ML060450731
| ML060450731 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 02/14/2006 |
| From: | William Jones NRC/RGN-IV/DRP/RPB-B |
| To: | Naslund C Union Electric Co |
| References | |
| IR-05-005 | |
| Download: ML060450731 (57) | |
See also: IR 05000483/2005005
Text
February 14, 2006
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 05000483/2005005
Dear Mr. Naslund:
On December 31, 2005, the NRC completed an inspection at your Callaway Plant. The
enclosed report documents the inspection findings which were discussed on January 6, 2006,
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.
Based on the results of this inspection, the NRC has determined that one Severity Level IV
violation of NRC requirements occurred. The NRC has also identified six additional issues that
were evaluated under the risk significance determination process as having very low safety
significance (Green). The NRC has determined that there are four violations associated with
the significance determination process issues. In addition, licensee-identified violations which
were determined to be of very low safety significance are listed in the report. All of the
violations are being treated as noncited violations (NCVs), consistent with Section VI.A of the
Enforcement Policy. The NCVs 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 05000483/2005005
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
-3-
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
Chief, Radiological Emergency
Preparedness Section
Kansas City Field Office
Chemical and Nuclear Preparedness
and Protection Division
Dept. of Homeland Security
9221 Ward Parkway
Suite 300
Kansas City, MO 64114-3372
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)
Regional State Liaison Officer (WAM)
J. Dixon-Herrity, OEDO RIV Coordinator (JLD)
ROPreports
CWY Site Secretary (DVY)
SUNSI Review Completed: __wbj_ ADAMS: : Yes
G No Initials: ___wbj__
- Publicly Available G Non-Publicly Available G Sensitive : Non-Sensitive
R:\\_REACTORS\\_CW\\2005\\CW2005-05RP-MSP.wpd
RI:DRP/B
SRI:DRP/B
C:DRS/EB2
C:DRS/EB1
DEDumbacher
MSPeck
LJSmith
JAClark
E - WBJones
E - WBJones
GDReplogle for
/RA/
2/9/06
2/9/05
2/13/05
2/13/05
C:DRS/PSB
C:DRS/OB
C:DRP/B
MPShannon
ATGody
WBJones
/RA/
/RA/
/RA/
2/13/05
2/13/05
2/14/05
OFFICIAL RECORD COPY
T=Telephone E=E-mail F=Fax
Enclosure
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ENCLOSURE
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket:
50-483
License:
Report No.:
Licensee:
Union Electric Company
Facility:
Callaway Plant
Location:
Junction Highway CC and Highway O
Fulton, Missouri
Dates:
September 24 through December 31, 2005
Inspectors:
M. S. Peck, Senior Resident Inspector
D. E. Dumbacher, Resident Inspector
R. W. Deese, Senior Resident Inspector
B. D. Baca, Health Physicist
P. J. Elkmann, Emergency Preparedness Inspector
T. F. Stetka, Senior Operations Engineer
M. E. Murphy, Senior Operations Engineer
J. F. Drake, Operations Engineer
Approved By:
W. B. Jones, Chief, Project Branch B
Enclosure
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SUMMARY OF FINDINGS
IR 05000483/2005005; 09/24 - 12/31/2005; Callaway Plant: Equipment Alignment, Fire
Protection, Personnel Performance During Nonroutine Plant Evolutions, Permanent Plant Mods,
Refueling & Outage Activities, Licensed Operator Requal Program, and Emergency Plan &
Emergency Action Level Change.
This report covered a 3-month inspection by region based reactor inspectors and resident
inspectors. One Severity Level IV noncited violation, four Green noncited violations, and two
Green findings were identified. The significance of most findings is indicated by their color
(Green, White, Yellow, 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.
A.
Inspector-Identified and Self-Revealing Findings
Cornerstone: Initiating Events
Green. The inspectors determined that the failure to adhere to ANSI/ANS 3.5-1998, as
endorsed by Regulatory Guide 1.149, "Nuclear Power Plant Simulation Facilities for Use
in Operator Training and License Examinations," Revision 3, October 2001, as
committed to in the Callaway Plant Simulation certification dated March 13, 2000, was a
finding. Specifically, the simulator performance testing did not meet the standards
specified in ANSI/ANS 3.5-1998 in that: (1) all required parameters during the simulator
test were not recorded; and (2) simulator to baseline data comparisons were
unavailable.
