident Investigation, Radioactive Material Detected By PM-7 Portal
Monitor Located At Security Exit (East) Badge # 4972
Radiological Incident Investigation, Radioactive Material Detected By PM-7 Portal
Monitor Located At Security Exit Badge # 5857 Plant Issue N-2004-1654
Response to Plant Issue N-2004-0435: Contaminated Eye Bolt Found Outside the
Section 4OA1: Performance Indicator Verification
Procedures
DNAP-2605, Emergency Preparedness Performance Indicators, Revision 1
Records and Data
Performance Indicator Monthly Data from January, 2004 thru June, 2005
Radiation Safety
Procedures
Procedure Number HPAP-2802, NRC Performance Indicator Program, Revision 3
SAP, Number VPAP-1501, Deviations, Revision 17
SAP, Number VPAP-1601, Corrective Action, Revision 20
Section 4OA5: Other Activities
ISFSI Radiological Controls
Procedures
Procedure Number 0-HSP-ISFSI-001, Independent Spent Fuel Storage
Installation (ISFSI), Health Physics TLD Survey Surveillance, Revision 3
Procedure Number HP-1020.012, Radiological Protection Action Plan During Dry
Storage Cask Activities, Revision 14
Radiation Work Permits
Radiation Work Permit 05-2-1107, Receive, prep, load, decon, leak test, and ship
the loaded NAC-LWT Cask includes all associated work
Temporary Instruction 2515/161, Transport of Control Rod Drive (CRD) in Type A Packages
Records
Radioactive Material Shipment Log, 01/02 - 06/05