ML050450129
| ML050450129 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/14/2005 |
| From: | Hay M NRC/RGN-IV/DRP/RPB-C |
| To: | Edington R Nebraska Public Power District (NPPD) |
| References | |
| IR-04-005 | |
| Download: ML050450129 (30) | |
See also: IR 05000298/2004005
Text
February 14, 2005
Randall K. Edington, Vice
President-Nuclear and CNO
Nebraska Public Power District
P.O. Box 98
Brownville, NE 68321
SUBJECT:
COOPER NUCLEAR STATION - NRC INTEGRATED INSPECTION
REPORT 05000298/2004005
Dear Mr. Edington:
On December 31, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Cooper Nuclear Station. The enclosed integrated inspection report
documents the inspection findings which were discussed on January 6, 2005, with
Mr. S. Minahan, General Manager of Plant Operations, and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, the NRC identified four findings that were evaluated
under the risk significance determination process as having very low safety significance
(Green). The NRC also determined that there were four violations associated with these
findings. However, because these violations were of very low safety significance and the issues
were entered into the licensees corrective action program, the NRC is treating these findings
as noncited violations (NCVs), consistent with Section VI.A.1 of the NRCs Enforcement Policy.
These NCVs are described in the subject inspection report. If you contest the subject or
significance of the NCVs, you should provide a response within 30 days of the date of this
inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional
Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive,
Suite 400, Arlington, Texas 76011-4005; the Director, Office of Enforcement, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the
Cooper Nuclear Station facility.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response, if any, will be made available electronically for public inspection
in the NRC Public Document Room or from the Publicly Available Records component of NRCs
document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Nebraska Public Power District
-2-
Should you have any questions concerning this inspection, we will be pleased to discuss them
with you.
Sincerely,
/RA/
Michael C. Hay, Chief
Project Branch C
Division of Reactor Projects
Docket: 50-298
License: DPR-46
Enclosure:
NRC Inspection Report 05000298/2004005
w/attachment: Supplemental Information
cc w/enclosure:
Michael T. Boyce, Nuclear Asset Manager
Nebraska Public Power District
1414 15th Street
Columbus, NE 68601
John C. McClure, Vice President
and General Counsel
Nebraska Public Power District
P.O. Box 499
Columbus, NE 68602-0499
P. V. Fleming, Licensing Manager
Nebraska Public Power District
P.O. Box 98
Brownville, NE 68321
Michael J. Linder, Director
Nebraska Department of
Environmental Quality
P.O. Box 98922
Lincoln, NE 68509-8922
Chairman
Nemaha County Board of Commissioners
Nemaha County Courthouse
1824 N Street
Auburn, NE 68305
Nebraska Public Power District
-3-
Sue Semerena, Section Administrator
Nebraska Health and Human Services System
Division of Public Health Assurance
Consumer Services Section
301 Centennial Mall, South
P.O. Box 95007
Lincoln, NE 68509-5007
Ronald A. Kucera, Deputy Director
for Public Policy
Department of Natural Resources
P.O. Box 176
Jefferson City, MO 65101
Jerry Uhlmann, Director
State Emergency Management Agency
P.O. Box 116
Jefferson City, MO 65102-0116
Chief, Radiation and Asbestos
Control Section
Kansas Department of Health
and Environment
Bureau of Air and Radiation
1000 SW Jackson, Suite 310
Topeka, KS 66612-1366
Daniel K. McGhee
Bureau of Radiological Health
Iowa Department of Public Health
401 SW 7th Street, Suite D
Des Moines, IA 50309
William J. Fehrman, President
and Chief Executive Officer
Nebraska Public Power District
1414 15th Street
Columbus, NE 68601
Jerry C. Roberts, Director of
Nuclear Safety Assurance
Nebraska Public Power District
P.O. Box 98
Brownville, NE 68321
Nebraska Public Power District
-4-
Chief Technological Services Branch
National Preparedness Division
Department of Homeland Security
Emergency Preparedness & Response Directorate
FEMA Region VII
2323 Grand Boulevard, Suite 900
Kansas City, MO 64108-2670
Nebraska Public Power District
-5-
Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
Senior Resident Inspector (SDC)
Branch Chief, DRP/C (KMK)
Senior Project Engineer, DRP/C (WCW)
Team Leader, DRP/TSS (RLN1)
RITS Coordinator (KEG)
RidsNrrDipmIipb
J. Dixon-Herrity, OEDO RIV Coordinator (JLD)
CNS Site Secretary (SLN)
NSIR/DPR/EPD (REK)
W. A. Maier, RSLO (WAM)
SISP Review Completed: _wcw___ ADAMS: / Yes
G No Initials: __wcw__
/ Publicly Available G Non-Publicly Available G Sensitive
/ Non-Sensitive
R:\\_CNS\\2004\\CN2004-05RP-SCS.wpd
RIV:RI:DRP/C
C:DRP/EB
C:DRS/OB
C:DRS/PSB
SDCochrum
SCSchwind
JAClark
ATGody
MPShannon
E - WCWalker
/RA/
/RA/
/RA/
/RA/
1/26/05
2/11/05
1/31/05
1/31/05
2/1/05
C:DRS/PEB
C:DRP/C
LJSmith
MCHay
/RA/
/RA/
1/29/05
2/14/05
OFFICIAL RECORD COPY
T=Telephone E=E-mail F=Fax
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket.:
50-298
License:
Report:
Licensee:
Nebraska Public Power District
Facility:
Cooper Nuclear Station
Location:
P.O. Box 98
Brownville, Nebraska
Dates:
September 24 through December 31, 2004
Inspectors:
S. Schwind, Senior Resident Inspector
S. Cochrum, Senior Resident Inspector (temporary)
D. Carter, Health Physicist
P. Elkmann, Emergency Preparedness Inspector
G. Replogle, Senior Reactor Inspector
Approved By:
M. Hay, Chief, Branch C, Division of Reactor Projects
Enclosure
SUMMARY OF FINDINGS
IR 05000298/2004005; 09/24/04 - 12/31/04; Cooper Nuclear Station, Fire Protection, ALARA
Planning and Controls and other activities.
