ML053090002
| ML053090002 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 11/05/2005 |
| From: | William Jones NRC/RGN-IV/DRP/RPB-B |
| To: | Naslund C Union Electric Co |
| References | |
| IR-05-004 | |
| Download: ML053090002 (7) | |
See also: IR 05000483/2005004
Text
November 5, 2005
Charles D. Naslund, Senior Vice
President and Chief Nuclear Officer
Union Electric Company
P.O. Box 620
Fulton, MO 65251
SUBJECT:
CALLAWAY PLANT - NRC INTEGRATED INSPECTION
REPORT 05000483/2005004
Dear Mr. Naslund:
On September 23, 2005, the NRC completed an inspection at your Callaway Plant. The
enclosed report documents the inspection findings which were discussed on September 26,
2005, with you 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.
Within these areas, the inspection consisted of selected examination of procedures and
representative records, observations of activities, and interviews with personnel.
This report documents two findings that were evaluated under the risk significance
determination process as having very low safety significance (Green). The NRC has
determined that violations are associated with these issues. These violations are being treated
as noncited violations (NCVs), consistent with Section VI.A of the Enforcement Policy. In
addition, an apparent violation was identified for the failure to adequately implement a
procedure for cold overpressure mitigation configurations control. The NRC is performing a
significant determination process Phase 3 review to determine the safety significance. The
NCVs and AV are described in the subject inspection report. If you contest these violations or
significance of these NCVs, you should provide a response within 30 days of the date of this
inspection report, with the basis for your denial, to the 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; the Director, Office of Enforcement, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the
Callaway Plant 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).
Union Electric Company
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Should you have any questions concerning this inspection, we will be pleased to discuss them
with you.
Sincerely,
/RA/
William B. Jones, Chief
Project Branch B
Division of Reactor Projects
Docket: 50-483
License: NPF-30
Enclosure:
NRC Inspection Report
05000483/2005004`
w/attachment: Supplemental Information
cc w/enclosure
Professional Nuclear Consulting, Inc.
19041 Raines Drive
Derwood, MD 20855
John ONeill, Esq.
Shaw, Pittman, Potts & Trowbridge
2300 N. Street, N.W.
Washington, DC 20037
Mark A. Reidmeyer, Regional
Regulatory Affairs Supervisor
Regulatory Affairs
AmerenUE
P.O. Box 620
Fulton, MO 65251
Missouri Public Service Commission
Governors Office Building
200 Madison Street
P.O. Box 360
Jefferson City, MO 65102
Mike Wells, Deputy Director
Missouri Department of Natural Resources
P.O. Box 176
Jefferson City, MO 65102
Rick A. Muench, President and
Chief Executive Officer
Wolf Creek Nuclear Operating Corporation
P.O. Box 411
Burlington, KS 66839
Dan I. Bolef, President
Kay Drey, Representative
Board of Directors Coalition
for the Environment
6267 Delmar Boulevard
University City, MO 63130
Les H. Kanuckel, Manager
Quality Assurance
AmerenUE
P.O. Box 620
Fulton, MO 65251
Director, Missouri State Emergency
Management Agency
P.O. Box 116
Jefferson City, MO 65102-0116
Scott Clardy, Director
Section for Environmental Public Health
P.O. Box 570
Jefferson City, MO 65102-0570
Union Electric Company
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Keith D. Young, Manager
Regulatory Affairs
AmerenUE
P.O. Box 620
Fulton, MO 65251
David E. Shafer
Superintendent, Licensing
Regulatory Affairs
AmerenUE
P.O. Box 66149, MC 470
St. Louis, MO 63166-6149
Certrec Corporation
4200 South Hulen, Suite 630
Fort Worth, TX 76109
Union Electric Company
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Electronic distribution by RIV:
Regional Administrator (BSM1)
DRP Director (ATH)
DRS Director (DDC)
DRS Deputy Director (RJC1)
Senior Resident Inspector (MSP)
Branch Chief, DRP/B (WBJ)
Senior Project Engineer, DRP/B (RAK1)
Team Leader, DRP/TSS (RLN1)
RITS Coordinator (KEG)
Only inspection reports to the following:
J. Dixon-Herrity, OEDO RIV Coordinator (JLD)
ROPreports
CWY Site Secretary (DVY)
SISP Review Completed: __WBJ__
ADAMS: / Yes
G No Initials: __WBJ___
/ Publicly Available G Non-Publicly Available G Sensitive / Non-Sensitive
R:\\_CW\\2005\\CW2005-04RP-MSP.wpd
RI:DRP/B
SRI:DRP/B
C:DRS/PEB
C:DRS/EB
DEDumbacher
MSPeck
LJSmith
CJPaulk
E-WBJones
E-WBJones
/RA/
/RA/
11/02/05
11/02/05
11/01/05
11/01/05
C:DRS/PSB
C:DRS/OB
C:DRP/B
MPShannon
ATGody
WBJones
ATGody for
/RA/
/RA/
11/02/05
11/02/05
11/05/05
OFFICIAL RECORD COPY
T=Telephone E=E-mail F=Fax
Enclosure
ENCLOSURE
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket:
50-483
License:
Report:
Licensee:
Union Electric Company
Facility:
Callaway Plant
Location:
Junction Highway CC and Highway O
Fulton, Missouri
Dates:
June 24 through September 23, 2005
Inspectors:
M. S. Peck, Senior Resident Inspector
D. E. Dumbacher, Resident Inspector
E. A. Owen, Reactor Inspector
G. A. Pick, Senior Reactor Inspector, Engineering Branch 2
Approved By:
W. B. Jones, Chief, Project Branch B
Enclosure
SUMMARY OF FINDINGS
IR 05000483/2005004; 06/24 - 9/23/2005; Callaway Plant: Personnel Performance During
Nonroutine Plant Evolutions, Operability Evaluations, and Identification and Resolution of
Problems
This report covered a 3-month inspection by region based reactor inspectors and resident
inspectors. Two Green noncited violations and an apparent violation were identified. The
significance of most findings is indicated by their color (Green, White, Yellow, or Red) using
Inspection Manual Chapter 0609, Significance Determination Process. Findings for which the
significance determination process does not apply may be Green or assigned a severity level
after NRC management review. 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.
Inspector-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. A self-revealing noncited violation of Technical Specification 5.4.1.a,
Procedures, was identified after AmerenUE failed to properly align the turbine driven
auxiliary feedwater pump mechanical overspeed trip mechanism after surveillance
testing. The trip mechanism was misaligned from August 1 - 18, 2005. The misaligned
trip mechanism increased the probability the turbine would trip if the pump would have
been required to respond to an event. This issue was entered into the corrective action
program as Callaway Action Request 200505801. This finding, which involved the
failure of an operator to follow procedure, was associated with the crosscutting area of
human performance.