The failure to evaluate and document simulator performance testing is more than minor
because it affected the Operator Requalification attribute of the Mitigating Systems and
Initiating Events cornerstone of reactor safety and is inconsistent with the requirements
of 10 CFR 55.46 in that simulator fidelity issues may not be identified which have the
potential of causing negative training. The finding was considered to be of very low
safety significance because the discrepancies have not yet impacted operator actions in
the plant such that safety-related equipment was made inoperable or that operators
failed to properly respond to plant transients. This issue is documented in the facility
licensees corrective action program as Callaway Action Request 200503956
(Section 1R11).
Cornerstone: Mitigating Systems
Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,
Criterion V, for the failure to adequately implement work order instructions and a
procedure for the inspection of the containment recirculation sump enclosure. The
licensees inspections failed to identify a 1.5-inch hole in the sump cover, which could
provide a path for foreign material to enter the containment sump. AmerenUE
completed a detailed inspection of the sump on April 27, 2004, in response to NRC
Enclosure
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Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump
Recirculation at Pressurized-Water Reactors, but failed to identify the 1.5-inch hole. A
subsequent inspection was performed on November 8, 2005, during Refueling
Outage RF 14 that also did not identify the hole in the containment sump enclosure.
This issue was entered into the corrective action program as Callaway Action
Request 200509189.
This finding is greater than minor because it is associated with the mitigating systems
cornerstone attribute of equipment performance and affects the associated cornerstone
objective to ensure availability and reliability of the containment recirculation sump
emergency core cooling system containment safety function. This finding is of very low
safety significance because the condition was a qualification deficiency confirmed not to
result in loss of function per Part 9900, Technical Assessment, Operability
Determination Process for Operability and Functional Assessment. The cause of this
finding is related to the crosscutting element of human performance in that personnel
failed to adequately implement a work instruction and procedure in inspecting the
containment sump configuration (Section 1R04).
Green. The inspectors identified a noncited violation of Technical Specification 5.4.1.d,
Fire Protection Program Implementation, associated with seven examples of
inadequately performed continuous fire watches. In September 2005, AmerenUE
provided verbal guidance to fire watch personnel that continuous fire watches may be
met by a 15-minute roving fire patrol. The roving patrol did not ensure adequate
compensatory action for fire areas with degraded detection or suppression capability.
As a result, fire watch personnel were not available to promptly detect, report, and
extinguish a fire while still in the incipient stage. AmerenUE did not evaluate this change
to ensure no adverse affect on the ability to achieve and maintain safe shutdown in the
event of a fire. This condition was entered into the corrective action program as
Callaway Action Request 200510325.
This finding is greater than minor because inadequate fire watches are associated with
the reactor safety mitigating systems cornerstone attribute to provide protection against
external factors and affect the associated cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. This finding is of very low safety significance
because the condition had an adverse affect on the Fixed Fire Protection Systems
element of fire watches posted as a compensatory measure for outages or
degradations. A low degradation rating was assigned to this finding as the provision
affected by this finding is expected to display nearly the same level of effectiveness and
reliability. The cause of this finding is related to the crosscutting element of human
performance in that the guidance provided was not adequate to ensure continuous fire
watches were appropriately conducted (Section 1R05).
Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,
Criteria V, Instructions, Procedures, and Drawings, associated with an inadequate
engineering procedure used for the verification of design calculations. The inadequate
procedure resulted in a nonqualified, nonsafety-related engineering calculation used to
demonstrate that the safety-related containment recirculation sump valves were capable
Enclosure
-4-
of performing the safety function described in the design bases. The performance
deficiency associated with this finding involved the failure of engineering personnel to
only use qualified calculations for safety-related applications. This finding was entered
into the Corrective Action Program as Callaway Action Request 200509849.