The report covered a 3-month period of inspection by resident inspectors and region-based
inspectors. Four Green noncited violations and one Green finding were identified. The
significance of the issues is indicated by their color (Green, White, Yellow, or Red) and was
determined by the significance determination process in Inspection Manual Chapter 0609.
Findings for which the significance determination process does not apply are indicated by the
severity level of the applicable violation. The NRC's program for overseeing the safe operation
of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight
Process, Revision 3, dated July 2000.
A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. The inspectors identified a noncited violation of Technical Specification 5.4.1.d for failure to implement the stations fire watch procedure. Specifically, on
October 22, 2004, the inspectors identified that a compensatory fire watch,
responsible for protecting equipment important to safety from fire damage, was not
alert and therefore was inattentive to the areas assigned as directed by procedural
requirements.
This finding was considered more than minor since the finding would become a
more significant safety concern if left uncorrected. The finding was determined to
be of very low safety significance, since the finding was assigned a moderate fire
protection barrier degradation rating and did not degrade the automatic water-
based fire suppression system in the fire area. This finding had crosscutting
aspects associated with problem identification and resolution due to the licensees
failure to enter this condition into the corrective action program until prompted by
the inspectors approximately 10 days following its identification (Section 1R05).
Green. The inspectors identified a noncited violation of 10 CFR Part 50,
Appendix B, Criterion XVI, involving the failure to promptly identify and correct
conditions adverse to quality. Specifically, on numerous occasions the licensee
failed to promptly identify that environmental temperatures outside design
specifications could potentially affect the function of equipment important to
safety. As a result, the licensee failed to promptly evaluate this adverse condition
in a timely manner. The failure to promptly identify and correct this condition
adverse to quality involved crosscutting aspects associated with problem
identification and resolution.
The inspectors determined that the issue had more than minor safety significance
because it impacted the mitigating systems cornerstone objective and could have
affected the ability of safety-related systems to perform their design basis
functions. The finding was of very low risk significance because it was a
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Enclosure
design/qualification deficiency that did not result in a loss of function per Generic Letter 91-18, Information to Licensees Regarding NRC Inspection Manual Section
on Resolution of Degraded and Nonconforming Conditions, Revision 1
(Section 4OA5).
Cornerstone: Occupational Radiation Safety
Green. The inspectors identified a noncited violation of Technical Specification 5.7.1, since the licensee failed to barricade and conspicuously post a high
radiation area. On November 30, 2004, the inspector identified piping located in
the Residual Heat Removal B heat exchanger room that had dose rates elevated
to greater than 100 millirem per hour. The licensee performed a survey and
confirmed dose rates were 600 millirem per hour on contact with the pipe and
160 millirem per hour at 12 inches from the pipe. The area was immediately
barricaded and posted. The licensee entered this issue into its corrective action
program.
This finding is greater than minor because it was associated with the cornerstone
attribute (exposure control) and affected the cornerstone objective because failure
to post a high radiation area with dose rates greater than 100 millirem per hour
could increase the risk of personnel dosage. The finding was of very low safety
significance because it did not involve: (1) ALARA planning and controls, (2) an
overexposure, (3) a substantial potential for overexposure, or (4) an impaired
ability to assess dose (Section 2OS2).
Green. The inspector reviewed a self-revealing, noncited violation of Technical Specification 5.7.1 because the licensee failed to provide an individual a radiation
monitoring device that could be detected when a preset integrated dose alarm was
received. On December 15, 2003, an individual unknowingly exceeded the alarm
setpoint of a required electronic dosimeter while working in an area with radiation
levels as high as 200 millirem per hour. The electronic dosimeter was set to alarm
at 20 millirem, but upon exiting the area, the electronic dosimeter read 31 millirem
and was alarming. The individual did not hear the alarm until the area was exited.
The licensee entered this issue into its corrective action program.
This finding is greater than minor because it was associated with the cornerstone
attribute (exposure control) and affected the cornerstone objective because the
inability to detect an alarming device in a high radiation area could increase
personnel dose. The finding was of very low safety significance because it did not
involve: (1) ALARA planning and controls, (2) an overexposure, (3) a substantial
potential for overexposure, or (4) an impaired ability to assess dose. This finding
also had crosscutting aspects associated with human performance
(Section 2OS2).
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Enclosure
B.
Licensee-Identified Violations
Violations of very low safety significance, which were identified by the licensee, have
been reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. These violations and
corrective actions are listed in Section 4OA7 of this report.
Enclosure
REPORT DETAILS
The plant was operating at full power at the beginning of this inspection period. On October 19,
2004, the reactor was shut down due to elevated main turbine rotor vibrations. Following repair
of the main turbine on November 10, 2004, full power operations resumed for the rest of the
inspection period.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R01 Adverse Weather Protection (71111.01)
a.
Inspection Scope
The inspectors completed an inspection sample of licensee activities involving
preparations for cold weather conditions on two risk significant systems. These
activities included:
A review of maintenance work orders completed to prepare the systems for cold
weather conditions
A review of deficiency tags and condition reports associated with cold weather
protection measures to determine their impact on the systems
A walkdown of Emergency Diesel Generator (EDG) 2 to verify proper ventilation
alignments were implemented
A walkdown of the ventilation screens in the intake structure to verify that the
licensee had completed the required actions identified in the work orders
The two risk significant systems evaluated during this inspection included:
Portions of the EDG 2 system
The intake structure and environmental controls located in the service water pump
room
b.