This finding is greater than minor because the degraded trip mechanism affected the
reactor mitigating systems cornerstone and the equipment performance attribute to
ensure availability of systems that respond to prevent core damage. This finding is only
of very low safety significance because the condition was not a design or qualification
deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result
in an actual loss of safety function of a system; did not increase the likelihood of a fire;
and did not screen as potentially risk significant due to a seismic, flooding, or severe
weather initiating event (Section 1R15).
Cornerstone: Barrier Integrity
TBD. A self-revealing apparent violation of Technical Specification 5.4.1.a,
Procedures, was identified after an operator error resulted in the failure to maintain the
required cold overpressure mitigation system configuration while the reactor was in
Mode 5. Technical Specification 3.4.12, Cold Overpressure Mitigation System,
prohibited more than one centrifugal charging pump from being capable of injecting into
the reactor vessel. An operator inadvertently defeated administrative controls and
enabled a centrifugal charging pump during a diesel generator and sequencer test
restoration lineup on September 20, 2005. Contributing causes to the event were
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Enclosure
inadequate procedural controls and pre-job brief. This issue was entered into the
corrective action program as Callaway Action Request 200507092. This finding, which
involved the failure of an operator to follow procedure, was associated with the
crosscutting area of human performance.
This finding is greater than minor because, if left uncorrected, it would have become a
more significant safety concern involving the integrity of the reactor coolant system
boundary (barrier integrity cornerstone). The finding was evaluated using Manual
Chapter 0609, "Significance Determination Process," Appendix G, Shutdown Operations
Significance, Checklist 2. Although the performance deficiency did not result in a
Technical Specification violation, discussions with the Office of Nuclear Reactor
Regulation identified a Phase 3 analysis should be performed and is currently under
evaluation (Section 1R14).
Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,
"Corrective Action," after ineffective corrective actions resulted in a repeat degradation
of a control building emergency ventilation habitability boundary door. AmerenUEs
work control organization twice authorized work on the essential switchgear room to
emergency diesel generator room door without approval of the shift operations
department. As a result, shift operations did not understand that the habitability
boundary had been compromised by the maintenance. This finding, which involved
ineffective corrective actions to prevent the repeat degradation of the ventilation system
habitability boundary door, was associated with the crosscutting area of problem
identification and resolution.
This finding was greater than minor because it was associated with the integrity of the
control building pressure envelope in that the degraded door would not meet its
habitability function. The finding was only of very low safety significance because the
finding only represented a degradation of the radiological barrier function provided for
the control room (Section 4OA2).
Enclosure
REPORT DETAILS
Summary of Plant Status: At the beginning of the inspection period, the Callaway Plant was
operating at full power. AmerenUE incurred an unplanned power reduction to 65 percent on
June 28, 2005 following a main feed pump control failure. AmerenUE completed control
system repairs and returned the plant to full power on June 29, 2005. AmerenUE operated the
plant at full power until performing a normal shutdown on September 17, 2005, for refueling
Outage 14. The plant remained shutdown the remainder 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 reviewed AmerenUEs site preparation and actions for actual extreme
hot weather conditions (one inspection sample). The inspectors performed a detailed
review of the stations adverse weather Procedures EIP-ZZ-00231, Response to
Severe Thunderstorms/High Winds/Tornado Watches and Warnings, and
OTN-EF-00001, Essential Service Water System, Section 5.10, Using Essential
Service Water (ESW) to Reduce Containment Temperatures. The inspectors also
performed a review of the containment air temperature controls and the ultimate heat
sink temperature controls to verify that AmerenUE properly implemented required
administrative and Technical Specification (TS) controls. The inspectors performed the
walkdown on July 25, 2005, when air temperatures reached 104 degrees.
b.
Findings
No findings of significance were identified.
1R04
Equipment Alignment (71111.04)
a.
Inspection Scope
Partial System Walkdowns. The inspectors completed three partial system walkdowns
during the inspection period (three inspection samples). The inspectors performed the
walkdowns to verify component alignment and subsystem operability. The inspectors
used the Final Safety Analysis Report (FSAR), TSs, and the procedures and drawings
listed in the attachment as the bases for acceptability. The following systems were
included in the scope of this inspection:
Residual heat removal (RHR) system, Train A while the redundant train was out
of service for scheduled testing. The inspectors walked down components
located in the auxiliary and control buildings on August 31 and September 1,
2005.
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Enclosure
Containment cooling system, Train B on September 8 and 12, 2005. The
inspectors walked down components in the auxiliary building and containment
including the cooling coils while at power.
Charging and safety injection systems cold overpressure mitigation protection
alignments on September 19, 2005.
b.
Findings
No findings of significance were identified.
1R05
Fire Protection (71111.05)
Routine Fire Inspection Walkdowns
a.
Inspection Scope
The inspectors performed twelve fire zone walkdowns to verify that AmerenUE
maintained plant areas in accordance with the Fire Hazards Analysis Report (twelve
inspection samples). 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 Area A-7, Boron injection room, June 30, 2005
Fire Area F-1, Fuel building general areas, July 11, 2005
Fire Area A-4, Train A emergency core cooling room, July 15, 2005
Fire Area A-2, Train B emergency core cooling room, July 15, 2005
Fire Area A-23, Containment isolation valve room (north), July 15, 2005
Fire Area A-24, Containment isolation valve room (south), July 15, 2005
Fire Area C-28, Service area, September 20, 2005
Fire Area C-29, Secondary alarm station, September 20, 2005
Fire Area C-14, Class 1E air conditioning equipment room, September 20, 2005
Fire Area C-11, South cable chases, September 21, 2005
Fire Area C-12, North cable chases, September 21, 2005
Fire Area C-31, North vertical cable chase, September 23, 2005
b.
Findings
No findings of significance were identified.
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Enclosure
1R06
Flood Protection Measures (71111.06)
a.
Inspection Scope
The inspectors completed one external flood protection walkdown during the inspection
period (one inspection sample). The inspectors walked down both 1974 foot elevation
auxiliary building and control building sumps, plant exterior at grade level, and the
ESW pump houses on September 16 and 19, 2005. The inspectors performed the
walkdown to review plant configuration for susceptibility to external flooding, such as
that caused by heavy rains or flash flooding. The inspectors conducted the walkdowns
to verify that AmerenUE had implemented adequate protection for equipment below the
postulated flood line, including electrical conduits, holes, and wall penetrations. The
inspection included common drains, sumps, sump pumps, level alarms, and control
circuits. The inspectors also reviewed Callaway Action
,
Evaluate Ground Water Recovery System, and engineering evaluation RFR 15336E to
ensure an adequate ground water recovery system was being maintained. The
inspectors used FSAR Section 2.4, Hydrological Engineering, as the basis for
acceptability of the observed plant configuration.
b.