This finding is greater than minor because, if left uncorrected, this finding would become
a more significant safety concern. This finding is determined to have very low safety
significance because this finding involves a design deficiency confirmed not to result in
loss of operability per Part 9900, Technical Guidance, Operability Determination
Process for Operability and Functional Assessment. The cause of this finding is related
to the crosscutting element of human performance in that the procedure did not ensure
the calculations were qualified to support a design basis function of a safety-related
component (Section 1R17).
Cornerstone: Barrier Integrity
Green. The inspectors identified a noncited violation of Technical Specification 5.4.1.a,
Procedures, after AmerenUE Operations personnel failed to maintain the reactor
coolant system heatup and cooldown temperature limits on two occasions. On
November 7, 2005, plant operators decreased the reactor coolant system pressurizer
surge line temperature 260EF in a one-hour period. The operators conducted the rapid
cooldown after several containment lead shield blanket polyvinylchloride covers located
on the pressurizer surge line melted. On November 8, 2005, plant operators increased
the surge line temperature about 175EF in a one-hour period. Plant Technical Specification 3.4.3, RCS [reactor coolant system] Pressure and Temperature (P/T)
Limits, and plant procedures required reactor coolant system component temperature
changes (except the pressurizer) be limited to 100EF in one hour. This finding was
placed in the Corrective Action Program as Callaway Action Requests 200509487
and 200509143.
This finding was greater than minor because it is associated with the reactor safety
barrier integrity cornerstone attribute of equipment performance and affects the
associated cornerstone objective to ensure reasonable assurance that the reactor
coolant system piping barrier will protect the public from radionuclide releases caused
by accidents or events. This finding is determined to have very low safety significance
because an engineering evaluation concluded that the temperature transient did not
significantly increase the likelihood of a loss of reactor coolant system inventory or
degrade the ability to terminate a leak path. The cause of this finding is related to the
crosscutting element of human performance in that the reactor coolant system
pressurizer surge line heatup and cooldown limits were exceeded (Section 1R14).
Cornerstone: Emergency Preparedness
Severity Level IV. The inspectors identified a violation of 10 CFR 50.54(q) for
implementing a change to emergency action levels which decreased the effectiveness
of the emergency plan. Emergency Implementing Plan Procedure EIP-ZZ-00101,
Classifying the Emergency, Revision 33, limited application of emergency action
Enclosure
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Level 3E, Fire within Protected Area Boundary NOT Extinguished with 15 minutes of
Verification, so that fires in some plant areas which would be classified under the
previous revision may no longer be classifiable.
Implementation of changes to emergency action levels which decreased the
effectiveness of the emergency plan was a performance deficiency. The finding is more
than minor because removal of a classifiable condition from licensee emergency action
levels has the potential to impact safety, and licensee implementation of a change to
their emergency plan, which decreases the effectiveness of the plan without prior NRC
approval, impacts the regulatory process. This finding is a violation of 10 CFR 50.54(q).
The licensee has entered this issue into their corrective action system as Corrective
Action Report 200510162 (Section 1EP4).
Cornerstone: Miscellaneous
Green. The inspectors identified a finding after AmerenUE implemented less than
adequate risk management controls of the spent fuel pool water inventory. On
September 29, 2005, the core had been off-loaded to the spent fuel pool and the
transfer canal weir wall removed. The inspectors identified that the shutdown safety
plan did not establish specific controls for reactor refueling canal transfer tube
Valve ECV-995, which isolated the fuel transfer canal from the containment cavity or
provided for installation of the associated fuel transfer canal flange. Valve ECV-995 was
closed but was not identified in the shutdown risk management system and did not have
administrative controls established through the shutdown risk plan. NRC Information Notice 2005-16, Outage Planning and Scheduling - Impacts on Risk, emphasized that
most spent fuel pool events had a common thread of human error and involved
equipment misalignment. This finding was entered into the Corrective Action Program
as Callaway Action Requests 200507593 and 200507693.
This finding is greater than minor because, if left uncorrected, it would have become a
more significant safety concern. Because Manual Chapter 0609, Significance
Determination Process, does not specifically address findings related to the spent fuel
pool inventory, this finding is determined to have very low safety significance based on
NRC management review with input from a senior reactor analyst. The review
considered that the procedure used to manipulate the valve was not in use during this
period and that borated water makeup capabilities were available to the spent fuel pool.