Findings
No findings of significance were identified.
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Enclosure
1R04
Equipment Alignment (71111.04)
Partial Equipment Alignment Inspections
a.
Inspection Scope
The inspectors performed two partial equipment alignment inspections. The inspections
verified that critical portions of the selected systems were correctly aligned in
accordance with system operating procedures. The following two equipment alignment
inspections were performed:
EDG 1, while EDG 2 was inoperable during cleaning and coating of Diesel Fuel
Storage Tank 2 on October 25, 2004. The walkdown included accessible portions
of the system in the diesel generator room as well as temporary diesel fuel tanks,
hoses, and other equipment staged to support operability of EDG 1 during this
work. The inspectors also performed an as-found inspection of Diesel Fuel
Storage Tank 2 to assess its condition and an as-left inspection prior to refilling the
tank.
EDG 2, while EDG 1 was inoperable during cleaning and coating of the Diesel Fuel
Storage Tank 1 on November 1, 2004. The walkdown included accessible
portions of the system in the diesel generator room as well as temporary diesel
fuel tanks, hoses, and other equipment staged to support operability of EDG 2
during this work. The inspectors also performed an as-found inspection of Diesel
Fuel Storage Tank 1.
a.
Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
Quarterly Walkdowns
a. Inspection Scope
The inspectors performed six fire zone inspections to verify the licensee was maintaining
those areas in accordance with the fire hazards analysis. The fire zones were chosen
based on their risk significance as described in the individual plant examination of
external events. The walkdowns focused on control of combustible materials and
ignition sources, operability and material condition of fire detection and suppression
systems, and the material condition of passive fire protection features. The following
fire zones were inspected:
Fire Zone 1F/G, Control and computer rooms
Fire Zone 2A, Control rod mechanism North
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Enclosure
Fire Zone 3A/B, Critical switchgear room
Fire Zone 8D, Control Building Elevation 903
Fire Zone 8F/8G, Division 1 battery room and DC switchgear room
Fire Zone 20A/B, SW pump room
b.
Findings
Introduction. The inspectors identified a noncited violation (NCV) of Technical Specification (TS) 5.4.1.d for failure to implement the stations fire watch procedure.
Specifically, on October 22, 2004, the inspectors identified that a compensatory fire
watch, responsible for protecting equipment important to safety from fire damage, was
not alert and therefore was inattentive to the areas assigned as directed by procedural
requirements.
Description. On October 22, 2004, the inspectors conducted an inspection of the fire
protection features in the northeast section of the reactor building 903 level. The
inspectors identified that a compensatory fire watch, assigned to watch for fires in this
area, was not alert nor attentive to the area assigned. Following questioning by the
inspectors, the fire watch stated he was tired and therefore was not attentive to
assigned fire watch duties. The inspectors discussed the requirements of Administrative
Procedure 0.39, Fire Watches, Section 6.3.1, with the fire watch. The section states,
in part, that the fire watch shall observe the affected area and be alert for signs of fire,
smoke, and changing conditions. The inspectors then informed the shift manager and
outage manager who confirmed the fire watch appeared very tired and not alert and had
the watch relieved.
Analysis. The failure to implement the procedural requirements of Administrative
Procedure 0.39, Fire Watches, Revision 31, was considered a performance deficiency
which affected the mitigating systems cornerstone since compensatory fire watches are
used throughout the plant to protect safety-related equipment when fire protection
systems are degraded. This finding was considered more than minor since the finding
would become a more significant safety concern if left uncorrected. Inspection Manual
Chapter 0609, Significance Determination Process, Appendix F, was used to assess
the safety significance of this finding. Based on the results of a significance
determination process Phase 1 evaluation, the finding was determined to have very low
safety significance (Green) since the finding was assigned a moderate fire protection
barrier degradation rating and did not degrade the automatic water-based fire
suppression system in the fire area.
This finding had crosscutting aspects associated with problem identification and
resolution. This assessment was based on the fact that the licensee failed to enter this
condition into the corrective action program until prompted by the inspectors
approximately 10 days later.
Enforcement. TS 5.4.1.d states, Written procedures shall be established, implemented,
and maintained covering the fire protection program. Administrative Procedure 0.23,
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Enclosure
CNS Fire Protection Plan, Revision 41, Section 3.3, states, Fire Watches are
controlled by Procedure 0.39. Administrative Procedure 0.39, Fire Watches,
Revision 31, Section 6.3.1, states, in part, that the fire watch shall observe the affected
area and be alert for signs of fire, smoke, and changing conditions. Contrary to this
requirement, the fire watch failed to observe the affected area and remain alert to fire,
smoke, and changing conditions. Because this violation was of very low safety
significance and was entered into the corrective action program as Condition
Report CR-CNS-2004-07109, this violation is being treated as an NCV consistent with
Section VI.A of the NRC Enforcement Policy: NCV 05000298/2004005-01, Failure to
Implement the Station Fire Watch Procedure.
1R11 Licensed Operator Requalification (71111.11)
a.
Inspection Scope
On December 7, 2004, the inspectors observed one session of licensed operator
requalification training in the plant simulator. The training evaluated operator ability to
recognize, diagnose, and respond to a loss of dc power and a reactor scram.
Observations were focused on the following key attributes of operator performance:
Crew performance in terms of clarity and formality of communications
Ability to take timely and appropriate actions
Prioritizing, interpreting, and verifying alarms
Correct implementation of procedures, including the alarm response procedures
Timely control board operation and manipulation, including high-risk operator
actions
Oversight and direction provided by the shift supervisor, including the ability to
identify and implement appropriate TS requirements, reporting, emergency plan
actions, and notifications
Group dynamics involved in crew performance
The inspectors also verified that the simulator response during the training scenario
closely modeled expected plant response during an actual event.
b.