Findings
No findings of significance were identified.
1R07
Biennial Heat Sink Performance (71111.07B)
a.
Inspection Scope
The inspectors reviewed design documents (e.g., calculations and performance
specifications), program documents, implementing documents (e.g., test and
maintenance procedures), and corrective action documents. The inspectors interviewed
chemistry personnel, maintenance personnel, engineers, and program managers.
For heat exchangers directly connected to the safety-related service water system, the
inspectors verified whether testing, inspection and maintenance, or the biotic fouling
monitoring program provided sufficient controls to ensure proper heat transfer.
Specifically, the inspectors reviewed: (1) heat exchanger test methods and test results
from performance testing, (2) if necessary, heat exchanger inspection and cleaning
methods and results, and (3) chemical treatments for microfouling and controls for
macrofouling.
For heat exchangers directly or indirectly connected to the safety-related service water
system, the inspectors verified the: (1) condition and operation consistent with design
assumptions in the heat transfer calculations, (2) potential for water hammer, as
applicable, (3) vibration monitoring controls for the heat exchangers, (4) chemistry
controls for heat exchangers indirectly connected to the safety-related service water
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Enclosure
system, and (5) redundant and infrequently used heat exchangers are flow tested
periodically at maximum design flow.
For the ultimate heat sink and its subcomponents, the inspectors reviewed the following
requirements: (1) capacity of the reservoir, (2) macrofouling controls, (3) biotic fouling
controls, (4) controls for ensuring functionality during adverse conditions,
and (5) performance tests for pumps and valves.
If available, the inspectors reviewed additional nondestructive examination results for
the selected heat exchangers that demonstrated structural integrity.
The inspectors selected heat exchangers that ranked high in the plant specific risk
assessment and were directly or indirectly connected to the safety-related service water
system. The inspectors selected the following specific heat exchangers:
P645732, EEG01A Component Cooling Water Heat Exchanger A, performed on
October 23, 2002
P660945, KKJ01B Diesel Generator B coolers, performed on August 14, 2003
P700637, SGN01B Containment Cooler B, performed on June 9, 2004
P715273, SGL15A Penetration Room Cooler A, performed on August 25, 2004
The inspectors reviewed four samples and completed three of the required two to
three samples.
b.
Findings
No findings of significance were identified.
1R11
Licensed Operator Requalification Activities Review by Resident Staff (71111.11Q)
a.
Inspection Scope
The inspectors observed one licensed operator simulator training
and critique
(one inspection sample). The inspectors observed
to assess operator
performance during high-risk operator actions, implementation of the site emergency
plan, and industry operating experience (OE). The inspectors observed licensed
operators when they responded to a steam generator tube rupture, with the loss of
onsite power, training scenario on August 10, 2005.
b.
Findings
No findings of significance were identified.
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Enclosure
1R12
Maintenance Effectiveness (71111.12Q)
a.
Inspection Scope
The inspectors reviewed two samples of equipment maintenance problems (two
inspection samples). The inspectors performed the review to verify that AmerenUE
effectively implemented 10 CFR 50.65, Requirements for Monitoring the Effectiveness
of Maintenance at Nuclear Power Plants. The inspectors focused on maintenance rule
characterization of failed components, risk significance, determination of the
(a)(1) classification, corrective actions, and the appropriateness of performance goals
and monitoring criteria. The inspectors also evaluated emergent equipment issues to
determine if problems were identified at the appropriate level and entered into the
corrective action program. The inspectors used Administrative
Procedure EDP-ZZ-01128, Maintenance Rule Program, Revision 6, during the review.
The inspectors performed an in-office review of the following Maintenance Rule (a)(1)
evaluations:
CARs 200505224 and 200506294, DSK33021 Diesel B door found not latched
CAR 200504286, Failure of volume control tank level Transmitter BG-LT-0149
b.
Findings
No findings of significance were identified.
1R13
Maintenance Risk Assessments and Emergent Work Control (71111.13)
a.
Inspection Scope
The inspectors reviewed four risk assessments for planned or emergent maintenance
activities to verify that AmerenUE met the requirements of 10 CFR 50.65(a)(4) for
assessing and managing increases in plant risk (four inspection samples). The
inspectors compared AmerenUEs risk assessment and risk management actions
against the requirements of 10 CFR 50.65(a)(4); the recommendations of Nuclear
Management and Resource Council 93-01, Industry Guidelines for Monitoring the
Effectiveness of Maintenance at Nuclear Power Plants, Revision 3; and Engineering
Department Procedure EDP-ZZ-01129, Callaway Plant Risk Assessment. The
inspectors reviewed the following risk assessments:
Preventative maintenance and testing of the turbine-driven auxiliary feedwater
pump (TDAFP) on August 1, 2005. The inspectors observed AmerenUEs risk
contingency activities from the control room and auxiliary building.
Unplanned inoperability of the TDAFP due to a misalignment of the overspeed
trip mechanism coincident with essential 4 kV load shed sequence testing on
August 18, 2005. The inspectors observed AmerenUEs risk management
activities from the control room and auxiliary building.
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Enclosure
Planned RHR Train A outage on August 23 and 24, 2005. The inspectors
observed AmerenUEs risk management activities from the control room.
The reactor coolant system (RCS) initial drain down (during refueling outage) to
the reactor vessel flange on September 20, 2005. The inspectors observed
AmerenUEs risk management activities from the control room and auxiliary and
reactor buildings.
b.
Findings
No findings of significance were identified.
1R14
Personnel Performance During Nonroutine Plant Evolutions (71111.14)
a.
Inspection Scope
The inspectors reviewed two non-routine plant events for personnel performance (two
inspection samples). The inspectors reviewed each event to verify proper operator
response. The inspectors used operator logs, plant computer data, charts, and
condition adverse to quality documents to determine what occurred, how the operators
responded, and whether the responses were in accordance with plant procedures. The
inspectors selected the following events:
Rapid power reduction due to the failure of a main feed pump turbine speed
control on June 28, 2005 (CAR 200504493)
Failure to maintain a centrifugal charging pump incapable of injection during cold
shutdown on September 20, 2005 (CAR 200507092)
b.
Findings
Failure to Maintain Cold Overpressure Mitigation Measures
Introduction. A self-revealing apparent violations of TS 5.4.1.a, Procedures, was
identified after an operator error resulted in the failure of AmerenUE to maintain the cold
overpressure mitigation system configuration while in Mode 5.