No violation of regulatory requirements occurred (Section 1R20).
B.
Licensee-Identified Violations
Violations of very low 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 licensee's corrective action program. These violations and
corrective action tracking numbers are listed in Section 4OA7 of this report.
Enclosure
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REPORT DETAILS
Summary of Plant Status
The Callaway Plant was shut down for Refueling Outage 14 at the beginning of the inspection
period. Outage work included steam generator replacement and a major turbine overhaul.
AmerenUE completed the refueling outage and synchronized the generator to the grid on
November 19, 2005. The licensee returned to full power operations on November 23, 2005.
AmerenUE operated the plant at full power for 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
Readiness for Seasonal Susceptibilities
The inspectors completed a review of the licensee's readiness of seasonal
susceptibilities involving extreme low temperatures. The inspectors: (1) reviewed plant
procedures, the Final Safety Analysis Report (FSAR), and Technical Specifications (TS)
to ensure that operator actions defined in adverse weather procedures maintained the
readiness of essential systems; (2) walked down portions of the two systems listed
below to ensure that adverse weather protection features (heat tracing, space heaters,
weatherized enclosures, temporary chillers, etc.) were sufficient to support operability,
including the ability to perform safe shutdown functions; (3) evaluated operator staffing
levels to ensure the licensee could maintain the readiness of essential systems required
by plant procedures; and (4) reviewed the corrective action program to determine if the
licensee identified and corrected problems related to adverse weather conditions.
November 17, 2005: Essential service water pump house, Trains A and B
Documents reviewed by the inspectors included:
Procedure OTS-ZZ-00007, Plant Cold Weather, Revision 10
Procedure OTN-QJ-00003, Plant Freeze Protection Heat Tracing Procedure,
Revision 3
The inspectors completed one sample.
b.
Findings
No findings of significance were identified.
Enclosure
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1R04
Equipment Alignment (71111.04)
Partial Walkdowns
a.
Inspection Scope
The inspectors: (1) walked down portions of three risk important systems and reviewed
plant procedures and documents to verify that critical portions of the selected systems
were correctly aligned; and (2) compared deficiencies identified during the walkdown to
the licensee's FSAR and corrective action program to ensure problems were being
identified and corrected.
October 17, 2005, Emergency diesel generator (EDG), Train A
November 8, 2005, Containment recirculation sump, Train A
December 21, 2005, Centrifugal charging pump, Train A
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed three samples.
b.
Finding - Inadequate Inspection of the Containment Recirculation Sump
Introduction: The NRC identified a Green noncited violation (NCV) of 10 CFR Part 50,
Appendix B, Criterion V, for the failure to adequately implement work order instructions
and a procedure for inspection of the containment recirculation sump enclosure. The
licensees inspections failed to identify a 1.5-inch hole in the sump cover which could
provide a path for foreign material to enter into the containment sump emergency core
cooling system (ECCS) containment recirculation sump.
Description: On November 8, 2005, the inspectors identified a 1.5-inch hole penetrating
the containment recirculation Sump A ceiling. FSAR Section 6.2.2.1.2.2 stated that the
recirculation sumps are covered with the concrete pads supporting the accumulator
tanks; thus, debris cannot fall directly upon the screening structure. FSAR
Table 6.2.2-1 established a maximum c-inch gap for the sump screen. The screen
prevents the introduction of foreign material and debris that could degrade long-term
core cooling during an ECCS recirculation mode of operation. NRC Bulletin 2003-01,
Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-
Water Reactors, alerted the licensee to the susceptibility of recirculation sump failures.
AmerenUEs August 8, 2003, response to the bulletin included a commitment to inspect
the containment sumps and verify screen gap tolerances. The Callaway quality control
technicians' detailed inspection on April 27, 2004 (Work Package W229952), did not
identify the 1.5-inch hole. During Refueling Outage RF-14, AmerenUE performed
Procedure OSP-EJ-00003, Containment Recirculation Sump Inspection, Revision 5,
that required quality control and operations personnel to verify that all sump
penetrations were sealed prior to reactor startup. This inspection performed on
November 8, 2005, did not identify the hole in the containment sump cover.