Findings
No findings of significance were identified.
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Enclosure
1R12
Maintenance Rule Implementation (71111.12)
a.
Inspection Scope
The inspectors reviewed one equipment performance issue to assess the licensees
implementation of their maintenance rule program. The inspectors verified that
components which experienced performance problems were properly included in the
scope of the licensees maintenance rule program and that the appropriate performance
criteria were established. Maintenance rule implementation was determined to be
adequate if it met the requirements outlined in 10 CFR 50.65 and Administrative
Procedure 0.27, Maintenance Rule Program, Revision 15. The inspectors reviewed
the following equipment performance problem:
Failure of breakers related to Lighting Panel EE-PNL- LPIS1 (CR 2004-07124)
b.
Findings
No findings of significance were identified
1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)
a.
Inspection Scope
The inspectors reviewed two risk assessments for planned or emergent maintenance
activities to determine if the licensee met the requirements of 10 CFR 50.65(a)(4) for
assessing and managing any increase in risk from these activities. Evaluations for the
following maintenance activities were included in the scope of this inspection:
Corrective maintenance on the reactor core isolation cooling system to replace a
fuse and fuse holder on September 30, 2004 (CR 2004-06582)
Surveillance procedure on Residual Heat Removal (RHR) Loop B requiring
shutdown cooling to be secured on November 4, 2004 (Work Order 4360509)
b.
Findings
No findings of significance were identified.
1R14 Personnel Performance During Nonroutine Evolutions (71111.14)
a.
Inspection Scope
For the nonroutine event described below, the inspectors reviewed operator logs, plant
computer data, and strip charts to determine what occurred, how the operators
responded, and whether the response was in accordance with plant procedures.
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Enclosure
On October 18, 2004, the inspectors responded to the control room following
identification of elevated main turbine bearing vibrations. Based on the elevated
vibrations, operators performed a normal shutdown to inspect the main turbine
blades. The inspectors observed and evaluated the reactor shutdown, followup
actions by the operators, actions required by procedures, and monitoring of plant
conditions.
b.
Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a.
Inspection Scope
The inspectors reviewed three operability determinations associated with mitigating
system capabilities to ensure that the licensee properly justified operability and that the
component or system remained available so that no unrecognized increase in risk
occurred. These reviews considered the technical adequacy of the licensees evaluation
and verified that the licensee considered other degraded conditions and their impact on
compensatory measures for the condition being evaluated. The inspectors referenced
the Updated Safety Analysis Report, TSs, and the associated system design criteria
documents to determine if operability was justified. The inspectors reviewed the
following equipment conditions and associated operability evaluations:
Reactor equipment cooling system leakage (CR 2004-06015)
Rod block monitor setpoint error (CR 2004-06893)
RHR Valve MO-13D failure to close (CR 2004-06776)
b.
Findings
No findings of significance were identified.
1R16
Operator Workarounds (71111.16)
a.
Inspection Scope
The inspectors performed a review of all open operator workaround items to evaluate
their cumulative effect on mitigating systems and the operators ability to implement
abnormal or emergency procedures. In addition, open operability determinations and
selected condition reports were reviewed, and operators were interviewed to determine
if there were additional degraded or nonconforming conditions that could complicate the
operation of plant equipment.
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Enclosure
b.
Findings
No findings of significance were identified.
1R17
Permanent Plant Modifications
a.
Inspection Scope
The inspectors reviewed plant modification Change Evaluation Document (CED)
6008140 that replaced the reactor feedwater startup valves.
b.
Findings
Introduction. A Green self-revealing finding was identified involving the failure to
perform an adequate design change for the reactor feed system startup flow control
valves. The inadequate design change failed to ensure component temperature ratings
were not exceeded resulting in adversely affecting valve operation. Specifically, the
licensees evaluation failed to recognize and address acceptable O-ring types for the
temperatures of the reactor feed system.
Description. In September and October of 2004, several reactor feed pump (RFP)
trouble alarms, related to startup flow control valve abnormal operation, were received in
the control room. These valves are used to control feed water flow to the reactor vessel
during startup, cooldown, and depressurization. Based on initial testing and operation
following the alarms, the licensee determined that failed valve positioners were causing
the alarms. Both valve positioners were replaced September 25, 2004, however, during
testing the valves did not cycle as expected. During the investigation into the cause of
valve positioner failures, the licensee discovered the valve piston o-ring seal was
degraded, failing to provide the required seal. As a result, the valves would operate
erratically and deviate from normal demand.
The licensees apparent cause investigation discovered that the actuator O-rings were
square in shape indicating they had taken a permanent set due to exceeding the O-ring
temperature rating. The vendor manual states the temperature limit for this type of
actuator with nitrile O-rings is 175EF. During normal operation, the reactor feedwater
temperatures flowing thought the startup flow control valves reach as high as 360EF
exceeding the nitrile O-ring 175EF limit. In October of 2002, the licensee implemented
Modification CED 6008140, installing the inappropriate O-rings for the reactor feed
system startup flow control Valves RF-AOV-FCV11AA and RF-AOV-FCV11BB.
Analysis. The inadequate design review of the RFP startup flow control valve
modification (CED 6008140) was considered a performance deficiency, which affected
the Mitigating Systems cornerstone. This finding is greater than minor because it
affected both the Initiating Events and Mitigating Systems cornerstone attribute of
design control, reducing the reliability and capability of the RFP startup flow control
valves to mitigate events or potentially result in an initiating event based on loss of
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Enclosure
feedwater control to the reactor vessel. This issue is unresolved for significance
determination and the appropriate regulatory characterization (URI 05000298/2004005-
02, Review Safety Significance or Degraded Startup Flow Control Valves).