Description. An operator error resulted in both centrifugal charging pumps being
capable of injecting into the reactor while in Mode 5. Technical Specification 3.4.12,
Cold Overpressure Mitigation System, prohibited more than one centrifugal charging
pump from being capable of injecting into the reactor vessel while in Mode 5.
AmerenUE used Procedure OSP-BG-00002, COMS-Verify One Centrifugal Charging
Pump Incapable of Injecting into RCS, to isolate and administratively control the
centrifugal charging pump flow path while in Mode 5. Using Procedure OSP-BG-00002,
AmerenUE isolated and locked the Train A centrifugal charging pump discharge valve
and hung a placard describing the requirement to maintain the valve closed.
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Enclosure
AmerenUE performed Procedure ISP-SA-2413A, Diesel Generator and Sequencer
Testing (Train A), on September 20, 2005. After the completion of the test, the test
director gave directions to an auxiliary operator to restore the system lineup.
Procedure ISP-SA-2413A included a step to open the Train A charging pump discharge
valve. The test director recognized that the discharge valve needed to remain isolated
for compliance with TS 3.4.12. The test director crossed out the step to open the
discharge valve on the system restoration sheet. However, the auxiliary operator used a
different copy of the procedure which did not have the step crossed out. The operator
ignored the administrative controls and unlocked and opened the discharge valve. The
auxiliary operator returned the placard and lock to the test director. The test director
recognized the inappropriate configuration and immediately had the improper alignment
corrected. The inspectors determinated that a less than adequate test procedure, a
poor pre-job brief, and poor supervisory direction contributed to this event.
Analysis. AmerenUEs failure to establish and follow adequate procedures was a
performance deficiency. This finding affected the barrier integrity cornerstone and the
configuration control, procedure quality, and human performance attributes of
maintaining functionality of the RCS. This finding is greater than minor because, if left
uncorrected, it would have become a more significant safety concern involving the
integrity of the reactor coolant system. The inspectors used the shutdown operation
situation's, Manual Chapter 0609, Significance Determination Process, Appendix G,
Shutdown Operations Significance, Checklist 2. Although the condition only
represented a degradation of the licensee controls to the barrier function and did not
result in a noncompliance with low-temperature overpressure protection (LTOP) TS,
discussion with the Office of Nuclear Reactor Regulation (NRR) identified that a Phase 3
analysis should be performed. This finding is currently under review by NRR. This
finding, which involved the failure of an operator to follow procedure, inadequate briefing
and a less than adequate test procedure, was associated with the crosscutting area of
human performance.
Enforcement. Technical Specification 5.4.1.a, Procedures, required the written
procedures specified in Regulatory Guide 1.33, Appendix A, to be implemented.
Appendix A required procedures for shutdown, appropriate for the RCS, to be
implemented. Procedure OSP-BG-00002 required AmerenUE to ensure only one
centrifugal charging pump was capable of injecting to the RCS. Contrary to the above,
AmerenUE did not ensure only one centrifugal charging pump was capable of injecting
to the RCS. Pending determination of the final safety significance of this issue, this
issue is being treated as an AV consistent with Section VI.A of the NRC Enforcement
Policy. This issue was entered into AmerenUE's corrective action program
(CAR 200507092) (AV 05000483/2005004-01).
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Enclosure
1R15
Operability Evaluations (71111.15)
a.
Inspection Scope
The inspectors reviewed seven operability determinations involving risk significant
equipment during the inspection (seven inspection samples). The inspectors reviewed
the technical adequacy of the operability determinations to verify that operability was
justified and compensatory measures were appropriate and controlled. The inspectors
reviewed plant status documents such as operator shift logs, emergent work
documentation, deferred modifications, and standing orders to determine if an
operability determination was warranted for degraded components. The inspectors
used the FSAR, TSs, and design basis documents as the bases to determine the
technical adequacy of licensee prepared operability determinations. The inspectors
reviewed the following equipment conditions and associated operability determinations:
Operability
200505194, Unanalyzed differential pressure across
RHR sump Valves EJHV8811A and EJHV8811B, on July 28, 2005
Operability
200505224, Degraded control room gas treatment
boundary, on July 27, 2005
Operability
200505701, Degraded diaphragm and regulator on
ESW Valve EFHV0044, on August 15, 2005
Operability
200505801, Degraded TDAFP trip throttle
Valve FCHV0312 discovered out of alignment, on August 18, 2005
Operability
200505994, Degraded ESW flow across the
component cooling water heat exchanger, on August 25, 2005
Operability Determination 200206245, Containment cooler standpipe level
control valve not closing properly, on August 26, 2005
Operability
Failure of main steam isolation Valve C
four-way valve, on December 29, 2004 (Unresolved
Item 05000483/2005002-05). This unresolved item is closed in Section 40A5 of
this report.
b.
Findings
.1
Misalignment of the TDAFP due to Personnel Error
Introduction. A self-revealing Green NCV of TS 5.4.1.a, Procedures, was identified
after AmerenUE failed to properly align the TDAFP mechanical overspeed trip
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Enclosure
mechanism after surveillance testing. The misaligned trip mechanism increased the
probability that the turbine would trip if the pump would have been required to respond
to an event.
Description. AmerenUE failed to properly reset the TDAFP mechanical overspeed trip
mechanism after surveillance testing on August 1, 2005. A plant engineer identified the
misaligned trip linkage on August 18, 2005. The mechanical trip system used a weight
attached to the outboard end of the turbine shaft to push a tappet if an overspeed
condition was reached. The tappet nut held the trip and throttle valve trip rod in position.
When the weight extended and hit the tappet, the tappet nut and head lever
disengaged, allowing spring tension to move a rod, which in turn closed the trip and
throttle valve. The tappet nut is required to be reset locally.
Procedure OOA-ZZ-SEC06, Turbine Building Area 5 Operator Aid, Revision 6,
specified that the trip linkage head lever must rest against the flat side of the tappet nut
when the overspeed trip was reset. The beveled portion of the tappet nut was
misaligned when locally reset on August 1, 2005. The misalignment resulted in less
than the necessary contact surface between the tappet nut and head lever to hold the
trip mechanism open. However, a small burr on one of the contact surfaces created
enough friction to prevent the trip mechanism from closing the trip and throttle valve.
The trip linkage misalignment resulted in increased probability of an inadvertent TDAFP
trip during accident conditions.