Enclosure
-8-
Analysis: The performance deficiency associated with this finding involved licensee
personnel failure to effectively inspect the containment sump to assure any opening or
gaps in the sump cover were in accordance with the design basis. This finding was
greater than minor because it is associated with the mitigating systems cornerstone
attribute of equipment performance and affects the associated cornerstone objective to
ensure availability and reliability of the containment recirculation sump ECCS safety
function. Using the Manual Chapter 0609, Significance Determination Process,
Phase 1 Worksheet, this finding is determined to have very low safety significance
because the condition is a qualification deficiency confirmed not to result in loss of
operability per Part 9900, Technical Guidance, Operability Determination Process for
Operability and Functional Assessment. The cause of this finding is related to the
crosscutting element of human performance in that personnel failed to adequately
implement a work instruction and procedure for inspecting the containment sump
configuration.
Enforcement: The inspectors identified an NCV of 10 CFR Part 50, Appendix B,
Criterion V, "Instructions, Procedures, and Drawings," because AmerenUE did not
properly implement work instructions and a test procedure for inspecting the ECCS
containment sump. Contrary to
verify conformance of
containment Sump A. The corrective actions to restore compliance included repair of
the hole and actions taken to improve inspection techniques. Because of the very low
safety significance and the licensees action to place this issue in their corrective action
program as Callaway Action Request (CAR) 200509189, this violation is being treated
as an NCV in accordance with Section VI.A.1 of the Enforcement
Policy (NCV 05000483/2005005-01).
1R05
Fire Protection (71111.05)
a.
Inspection Scope
Quarterly Inspection
The inspectors walked down the nine listed plant areas to assess the material condition
of active and passive fire protection features and their operational lineup and readiness.
The inspectors: (1) verified that transient combustibles and hot work activities were
controlled in accordance with plant procedures; (2) observed the condition of fire
detection devices to verify they remained functional; (3) observed fire suppression
systems to verify they remained functional and that access to manual actuators was
unobstructed; (4) verified that fire extinguishers and hose stations were provided at their
designated locations and that they were in a satisfactory condition; (5) verified that
passive fire protection features (electrical raceway barriers, fire doors, fire dampers,
steel fire proofing, penetration seals, and oil collection systems) were in a satisfactory
material condition; (6) verified that adequate compensatory measures were established
for degraded or inoperable fire protection features and that the compensatory measures
were commensurate with the significance of the deficiency; and (7) reviewed the FSAR
to determine if the licensee identified and corrected fire protection problems.
Enclosure
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September 25, 2005, Fire Area RB, Reactor building
November 10, 2005, Fire Area RB, Reactor building
November 22, 2005, Fire Area A-2, ECCS, Train A
November 22, 2005, Fire Area A-4, ECCS Rooms, Train A
November 22, 2005, Fire Area A-9, Residual heat removal (RHR) heat
exchanger room, Train A
November 22, 2005, Fire Area A-10, RHR heat exchanger room, Train B
November 30, 2005, Fire Area C-9, Switchgear room, Train A
November 30, 2005, Fire Area C-10, Switchgear room, Train B
November 30, 2005, Fire Area D-1, Diesel generator, Train A
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed nine samples.
b.
Finding - Failure to Adequately Implement Continuous Compensatory Fire Watches
Introduction: The inspectors identified a noncited violation of TS 5.4.1.d, Fire
Protection Program Implementation, associated with seven examples of inadequately
performed continuous fire watches.
Description: Procedure APA-ZZ-0703, Fire Protection Operability Criteria and
Surveillance Requirements, required AmerenUE to establish compensatory continuous
watches in specified fire areas as a result of degraded fire detection or suppression
capability. The continuous fire watch is an uninterrupted observation post within a single
fire area. The physical presence of fire watch personnel provides reasonable assurance
that a fire would be prevented through prompt recognition and disposition of fire
hazards. If a fire occurred, despite these efforts, fire watch personnel would promptly
detect, report, and extinguish the fire while still in the incipient stage.