Enforcement. No violation of regulatory requirements occurred because the RFP
startup flow control valves are not classified as safety-related. The licensee entered this
finding into their corrective action program as Condition Report CR-CNS-2004-06997.
This finding is identified as FIN 05000298/2004005-02, Inadequate Design Review of
System Modification.
1R19 Postmaintenance Testing (71111.19)
a.
Inspection Scope
The inspectors reviewed or observed four selected postmaintenance tests (four
inspection samples) to verify that the procedures adequately tested the safety
function(s) that were affected by maintenance activities on the associated systems. The
inspectors also verified that the acceptance criteria were consistent with information in
the applicable licensing basis and design basis documents and that the procedures
were properly reviewed and approved. Postmaintenance tests for the following
maintenance activities were included in the scope of this inspection:
Emergency Diesel Generator 2 fuel oil pump replacement (Work Order 4376678)
Emergency Diesel Generator 2 fuel oil strainer cleaning and inspection (Work Order 4384984)
High pressure coolant injection exhaust drip leg drain extension (Work Order 4327601)
RHR system Valve RHR-MO-13D control switch replacement (Work Order 4406741)
b.
Findings
No findings of significance were identified.
1R20
Refueling and Outage Activities
a.
Inspection Scope
The inspectors observed outage-related activities during a forced outage following a
main turbine blade failure on October 18, 2004. Activities included reactor shutdown,
plant cooldown, placing the RHR system in the shutdown cooling mode of operation,
and startup activities.
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Enclosure
b.
Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a.
Inspection Scope
The inspectors observed or reviewed the following four surveillance tests (four
inspection samples) to ensure that the systems were capable of performing their safety
function and to assess their operational readiness. Specifically, the inspectors verified
that the following surveillance tests met TS requirements, the Updated Safety Analysis
Report, and licensee procedural requirements:
6.HPCI.103, HPCI IST and 92 Day Test Mode Surveillance Operation,
Revision 26, performed on Oct 15, 2004
6.1DG.105, Diesel Generator Starting Air Compressor Operability (IST) (DIV 1),
Revision 13C1, performed on October 4, 2004
6.2RHR.706, RHR Loop B Injection Valve Time Delay Channel Functional Test
(DIV 2), Revision 2, performed on November 4, 2004
6.DWLD.302, Drywell Floor Drain Sump 1F Flow Loop Channel Calibration,
Revision 6, performed on December 14, 2004
b.
Findings
No findings of significance were identified.
Cornerstones: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a.
Inspection Scope
The inspector performed an in-office review of Revision 48 to the Cooper Nuclear
Station Emergency Plan, submitted October 20, 2004, and Revision 49 to the Cooper
Nuclear Station Emergency Plan, submitted October 25, 2004. This review included the
following changes that had been made to the plan:
Revised the physical description of installed meteorological instruments because
of replacement of the instruments, along with their associated ranges, tolerances,
and accuracy
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Enclosure
Revised the description of the emergency notification system from the Emergency
Broadcast System to the Emergency Alerting System, with associated details
related to system activation
Revised the location from which the emergency notification system is activated in
Atchinson County, Missouri, from the Sheriffs Department to the 911 Center
Revised the site emergency preparedness training guide from a guide document to
a procedure
Removed emergency preparedness from the site Quality Assurance for Operation
Policy Document
Revised the titles of several procedures.
The revision was compared to: its previous revision; the criteria of NUREG-0654,
Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and
Preparedness in Support of Nuclear Power Plants, Revision 1; the criteria of
ANSI/ANS-2.5-1984, American National Standard for Determining Meteorological
Information at Nuclear Power Sites; the criteria of Federal Emergency Management
Agency (FEMA) Report REP-10, Guide for the Evaluation of Alert and Notification
Systems for Nuclear Power Plants; and the requirements of 10 CFR 50.47(b) and
50.54(q) to determine if the revision decreased the effectiveness of the plan.
b.
Findings
No findings of significance were identified.
1EP6 Drill Evaluation (71114.06)
a.
Inspection Scope
The inspectors observed the licensee perform one emergency preparedness drill on
December 9, 2004. Observations were conducted in the control room, technical support
center, and emergency operations facility. During the drill, the inspectors assessed the
licensees performance related to classification, notification, and protective action
recommendations. Following the drill, the inspectors reviewed the licensees critique to
determine if issues were appropriately identified and documented. The following
documents were reviewed during this inspection:
Emergency Plan for Cooper Nuclear Station
Emergency Plan Implementing Procedures for Cooper Nuclear Station
Cooper Nuclear Station Emergency Preparedness Drill Scenario for December 9,
2004
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Enclosure
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS2 ALARA Planning and Controls (71121.02)
a.
Inspection Scope
The inspector assessed licensee performance with respect to maintaining individual and
collective radiation exposures as low as is reasonably achievable (ALARA). The
inspector used the requirements in 10 CFR Part 20 and the licensees procedures
required by TSs as criteria for determining compliance. The inspector interviewed
licensee personnel and reviewed:
Current 3-year rolling average collective exposure
Three on-line maintenance work activities scheduled during the inspection period
and associated work activity exposure estimates which were likely to result in the
highest personnel collective exposures.
Site-specific trends in collective exposures, plant historical data, and source-term
measurements
Site-specific ALARA procedures
Intended versus actual work activity doses and the reasons for any inconsistencies
Integration of ALARA requirements into work procedure and radiation work permit
(or radiation exposure permit) documents
Person-hour estimates provided by maintenance planning and other groups to the
radiation protection group with the actual work activity time requirements
Shielding requests and dose/benefit analyses
Method for adjusting exposure estimates, or replanning work, when unexpected
changes in scope or emergent work were encountered
Use of engineering controls to achieve dose reductions and dose reduction
benefits afforded by shielding
Workers use of the low dose waiting areas
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Enclosure
Exposures of individuals from selected work groups
Declared pregnant workers during the current assessment period, monitoring
controls, and the exposure results
The inspector completed 8 of the required 15 samples and 5 of the optional samples.
b.