Analysis. AmerenUE's failure to properly reset the turbine mechanical overspeed trip
mechanism after surveillance testing was a performance deficiency. The inspectors
used the at-power significance determination process to analyze this finding. This
finding is greater than minor because the degraded trip mechanism affected the reactor
mitigating systems cornerstone and the equipment performance attribute to ensure
availability of systems that respond to prevent core damage. This finding is only of very
low safety significance because the condition was not a design or qualification
deficiency confirmed to result in loss of function per Generic Letter 91-18; did not result
in an actual loss of safety function of a system; did not increase the likelihood of a fire;
and did not screen as potentially risk significant due to a seismic, flooding, or severe
weather initiating event. This finding, which involved the failure of an operator to follow
procedure, was associated with the crosscutting area of human performance.
Enforcement. Technical Specification 5.4.1.a required written procedures specified in
Regulatory Guide 1.33, Appendix A, to be implemented. Appendix A, required
procedures for operation of auxiliary feedwater systems to be implemented.
Procedure OOA-ZZ-SEC06 provided for operation of the TDAFP and required that the
operator verify the tappet nut was completely seated when resetting the turbine
mechanical overspeed trip mechanism. Contrary to the above, on August 1, 2005, the
operator did not verify the tappet nut was completely seated when resetting the turbine
mechanical overspeed trip mechanism. Because this finding is of very low safety
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Enclosure
significance and was entered into AmerenUE's corrective action program
(CAR 200505801), this violation is being treated as an NCV, consistent with
Section VI.A of the NRC Enforcement Policy (NCV 05000483/2005004-02).
.2
Potential Failure of the RHR Containment Suction Valves During Certain Design Bases
Events
Introduction. The inspectors are reviewing AmerenUE's actions associated with the
RHR containment suction valves as an unresolved item. This issue will remain
unresolved pending additional review by the inspectors. No analysis or enforcement
reviews were performed for this unresolved item.
Description. AmerenUEs analyzed maximum differential pressure between the
containment recirculation sump and RHR systems was not the limiting case. AmerenUE
used this maximum differential pressure value to ensure that the containment
recirculation sump RHR suction valve actuators were capable of opening during all
design bases accidents. The RHR system suction path was designed to automatically
transfer to the containment recirculation sumps after the refueling water storage tank
level deceases to thirty-six percent. The RHR containment sump valves must open for
the transfer to occur. The reactor operator subsequently aligns the RHR pump
discharge to the safety injection and high head emergency core cooling pump suctions
to ensure adequate net positive suction head for the cold leg recirculation phase of
accident mitigation. AmerenUE concluded a 53 psid maximum differential pressure for
the RHR sump valves in Calculation RFR 05353, Revision F. However, during some
small and medium size loss of coolant accidents, the RHR suction pressure may
exceeded the 53 psid maximum differential pressure analyzed by AmerenUE. During
these accident conditions, the reactor pressure may stay above the 195 psig discharge
pressure of the RHR pumps. Flow from the RHR heat exchanger outlet would be
diverted back to the pump suction. A minimum flow valve automatically opens to ensure
RHR pump discharge flow is maintained at least 816 gpm. Based on the pump curve,
the RHR pump will develop a 215 psid head at 816 gpm. The RHR suction
configuration also included a relief valve capable of passing 700 gpm at 450 psig.
AmerenUE evaluated OE from the Catawba and McGuire (CAR 200504370) plants
during June 2005. This OE alerted the industry to the potential of higher than previously
considered differential pressure across the RHR sump valves. In response to the OE,
AmerenUE operated the RHR pumps for about 30 minutes in the minium flow
configuration and observed 189 psid across the sump valve. AmerenUE concluded that
the valve actuator would still open based on a linear extrapolation of the actuator torque
at the higher differential pressure. The inspectors determined that, based on the
215 psid developed RHR pump head with no condensable gases present in the system,
differential pressures greater than 189 psid are possible in the RHR suction line.
Additionally, Westinghouse Calculation WCAP 13097, Revision 0, Section 1.2, Design
Basis Review, recommended that 464 psid differential pressure generically be used for
sizing the containment sump valve actuator torque. This issue is considered unresolved
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Enclosure
pending additional inspector review of the supporting RHR differential pressure
calculations and of current and past RHR sump valve operability (Unresolved
Item 05000483/2005004-03) .
1R19
Post Maintenance Testing (71111.19)
a.
Inspection Scope
The inspectors reviewed two risk significant post maintenance tests to verify that
AmerenUE adequately demonstrated the safety function of components affected by
maintenance activities (two inspection samples). The inspectors verified that testing
procedures were properly reviewed and approved and incorporated appropriate
acceptance criteria. The inspectors used information in the TSs, the FSAR, and
Section XI of the American Society of Mechanical Engineers Code, as the bases for
acceptability of sampled postmaintenance tests. The inspectors completed an in-office
review of the completed work packages. The sample included the following post
maintenance tests:
PMT 05101938/910, PMT 718923/904, and PMT P973710/910, after emergency
core cooling room cooler and Valve HBHV7150 maintenance on August 23,
2005
PMT P721525/910 and PMT 721525/910, after RHR pump motor maintenance
on August 23, 2005 (pump, room cooler, and Valve EJHV0610)
b.
Findings
No findings of significance were identified.
1R20
Refueling and Outage Activities (71111.20)
a.
Inspection Scope
The inspectors evaluated and observed selected refueling outage activities to ensure
that AmerenUE appropriately considered plant risk when developing outage schedules
and adequately controlled plant configuration. The inspectors also reviewed refueling
activities to verify that AmerenUE developed appropriate mitigation strategies for losses
of key safety functions and complied with the operating license and TS requirements.
Outage Plan Review
Prior to the outage, the inspectors performed an in-office review of the refueling risk
analysis and schedule to verify that AmerenUE appropriately considered risk, industry
experience, and previous site-specific problems. The inspectors compared AmerenUEs
outage plan with Administrative Procedure APA-ZZ-00150, Outage Preparation and
Execution, Revision 15, and Nuclear Utility Management and Resource Council 91-06,
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Enclosure
Guidelines for Industry Actions to Assess Shutdown Management, December 1991, as
a basis for acceptability. The inspectors also reviewed refueling activities against the
requirements of Procedure APA-ZZ-00322, Integrated Work Management Process
Description, Revision 1.
Monitoring of Shutdown Activities
The inspectors reviewed the RCS cooldown on September 17, 2005, to verify that
AmerenUE did not exceed TS cooldown limits. The inspectors compared the plant
cooldown data against Procedure OTG-ZZ-0006, Plant Cooldown Hot Standby to Cold
Shutdown, Revision 6 and the Curve Book, Figure 8-6, RCS Pressure-Temperature
Limitations.