Procedure SDP-KC-00001, Requirements for and Duties of Compensatory Fire
Watches, Revision 5, required fire watches to maintain watch over the entire assigned
space with a minimum of patrolling.
In September 2005, AmerenUE provided verbal guidance to fire watch personnel that
continuous watch requirements may be met by a 15-minute roving fire patrol. Callaway
Facility Operating License, Amendment 169 (5) (d), required that changes that
adversely affect the ability to achieve and maintain safe shutdown in the event of a fire
receive prior NRC approval. The inspectors concluded that reducing continuous watch
requirements to a 15-minute roving patrol adversely affected the ability to achieve and
Enclosure
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maintain safe shutdown in the event of a fire. The inspections identified seven
examples of compensatory continuous fire watches where one fire watch person was
assigned simultaneously to multiple fire areas and building levels:
Date
Number
Continuous Fire Areas Concurrently
Watched by a Single Individual
September 5
12260
A-1, A-8, A11, A12, A-24, A-25
September 6
12260
A-1, A-8, A11, A12, A-24, A-25
September 7
12260
A-1, A-8, A11, A12, A-24, A-25
September 8
12260
A-1, A-8, A11, A12, A-24, A-25
September 25
12269
A-1, A-8, A11, A12, A-24, A-25
September 26
12269
A-1, A-8, A11, A12, A-24, A-25
September 30
12244
A-1, A-13, A-14, A-15
Analysis: The performance deficiency associated with this finding involved the failure of
AmerenUE to establish adequate continuous fire watches. This finding is greater than
minor because this finding was associated with the reactor safety mitigating systems
cornerstone attribute to provide protection against external factors and affects the
associated cornerstone objective to ensure the availability, reliability, and capability of
systems that respond to initiating events to prevent undesirable consequences. The
inspectors used Manual Chapter 0609, Appendix F, Fire Protection Significance
Determination Process, to analyze this finding because the condition had an adverse
affect on the Fixed Fire Protection Systems element of fire watches posted as a
compensatory measure for outages or degradations. A low degradation rating was
assigned to this finding as the provision affected by this finding is expected to display
nearly the same level of effectiveness and reliability. Using Manual Chapter 0609,
Appendix F, this finding is determined to have very low safety significance. The
inspectors concluded that the new guidance created situations which resulted in
inadequate compensatory fire watch coverage. The cause of this finding is related to
the crosscutting element of human performance in that the guidance was not adequate
to ensure continuous fire watches were appropriately implemented.
Enforcement: Callaway Plant Technical Specification 5.4.1.d, Fire Protection Program
Implementation, required that the Fire Prevention Program be implemented and
maintained per written procedures. The Fire Prevention Program requirements for fire
watches were implemented by Procedure SDP-KC-00001, Requirements for and Duties
of Compensatory Fire Watches, Revision 5. Procedure SDP-KC-00001 established a
requirement for compensatory continuous watches within specified fire areas as a result
of degraded fire detection or suppression capability. Contrary to
Procedure SDP-KC-00001 and the fire program, AmerenUE failed to perform
compensatory continuous watches within certain specified fire areas with degraded fire
detection or suppression capability between September 5 and 30, 2005. Because this
finding is of very low safety significance and was entered into the licensee's corrective
Enclosure
-11-
action program (CAR 200510325), it is being treated as an NCV, consistent with
Section VI.A of the NRC Enforcement Policy (NCV 05000483/2005005-02).
1R07
Heat Sink Performance (71111.07)
a.
Inspection Scope
The inspectors reviewed licensee programs, verified performance against industry
standards, and reviewed critical operating parameters and maintenance records for the
containment cooler heat exchangers. The inspectors verified that: (1) performance
tests were satisfactorily conducted for heat exchangers/heat sinks and reviewed for
problems or errors; (2) the licensee utilized the periodic maintenance method outlined in
Electric Power Research Institute NP-7552, Heat Exchanger Performance Monitoring
Guidelines; (3) the licensee properly utilized biofouling controls; (4) the licensees heat
exchanger inspections adequately assessed the state of cleanliness of their tubes, and
(5) the heat exchanger was correctly categorized under the maintenance rule.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b.