Findings
(1)
Introduction. The inspector identified an NCV of TS 5.7.1 for failure to barricade and
conspicuously post a high radiation area.
Description. On November 30, 2004, during walkdowns of the Reactor Building 931-foot
elevation, the inspector performed independent radiation measurements and identified
dose rates greater than 100 millirem per hour coming from RHR B heat exchanger
piping. This area was not posted and barricaded as a high radiation area, and it was
accessible from a scaffold platform. The radiation survey tag attached to the scaffold
ladder indicated general area dose rates of approximately 20 to 35 millirem per hour.
The licensee performed a survey of the area and confirmed dose rates of 600 millirem
per hour on contact with the pipe and 160 millrem per hour at one foot from the pipe.
The area was immediately barricaded and posted as a high radiation area.
Analysis. The failure to barricade and post a high radiation area is a performance
deficiency. This NRC-identified finding is greater than minor because it was associated
with a cornerstone attribute (exposure control), and affected the associated cornerstone
objective, to ensure the adequate protection of workers health and safety from
exposure to radiation, because not properly controlling high radiation areas could
increase personnel dose. The finding involved the potential for an individuals
unplanned or unintended dose, which could have been significantly greater as a result of
a single minor reasonable alteration of the circumstances. When processed through the
Occupational Radiation Safety Significant Determination Process, the finding was of
very low safety significance because it did not involve: (1) ALARA planning and
controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an
impaired ability to assess dose.
Enforcement. TS 5.7.1 requires that each area in which radiation levels are in excess of
100 millirem per hour but less than 1000 millirem per hour shall be barricaded and
conspicuously posted as a high radiation area and entrance thereto shall be controlled
by requiring issuance of a specific work request. Contrary to this, the licensee failed to
barricade and post a high radiation area. Because the failure to barricade and post a
high radiation area was determined to be of very low safety significance and has been
entered into the licensees corrective action program as Condition Report CR-CNS-
2004-07496, this violation is being treated as an NCV consistent with Section VI.A of the
NRC Enforcement Policy: NCV 05000298/2004005-03, Failure to Barricade and Post a
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Enclosure
(2)
Introduction. The inspector reviewed a self-revealing NCV of TS 5.7.1 for failure to
provide an individual a radiation monitoring device that could be detected when a preset
integrated dose alarm was received in a high radiation area.
Description. On December 15, 2003, an individual unknowingly exceeded the alarm
setpoint of a required radiation monitoring device (electronic dosimeter) while working in
an area with radiation levels as high as 200 millirem per hour. A worker exited the high
radiation area (the condenser area) with the electronic dosimeter in alarm. The workers
electronic dosimeter alarm was set at 20 millirem; upon exiting the area the electronic
dosimeter was reading 31 millirem. The worker did not hear the electronic dosimeter
alarm until the area was exited, and the alarm became self-revealing to the worker.
In addition to being unable to hear the electronic dosimeter alarm, the licensees
apparent cause determination identified that: (1) the worker failed to properly monitor his
dose during work, (2) the worker and the radiation protection technician failed to
communicate a specific stay time for a job, and (3) the radiation protection technician
failed to ensure the proper electronic dosimeter was used for the job, as required by the
radiation work permit.
Analysis. The failure to provide a radiation monitoring device that could be detected
when it alarms in a high radiation area is a performance deficiency. This self-revealing
finding is greater than minor because it was associated with a cornerstone attribute
(exposure monitoring), and affected the associated cornerstone objective, to ensure the
adequate protection of the workers health and safety from exposure to radiation,
because being unable to detect a radiation alarming device in a high radiation area
could increase personnel dose. The finding involved an individuals unplanned or
unintended dose. When processed through the Occupational Radiation Safety
Significant Determination Process, the finding was of very low safety significance
because it did not involve: (1) ALARA planning and controls, (2) an overexposure, (3) a
substantial potential for overexposure, or (4) an impaired ability to assess dose. The
finding also had crosscutting aspects associated with human performance.
Enforcement. TS 5.7.1 requires that any individual entering an area in which radiation
levels are in excess of 100 millirem per hour but less than 1000 millirem per hour shall
be provided with a monitoring device which continuously integrates the radiation dose in
the area and alarms when a preset integrated dose is received. Contrary to this, the
licensee failed to provide a monitoring device that can be detected when it alarms.
Because the failure to provide a radiation monitoring device that could be detected when
it alarms was determined to be of very low safety significance and has been entered into
the licensees corrective action program as Condition Report CR-CNS-2003-02009, this
violation is being treated as an NCV consistent with Section VI.A of the NRC
Enforcement Policy: NCV 05000298/2004005-04, Failure to Provide a Monitoring
Device that could Detect High Radiation in a Work Area.
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Enclosure
4.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification (71151)
a.
Inspection Scope
The inspectors sampled two licensee PIs listed below for the period October 2003
through September 2004 (two inspection samples). The definitions and guidance of
Nuclear Energy Institute 99-02, Regulatory Assessment Indicator Guideline,
Revision 2, were used to verify that the licensee accurately reported PI data during the
assessment period. Licensee PI data was reviewed against the requirements of
Procedure 0-PI-01, Performance Indicator Program, Revision 16.
Reactor Safety Strategic Area
Reactor Coolant System Specific Activity
Reactor Coolant System Leak Rate
The inspectors reviewed a selection of licensee event reports, portions of operator log
entries, monthly reports, and PI data sheets to determine whether the licensee
adequately collected, evaluated, and distributed PI data for the period reviewed. The
inspectors also interviewed licensee personnel responsible for collecting and evaluating
PI data.
b.