Licensee Control of Outage Activities
The inspectors verified that AmerenUE maintained key safety functions and applicable
TSs when taking equipment out of service and transitioning the plant. The inspectors
attended daily outage status meetings and observed AmerenUEs control of outage
activities to verify that defense-in-depth risk was commensurate with the outage risk
control plan. The inspectors compared AmerenUEs evaluation of emergent work risk
with Engineering Department Procedure EDP-ZZ-1129, Callaway Plant Risk
Assessment, Revision 8; and Nuclear Utility Management and Resource Council 91-06,
Guidelines for Industry Actions to Assess Shutdown Management.
Inventory Control
The inspectors observed the RCS drain down from the control room on September 20,
2005, to verify that the flow paths, configurations, and alternative means for inventory
addition were consistent with the outage risk plan. The inspectors verified AmerenUEs
Limited Inventory Controls contingencies were implemented prior to the RCS drain
down.
Reactivity Control
The inspectors reviewed AmerenUEs outage reactivity controls to verify that the TS
reactivity control requirements were met. The inspectors performed auxiliary building
and control room walkdowns on September 17 and 19 to verify that AmerenUE
maintained the required boron injection flow paths. The inspectors also reviewed
outage activities that could cause unexpected reactivity changes.
Refueling Activities
The inspectors observed fuel handing activities from the reactor building and control
room on September 24 and 26, 2005, to verify that operations were performed in
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Enclosure
accordance with the FSAR, plant TSs, and fuel handling procedures. The inspectors
observed fuel handing coordination from the control room to verify that AmerenUE
tracked the location of fuel assemblies during core offload.
b.
Findings
No findings of significance were identified.
1R22
Surveillance Testing (71111.22)
a.
Inspection Scope
The inspectors observed and/or reviewed seven risk significant surveillance tests to
verify that AmerenUE adequately demonstrated component safety functions and to
assess operational readiness (seven inspection samples). The inspectors verified that
testing procedures were properly reviewed and approved with appropriately incorporated
acceptance criteria. The inspectors used information in the TSs, the FSAR,Section XI
of the American Society of Mechanical Engineers Code, and licensee procedural
requirements as the bases for acceptability of sampled surveillance tests. The samples
included the following surveillance tests:
Surveillance 05510445, Containment cooler Train B flow test performed on
July 13, 2005. The inspectors completed an in-office review of the completed
surveillance test package.
Surveillance 05508639, Emergency exhaust system test on July 20, 2005. The
inspectors completed an in-office review of the completed surveillance test
package.
Surveillance 05510817, Containment isolation verification performed on July 21,
2005. The inspectors observed a portion of the test from the control room and
completed an in-office review of the completed surveillance test package.
Surveillance 05508797/500, RHR Train A inservice test performed on July 26,
2005. The inspectors observed AmerenUEs brief, equipment line-up, and field
performance of the test and completed an in-office review of the completed
surveillance test package.
Surveillance 05509057, Component cooling Train B inservice test performed on
August 3, 2005. The inspectors observed a portion of the test from the control
room and completed an in-office review of the completed surveillance test
package.
Surveillance 05512469, TDAFP inservice test performed on August 29, 2005.
The inspectors completed an in-office review and observed a portion of the test
from the auxiliary building and control room.
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Enclosure
Surveillance 05502140, Train A RHR valve inservice test performed on
September 15, 2005. The inspectors observed the test from the control room
and performed and in-office review of the completed surveillance test package.
b.
Findings
No findings of significance were identified.
1EP6 Drill Evaluation (71114.06)
a.
Inspection Scope
The inspectors observed one emergency drill during the inspection period (one
inspection sample). The inspectors observed AmerenUEs response to a simulated
steam generator tube rupture accident without onsite power on August 10. The
inspectors observed the drill to evaluate the adequacy of AmerenUEs emergency
response and to verify that AmerenUE implemented proper emergency action level
classification and protective action recommendations. The inspectors observed the
exercise from the Technical Support Center. The inspectors compared drill
observations against Emergency Plan Implementing Procedure EIP-ZZ-00101,
Classification of Events, and Emergency Plan Implementing Procedure EIP-ZZ-00201,
Notifications, to evaluate licensee performance.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1
Daily Reviews
a.
Inspection Scope
The inspectors performed a daily review of items entered into AmerenUEs corrective
action program. The inspectors performed this screening to identify any repetitive
equipment failures or adverse human performance trends for followup. The inspectors
also attended selected conditions adverse to quality report screenings and daily plant
status meetings.
b.
Findings
No findings of significance were identified
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Enclosure
.2
Annual Sample Review
Routine Review of Identification and Resolution of Problems
a.
Inspection Scope
The inspectors performed detailed in-office reviews and walkdowns of plant equipment
related to three significant conditions adverse to quality (three inspection samples). The
inspectors reviewed AmerenUE's CAR reports to verify that the full extent of the issues
was identified, that AmerenUE performed appropriate evaluations, and that adequate
corrective actions were specified and prioritized. The inspectors evaluated the reports
against the requirements of Administrative Procedure APA-ZZ-00500, Corrective Action
Program, and 10 CFR Part 50, Appendix B. The inspectors reviewed the following two
samples:
CAR 200505224, Control building ventilation boundary degraded due to failed
door latch
CAR 200504163, Unexpected main steam feedwater isolation signal logic
Cabinet SA075B channel failure
CAR 200505194, Containment recirculation sump suction valves
b.
Findings
Ineffective Corrective Actions for Control Building Habitability Boundary Maintenance
Introduction. The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B,
Corrective Action, after ineffective corrective actions resulted in a repeat degradation
of the control building emergency ventilation habitability boundary.
Description. AmerenUE's corrective actions to prevent a repeat failure of the control
building emergency ventilation habitability boundary were ineffective. Technical Specification 3.7.10, Control Room Emergency Ventilation System, established the
requirement for maintaining the habitability boundary. Door DSK33021, located
between the essential switchgear and emergency diesel generator rooms, functions as
a control building pressure and switchgear room Halon boundary. On July 25, 2005,
AmerenUE identified that the door latching mechanism was bent and a corrective
maintenance request was initiated. A plant operator identified that the door latch had
failed on July 27, 2005, resulting in the inoperability of a control building ventilation
habitability boundary. AmerenUE repaired the door. AmerenUE identified that the work
control organization had authorized work on the door without approval of the operations
department. As a result, operations did not understand that the habitability boundary
had been compromised by the maintenance and that TS Limiting Condition for
Operations 3.7.10 should have been applied. A plant operator identified that the door
latch had been removed for maintenance on July 29, 2005. The work control
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Enclosure
organization had again authorized work on the door without approval of the operations
department. As a result, operations was not informed that the habitability boundary had
been compromised by the maintenance. AmerenUEs corrective actions following the
failure of maintenance to gain approval by the operations department, following the
July 27, 2005, event, failed to prevent reoccurrence of a similar event on July 29, 2005.