Findings - Indeterminate Containment Cooler Operability and Heat Removal Capability
Introduction: An unresolved item was identified for containment cooler heat removal
capability. AmerenUE will provide the inspectors additional testing results to complete
the inspection. This issue will remain unresolved pending additional review by the
inspectors. No analysis or enforcement reviews were performed for this unresolved
item.
Description: The inspectors reviewed available containment cooler testing data but
were not able to confirm that the heat exchangers were capable of the design bases
heat removal duty. FSAR Section 6.2.1.3, Mass and Energy Release Analyses for
Postulated Loss-of-Coolant Accidents, and Section 6.2.1.4, Mass and Energy Release
Analysis for Postulated Secondary Pipe Ruptures Inside Containment, stated that a
containment cooler duty of 141 million British Thermal Units per hour, at 277EF was
used in the accident analysis. TS Surveillance Bases 3.6.6.7, Containment Spray and
Cooling Systems, stated that the heat removal capability of each cooler train was
verified on an 18-month frequency. TS bases, Figure B.3.6.6-1, Containment Cooler
Heat Removal Minimum Cooling Flow Rate, established the minimum heat removal
capability as a function of essential service water (ESW) flow, assuming no fouling, to
meet design bases requirements. Plant engineering monitored ESW flow but not heat
removal capability. AmerenUE committed, by letter, Response to Generic Letter 89-13,
Service Water System Problems Affecting Safety Related Equipment, January 29,
1990, to verify the heat transfer capability of all safety-related heat exchangers cooled
by ESW. In addition, AmerenUE also committed to trend and compare the containment
cooler heat removal rates to the design requirements to promote identification of
degraded cooling equipment. Title 10 of the Code of Federal Regulations, Part 50,
Appendix B, Test Control, required AmerenUE to establish a test program to assure
Enclosure
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that the containment cooler's performance satisfactorily met acceptance limits
established in applicable design documents. Based on the information provided by
AmerenUE, the inspectors were not able to conclude that the containment coolers were
capable of removing design basis heat loads.
AmerenUE identified high differential pressure across the ESW side of containment
Cooler SGN01A on May 17, 2004 (Refueling Outage 14 Work Document P701990).
The high differential pressure was indicative of heat exchanger degradation due to
macrofouling. AmerenUE restarted and operated the plant until September 17, 2005,
without adequately assessing the affect of fouling on heat exchanger performance.
AmerenUE cleaned the heat exchanger during Refueling Outage 14. AmerenUE did not
perform testing prior to the cleaning to determine if any additional degradation had
occurred during the 18-month operating cycle. The inspectors were not able to verify,
based on the documentation reviewed, that the heat exchanger was capable of
performing the design bases function during Cycle 14. This issue is considered
unresolved pending additional NRC review of AmerenUE containment cooler testing
(Unresolved Item 05000483/2005005-03).
1R11
Licensed Operator Requalification Program (71111.11Q and 71111.11B)
.1
Quarterly Inspections
a.
Inspection Scope
The inspectors observed testing and training of senior reactor operators and reactor
operators to identify deficiencies and discrepancies in the training, to assess operator
performance, and to assess the postexercise critique. The inspectors observed a Just
In-Time Reactor Startup training scenario conducted on November 13, 2005.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b.
Findings
No findings of significance were identified.
.2
Biennial Inspection
a.
Inspection Scope
To assess the performance effectiveness of the licensed operator requalification
program, the inspectors conducted both on-site and in-office reviews involving personnel
interviews, operating and written examinations, and operating examination activities.
During the on-site review, the inspectors interviewed five licensee personnel, consisting
of three instructors, one operator and a training supervisor, to determine their
understanding of the policies and practices for administering requalification
Enclosure
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examinations. The inspectors also reviewed operator performance on the written and
operating examinations. These reviews included observations of portions of the
operating examination by the inspectors. The operating examinations observed
included job performance measures and four scenarios that were used in the current
biennial requalification cycle. These observations allowed the inspectors to assess the
licensee's effectiveness in conducting the operating test to ensure operator mastery of
the training program content.
The results of these examinations were reviewed to determine the effectiveness of the
licensees appraisal of operator performance and to determine if feedback of
performance analysis into the requalification training program was being accomplished.