Findings
No findings of significance were identified
4OA2 Identification and Resolution of Problems (71152)
.1
Routine Review of Identification and Resolution of Problems
a.
Inspection Scope
The inspectors reviewed a selection of condition reports written during the inspection
period to verify the licensee was entering conditions adverse to quality into the
corrective action program at an appropriate threshold. Additionally, the inspectors
verified that condition reports were appropriately categorized and dispositioned in
accordance with the licensees procedures, and in the case of significant conditions
adverse to quality, to review the adequacy of licensee root cause determinations, extent
of condition reviews, and implemented corrective actions.
b.
Findings
No findings of significance were identified.
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Enclosure
.2
Occupational Radiation Safety Sample Review
a.
Inspection Scope
The inspectors evaluated the effectiveness of the licensee's problem identification and
resolution processes regarding exposure tracking, higher than planned exposure levels,
and radiation worker practices. The inspector reviewed the corrective action documents
listed in the attachment against the licensees problem identification and resolution
program requirements.
b.
Findings
No findings of significance were identified.
.3
Semiannual Trend Review
a.
Inspection Scope
The inspectors performed a semiannual assessment of trends in the licensees
corrective action program to determine if any more significant safety issues exist.
Specifically, the inspectors reviewed the licensees corrective action program database
to determine if the licensee had identified trends in any of the following areas:
RFP controller alarms
Reactor equipment cooling leakage results
Intake structure silting
Contractor control
Average power range monitor alarms
Ronan computer multiplexer failures
345kv transformer alarms
120v ac lighting breaker tripping
Service air compressor failures
Procedure adherence
These areas were chosen based on information gathered by the inspectors during the
previous 6 months. For those areas where trends were documented in the corrective
action program, the inspectors verified that the licensee had corrective actions planned
or in place to address the trend. For the remainder of the issues in the scope of this
inspection, the inspectors reviewed control room logs, system health reports, Quality
Assurance Audits, and department self-assessments and interviewed selected licensee
staff to determine if any adverse trends existed.
-16-
Enclosure
b.
Findings
The inspectors concluded that, in general, the licensee had adequately identified trends
in areas within the scope of this inspection; however, these trends were not always
explicitly documented in the corrective action program. This was a result of the
licensee's practice of closing new condition reports regarding similar equipment issues
to an existing condition report which was already open to evaluate the condition. In all
cases, the licensee was taking adequate corrective actions to address the trends.
There were six condition reports written during this 6-month period regarding lighting
panel breaker tripping related to Panel EE-PNL-LPIS1, which did not cross any
statistical thresholds in the licensee's trending program. The inspectors identified
several additional lighting breaker trips that were not documented in the correction
action program. During a discussion of these events, engineering personnel concluded
that a trend may exist. Condition Report CR-CNS-2004-07124 was written to document
a potential trend regarding breaker trips related to lighting Panel EE-PNL-LPIS1.
.4
Identification and Resolution of Problems Crosscutting Aspects of Findings
Sections 1R05 and 4OA5 describe findings with crosscutting aspects associated with
problem identification and resolution.
4OA4 Human Performance Crosscutting Aspects of Findings
Sections 1R17 and 2OS2 describe findings with crosscutting aspects associated with
human performance.
4OA5 Other Activities
(Closed) Unresolved Item 05000298/2004004-05: Plant temperatures outside Updated
Safety Analysis Report Limits
Introduction. The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B,
Criterion XVI, involving the failure to promptly identify and correct conditions adverse to
quality. Specifically, on numerous occasions the licensee failed to promptly identify that
environmental temperatures outside design specifications could potentially affect the
function of equipment important to safety. As a result, the licensee failed to promptly
evaluate this adverse condition in a timely manner. The failure to promptly identify and
correct this condition adverse to quality involved crosscutting aspects associated with
problem identification and resolution.
Description. The Cooper Updated Safety Analysis Report states, in part, that the design
of the station heating, ventilating, and air conditioning systems are based on a minimum
outdoor temperature of -5EF and a maximum outdoor temperature of 97EF. The
inspectors reviewed historical plant temperature data and noted that in the last 2 years
site temperatures had exceeded the 97 degree limit on 12 occasions and had dropped
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Enclosure
below the -5EF threshold five times. During this same period plant temperatures were
as high as 104EF and as low as -10EF. Plant areas that directly rely on outside air for
temperature control during design basis accidents include the emergency diesel
generator rooms, portions of the reactor building, the control room, and the control
building.
The inspectors determined that engineering personnel failed to follow corrective action
program requirements resulting in the failure to promptly identify and correct a condition
adverse to quality. Specifically, although engineering was aware of the deviations from
the design temperature specifications, this nonconforming condition was not entered
into the corrective action program for resolution, resulting in the licensees failure to
evaluate the potential adverse affect to equipment being subjected to temperatures
outside design values.
Analysis. The inspectors determined that the issue had more than minor safety
significance because it impacted the mitigating systems cornerstone objective and could
have affected the ability of safety-related plant systems to perform their design basis
functions. The finding was of very low risk significance because it was a
design/qualification deficiency that did not result in a loss of function per Generic Letter 91-18, Information to Licensees Regarding NRC Inspection Manual Section on
Resolution of Degraded and Nonconforming Conditions, Revision 1.
This finding had crosscutting aspects associated with problem identification and
resolution. This assessment was based on the fact that the licensee failed to enter this
known nonconforming condition into the corrective action program.