This finding is related to less than adequate corrective action and is associated with the
crosscutting area of problem identification and resolution.
AmerenUEs operability evaluation of the degraded door was less than adequate.
AmerenUE concluded the control building emergency ventilation habitability boundary
was operable because the unlatched door was kept in the closed position by the higher
pressure in the diesel generator building. However, the operability evaluation did not
consider that the control room emergency pressurization fan would pressurize the other
side of the door at least 1/8 inch of water during accident conditions. AmerenUE did not
consider that the diesel generator air inlet was located next to the door, which would
cause a lower pressure on the other side of the door when the diesel generator was
operating. Additionally, the diesel room supply fan is designed to turn off when the room
temperature decreases to less than 85 degrees.
Analysis. AmerenUEs failure to implement effective corrective actions to ensure plant
configuration was consistent with accident analysis assumptions was a performance
deficiency. This finding was greater than minor because it was associated with the
integrity of the control building pressure envelope. Because this finding involved the
degradation of barrier integrity, the finding was evaluated using the significance
determination process for at-power situations. The inspectors concluded that the finding
was only of very low safety significance because the finding only represented a
degradation of the radiological barrier function provided for the control room. A
crosscutting aspect associated with problem identification and resolution was identified
for the failure to prevent a repeat degradation of the ventilation system boundary.
Enforcement. Title 10 of the Code of Federal Regulations, Part 50, Appendix B,
Criterion XVI, Corrective Action, required that measures be established to assure that
conditions adverse to quality are promptly identified and corrected. For significant
conditions adverse to quality, measures shall assure that the cause of the condition is
determined and corrective action is taken to preclude repetition. Contrary to the above,
AmerenUEs corrective actions failed to preclude recurrence of degraded control
building habitability barriers. Because of the very low safety significance and
AmerenUEs action to place this issue in their corrective action program
(CARs 200505224 and 200505279), this violation is being treated as an NCV in
accordance with Section VI.A.1 of the Enforcement Policy (0500483/2005-05).
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Enclosure
4OA4 Crosscutting Aspects of Findings (71152)
Section 1R14 documents one finding with human performance crosscutting aspects
associated with failure of an operator to follow a procedure which resulted in the loss of
configuration control of the cold overpressure mitigation system configuration while the
reactor was in Mode 5 (AV 05000483/2005004-01).
Section 1R15 documents one finding with human performance crosscutting aspects
associated with failure of an operator to follow a procedure which resulted in the
misalignment of the TDAFP mechanical overspeed trip mechanism
4OA6 Management Meetings
Exit Meeting Summary
On August 15, 2005, the inspectors presented the results of the follow up to Temporary
Inspection 2515/163, Operational Readiness of Offsite Power, to Mr. T. Herrman,
Manager, Engineering, and other members of his staff who acknowledged the findings.
The inspectors presented the preliminary inspection results to Mr. A. Heflin, Site Vice
President, and other members of licensee management at the conclusion of onsite
portion the Heat Sink Performance biennial inspection on July 1, 2005. The inspectors
presented the final inspection results to Mr. M. Reidmeyer, Supervisor, Regulatory
Affairs, on August 5, 2005. No proprietary information was reviewed.
On September 26, 2005, the resident inspectors presented their inspection results to
Mr. C. Naslund, Senior Vice President and Chief Nuclear Officer, and other members of
his staff who acknowledged the findings.
The inspectors verified that no proprietary information was reviewed during the
inspection.
ATTACHMENT: SUPPLEMENTAL INFORMATION
A-1
ATTACHMENT
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
A. Heflin, Site Vice President
T. Herman, Supervisor, Design Engineering
M. Reidmeyer, Supervisor, Regional Regulatory Affairs
T. Sharkey, Superintendent, Engineering Technical Support
L. Thibault, General Plant Manager
LIST OF ITEMS OPENED AND CLOSED
Opened
Potential Failure of the RHR Containment Suction Valves
During Certain Design Bases Events (Section 1R15)
Opened and Closed
Failure to Maintain Cold Overpressure Mitigation Measures
as Required by TSs (Section 1R14)05000483/2005004-02
Misalignment of the TDAFP due to Personnel Error
(Section 1R15)05000483/2005004-04
Ineffective Corrective Actions for Door
Failures
(Section 4OA2)
DOCUMENTS REVIEWED
Section 1R04: Equipment Alignment
Procedures
OSP-BB-00003, PORV/RHR COMS Alignment Verification, Revision 10
OSP-BG-00002, Verify One Centrifugal Charging Pump Incapable of Injection into RCS,
Revision 13
OTN-EJ-00001, Residual Heat Removal System, Revision 20
OTS-EJ-0004A, RHR Pump A Non-Surveillance Run, Revision 2
A-2
ATTACHMENT
Miscellaneous
RHR Train A Inservice Test, Revision 34
UFSAR Section 6.3, Emergency Core Cooling System
Drawing M-22EJ01, Residual Heat Removal System, Revision 54
Section 1R05: Fire Protection
Procedures
EIP-ZZ-00226, Fire Response Procedure for Callaway Plant, Revision 11
APA-ZZ-00743, Fire Team Organization and Duties, Revision 18
Section 1R07: Heat Sink Performance
Procedures
CDP-ZZ-00200, Appendix D, Closed Cooling Systems Tables, Revsion 1
CDP-ZZ-00940, Auxiliary Water Systems Chemistry Optimization Plan, Revision 4
CDP-ZZ-00950, Raw Water Systems Control Program, Revision 4
CTP-ZZ-06000, Circ and Service Water Chemical Additions, Revision 47
EDP-ZZ-01112, Heat Exchanger Predictive Performance Manual, Revision 8
EDP-ZZ-01121, Raw Water Systems Predictive Performance Manual,Revision 7
ETP-GL-00001, Area Room Cooler Test, Revision 4