Enforcement. The inspectors identified a violation of 10 CFR Part 50, Appendix B,
Criterion XVI (Corrective Actions), because the licensee had failed to properly identify
conditions adverse to quality. The noted regulation requires licensees, in part, to
promptly identify and correct conditions adverse to quality. Contrary to this requirement,
the inspectors identified 17 instances, in the last 2 years, where outside ambient
temperatures had exceeded design specifications for safety-related heating and
ventilation systems (conditions adverse to quality) and the licensee had failed to address
or evaluate the occurrences. Because the violation was of very low safety significance,
and was entered into the licensee's corrective action program (Condition Report 2004-
6820), this violation is being treated as an NCV, consistent with Section VI.A of the NRC
Enforcement Policy (NCV 05000298/2004005-05).
4OA6 Meetings, Including Exit
On December 2, 2004, the inspectors presented the ALARA Planning and Controls
inspection results to Mr. S. Minahan, General Manager of Plant Operations, and other
members of his staff who acknowledged the findings.
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Enclosure
On December 13, 2004, the inspectors conducted a telephonic exit meeting of the
emergency preparedness inspection results to Mr. J. Bednar, Emergency Preparedness
Manager, and other members of his staff who acknowledged the findings.
On January 6, 2005, the inspectors presented the results of the resident inspector
activities to Mr. S. Minahan, and other members of his staff, who acknowledged the
findings.
The inspectors confirmed that proprietary information was not provided or examined
during the inspection.
40A7 Licensee-Identified Violations
The following violation of very low significance (Green) was identified by the licensee
and is a violation of NRC requirements which meet the criteria of Section VI of the
NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
Cornerstone: Barrier Integrity
TS 5.6.5(a)3 required the licensee to determine the rod block monitor upscale
allowable values and document the values in the core operating limits report
prior to each reload cycle. Contrary to this requirement, this determination was
not completed prior to Reload Cycle 22. The licensee used generic values
provided by the vendor vice determining cycle-specific values. This resulted in
nonconservative rod block monitor upscale setpoints. This finding affected the
Barrier Integrity cornerstone and was of very low safety significance since it did
not represent an actual degradation of a fission product barrier. This was
identified in the licensees corrective action program as Condition Report CR-
ATTACHMENT: SUPPLEMENTAL INFORMATION
A-1
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
J. Bednar, Emergency Preparedness Manager
C. Blair, Engineer, Licensing
D. Cook, Technical Assistant to General Manager
J. Christensen, Co-Director of Nuclear Safety Assurance
S. Minahan, General Manager of Plant Operations
T. Chard, Radiological Manager
K. Chambliss, Operations Manager
K. Dalhberg, General Manager of Support
J. Edom, Risk Management
R. Estrada, Corrective Actions Manager
J. Flaherty, Site Regulatory Liaison
P. Fleming, Licensing Manager
D. Knox, Maintenance Manager
W. Macecevic, Work Control Manager
J. Roberts, Director, Nuclear Safety Assurance
R. Shaw, Shift Manager
J. Sumpter, Senior Staff Engineer, Licensing
K. Tanner, Shift Supervisor, Radiation Protection
R. Hayden, Emergency Preparedness Staff
NRC Personnel
L. Ricketson, Senior Health Physicist
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Review Safety Significance of Degraded Startup Flow
Control Valves (Section 1R17)
Opened and Closed
Failure to Implement the Station Fire Watch Procedure
(Section 1R05)05000298/2004005-03
Failure to barricade and conspicuously post a high radiation
area (Section 2OS2)
A-2
Attachment
Failure to provide a radiation monitoring device that could
detect high radiation in a work area (Section 2OS2)05000298/2004005-05
Plant Temperatures Outside Design Specifications
(Section 4A05)
Closed
Plant temperatures outside Updated Safety Analysis Report
limits (Section 4A05)
LIST OF DOCUMENTS REVIEWED
Condition Reports
2004-06045
2004-06015
2004-07124
2004-06567
2004-06582
2004-06757
2004-06835
2004-07120
Section 2OS2: ALARA Planning and Controls (71121.02)
Corrective Action Documents
2003-1787, 2003-61136, 2003-7706, 2004-1699, 2004-5224, 2004-5969, 2004-5970, 2004-
5973, 2004-6327, and 2004-6360 and 10299462 and 10314221
Audits and Self-Assessments
Snap Shot Assessment dated July 19-23, 2004
Snap Shot Assessment dated May 24, 2004
Snap Shot Assessment dated June 2, 2004
Radiation Work Permits
2003-1111
Condenser tube leak and repair
2004-1047
Condenser water box cleaning
2004-1050
Fan B cooling unit bearing replacement
Procedures
0.ALARA.1
0.ALARA.2
ALARA Organization and Management, Revision 7
9.ALARA.4
Radiation Work Permits, Revision 4
9.RADOP.3
Area Posting and Access Control, Revision 16
9.ALARA.1
Radiation Protection at CNS, Revision 4
9.ALARA.5
ALARA Planning and Controls, Revision 12
9ALARA.12
Hot Spot Reduction Program, Revision 0
9.EP 3.14
Temporary Shielding, Revision 15
ALARA Committee Meeting Minutes
2003
February 19, 2003
June 6, 2003, ALARA Special Committee Meeting Minutes
1st Quarter 2003
3rd Quarter 2003
September 11, 2003, Water Box Cleaning
4th Quarter 2003
2004
2004-02, 2004-03, 2004-04, 2004-05, 2004-06, 2004-07, 2004-08, 2004-09, 2004-10, 2004-11,
2004-12, and 2004-13
Miscellaneous
2003 ALARA Program and RE21 Review
LIST OF ACRONYMS
as low as is reasonably achievable
CED
change evaluation document
CFR
Code of Federal Regulations
finding
noncited violation
NRC
U.S. Nuclear Regulatory Commission
performance indicator
reactor feed pump
TS
Technical Specification
A-3
Attachment