ETP-EF-0002B, Essential Service Water Train B Flow Verification, Revision 9
ETP-EF-0002A, Essential Service Water Train A Flow Verification, Revision 8
ETP-EG-00001, Component Cooling Water Heat Exchanger Test, Revision 5
ETP-ZZ-03001, GL 89-13 Heat Exchanger Inspection, Revision 5
MPM-ZZ-QQ001, Room Cooler Inspection, Revision 12
OSP-EF-P001B, ESW Train B Inservice Test, Revision 41
OSP-EF-V001B, ESW Train B Valve Operability, Revision 29
OSP-ZZ-00001, Control Room Shift and Daily Log Readings/Channel Checks, Revision 45
OTS-EF-P001B, Performance Testing of Essential Service Water Pump B, Revision 0
Specifications
10466-M-072(Q), Design Specification for Component Cooling Water Heat Exchangers for the
Standardized Nuclear Unit Power Plant System (SNUPPS), Revision 11
072-00024, Instruction Manual for Component Cooling Water Heat Exchangers for the
SNUPPS Project, Revision 6
Callaway Action Requests
200304241
200304595
200308377
200308548
200401550
200403567
200403590
200403664
200404441
200407327
A-3
ATTACHMENT
Requests for Resolution
07809C
19513
22364A
Calculations
EF-45, Four Containment Coolers with New Coils, Revision 5
EF-49, Ultimate Heat Sink Thermal Transient Analysis, Revision 0, Addendum 1
EG-20, Max Component Cooling Water (CCW) Temperature During Post - LOCA, Revision 0,
Addendum 1
EG-42, Calculate the Number of CCW Tubes That Can Be Plugged, Revision 0
GN-03, Determine the Minimum ESW Flow Rate to GN Coolers with New Coils, Revision 5,
Addendum 3
KJ-10, Determine Tube Plugging Limits For DG Intercooler Heat Exchangers, EKJ03A/B, DG
Jacket Water Heat Exchangers, EKJ06A/B, and the Lube Oil Coolers, EKJ04A/B, Revision 0
EF-049, UHS Thermal Transient Analysis, Revision 0
EF-049, UHS Thermal Transient Analysis, Revision 0, Addendum 1
EF-52, Heat Exchanger Performance Based on Reduced ESW Temperature and Flow,
Revision 1
NESE-1081, Aeorfin Containment Cooler Performance Data Assuming 33EF or 95EF ESW
Water Temperature, Addendum 1
Maintenance Orders
P676141
P676150
P701990
P701992
P718659
W236012
Miscellaneous
Inservice Test Data for ESW Train B Pump and Valves from June 2002 through June 2005
Report PD04594.03, Record of Eddy Current Inspection of CCW Heat Exchanger A, April 2004
Report PR 17-17, Emergency Diesel Generator B - Heat Exchangers, dated September 2002
Final Safety Analysis Report, Section 9.2.1.2, Essential Service Water System
Report UOTCR 03-018, Cycle 12 Raw Water Report, dated March 27, 2003
A-4
ATTACHMENT
Report UOTCR 05-006, Cycle 13 Raw Water Report, dated February 9, 2005
Generic Letter 89-13, Service Water System Problems Affecting Safety-Related Equipment,
dated July 18, 1989
Generic Letter 89-13, Supplement 1, Service Water System Problems Affecting Safety-Related
Equipment, dated April 4, 1990
Letter ULNRC-2146, Response to Generic Letter 89-13 - Service Water System Problems
Affecting Safety-Related Equipment, dated January 29, 1990
PROTO-HX, Version 4.10, Shell and Tube Heat Exchangers User Documentation
NUPIC Audit 19290, Audit of Proto Power Corporation, dated June 22, 2005
Essential Service Water - EF - Lesson Plan
Figure B.3.6.6-1, Containment Cooler Heat Removal Minimum Cooling Flow Rate
Heat Exchanger Specification Sheets for EKJ03A/B, EKJ04A/B, EKJ06A/B, SGN01A/B/C/D,
SGL15A/B, and EEG01A/B
Section 1R13: Maintenance Risk Assessments and Emergent Work Evaluation
Procedures
EDP-ZZ-01128, Maintenance Rule Program, Revision 6
EDP-ZZ-01129, Callaway Plant Risk Assessment, Revision 8
ODP-ZZ-00001, Operations Department - Code of Conduct, Revision 23
Other
Nuclear Management and Resource Council 93-01, Industry Guidelines for Monitoring the
Effectiveness of Maintenance at Nuclear Power Plants, Revision 3
Section 1R20: Refueling and Outage Activities
Procedures
APA-ZZ-00150, Outage Preparation and Execution, Revisions 16 and 17
APA-ZZ-00500, Corrective Action Program, Revision 38
EDP-ZZ-01129, Callaway Plant Risk Assessment, Revision 8
A-5
ATTACHMENT
MOA-SM-00001, Containment Equipment Hatch Operation for Temporary Opening and
Closing, Revision 0
ODP-ZZ-00002, Equipment Status Control, Revision 28
OTO-ZZ-00012, Severe Weather, Revision 3
Miscellaneous
NUMARC 93-01, Revision 3, Industry Guidelines for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants
FSAR Section 3.5, Missile Protection
FSAR Section 9.5, Other Auxiliary System
FSAR Section 16.7.11, Area 5 Missile Shields
Section 1R22: Surveillance Testing
Procedures
OTN-EJ-00001, Residual Heat Removal System, Revision 20
OSP-EG-P01BD, CCW Train B Pump and Valve Inservice Test, Revision 19
OSP-EJ-P001A, RHR Train A Inservice Test, Revision 39
OSP-GP-00001, Containment Isolation Verification, Revision 14
OTS-EJ-0004A, RHR Pump A Non-Surveillance Run, Revision 2
Miscellaneous
RHR Train A Inservice Test, Revision 34
UFSAR Section 6.3, Emergency Core Cooling System
Drawing M-22EJ01, Residual Heat Removal System, Revision 54
TSs, the FSAR,Section XI of the American Society of Mechanical Engineers Code, and
licensee procedural requirements as the bases for acceptability
Section 4OA2: Identification and Resolution of Problems
Quality Assurance Audits and Surveillance Reports
AP05-004, Audit of testing, May 6, 2005
AP05-005, Chemistry program audit, June 27, 2005
AP05-006, Personnel qualifications and training audit, July 8, 2005
A-6
ATTACHMENT
AP05-008, Quality Assurance Independent Audit of the Operating Quality Assurance Program
AP05-009, Quality Assurance audit of emergency preparedness, September 6, 2005
AUCA 05-017, Underground electrical line hit during concrete demolition, June 1, 2005
SEGR 05-06-009, Independent technical review report, June 29, 2005
SP05-006, Worker practices, June 8, 2005
SP05-025, Readiness assessment for Refuel 14 for supplemental personnel, September 13,
2005
SP05-033, Surveillance of the method of identifying procedures affected by Refuel 14
modifications, September 15, 2005
CAR 200504855, Audit of Internal Audit Program Findings
CAR 200504856, QME Audit of ITR-ORC Programs - Recommendations
APA-ZZ-00500, Corrective Action Program, Revision 38
LIST OF ACRONYMS
Callaway Action Request
component cooling water
Essential Service Water
Final Safety Analysis Report
operating experience
turbine-driven auxiliary feedwater pump
TS
Technical Specification