The inspectors interviewed members of the training department and reviewed minutes
of training review group meetings to assess the responsiveness of the licensed operator
requalification program in incorporating the lessons learned from both plant and industry
events. Examination results were also assessed to determine if they were consistent
with the guidance contained in NUREG 1021, "Operator Licensing Examination
Standards for Power Reactors," Revision 9, and NRC Manual Chapter 0609, Appendix I,
"Operator Requalification Human Performance Significance Determination Process."
Additionally, the inspectors assessed the Callaway Plant-referenced simulator for
compliance with 10 CFR 55.46, "Simulator Facilities." This assessment included the
adequacy of the licensees simulation facility for use in operator licensing examinations
and for satisfying experience requirements as prescribed by 10 CFR 55.46. In addition,
the inspectors reviewed selected applicant personnel qualitative statements (NRC
Form 398) to verify their accuracy. During the Form 398 reviews, the inspectors noted
that several applicants were given credit for reactivity and control manipulations on the
simulator instead of on the actual plant. While this simulator usage is permitted by
10 CFR 55.46, the simulator must meet the standards of fidelity as required by
10 CFR 55.46(c)(2). Based on this observation and the requirements of 10 CFR 55.46,
the inspectors expanded their review of the simulator testing. This review expansion
included a review of the simulator annual performance test book. The inspectors
reviewed a sample of simulator performance test records (transient tests, surveillance
tests, and malfunction tests), simulator deficiency report records, and processes for
ensuring simulator fidelity commensurate with 10 CFR 55.46. The inspectors reviewed
selected simulator deficiency reports generated by the licensee that did not result in
changes to the configuration of the simulator to assess the responsiveness of the
licensee's simulator configuration management program. The inspectors also
interviewed members of the licensees simulator configuration control group as part of
this review.
During the in-office review, the inspectors evaluated whether the written examination
was developed and administered in accordance with the standards described in
NUREG 1021 and evaluated any issues identified in accordance with NRC Manual
Chapter 0609, Appendix I. The written examination review was focused on quality
aspects of the examination, such as discrimination validity, examination question
psychometric quality, and examination integrity.
Enclosure
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b.
Findings
1.
Evaluation of the Written Examination
As a result of the review of the written requalification examinations, the inspectors
identified that the quality of the examinations developed by the licensee appeared to not
meet the guidance set forth in NUREG 1021, ES-602, Attachment 1, Section B, "Open-
Reference Guidelines." The term "open reference" means that the candidates are
allowed to use any reference to assist them when taking the examination.
Since the operators are allowed to use examination question references while taking the
examination, test questions should be developed that do more than test for mere recall
and/or memorization. Open-reference questions should have the operators
demonstrate an understanding of an issue by using their knowledge to address real-life
situations and problems. The NUREG further states with regard to direct look up
questions that removing from the stem of the question any information that cues the
operator to the answer's location does not make the question acceptable.
With regard to the open-reference questions, the NUREG also addresses "Direct
Lookup" questions. Direct lookup questions only test memory because the information
is readily available. This is a less valid means of testing candidate knowledge and only
demonstrates that a candidate knows where to find information. Therefore, the
discrimination validity of the question is critical to differentiate the safe operator from the
unsafe operator.
Additionally, other than demonstrating that a candidate knows where to find information,
the licensees biennial requalification examinations appeared to not test the
understanding or analysis of the information that would be applied on the job. These
issues will be reviewed as Unresolved Item (URI)05000483/2005005-04, Adequacy of
the Biennial Requalification Written Examination (CAR 200600528).
2.
Simulation Facility Performance
Introduction: During a review of the simulator annual performance test book,
the inspectors identified a Green finding for the failure to conduct simulator performance
testing in accordance with ANSI/ANS 3.5, "Nuclear Power Plant Simulators for use in
Operator Training and Examination," 1998.
Description: A review of the Steady State and Normal Evolution tests contained in the
annual performance test book for the simulator revealed that the licensee did not
compare all of the required parameters listed in ANSI 3.5-1998 to actual plant data;
specifically, Thot, Tcold, core megawatt thermal, steam flow, feed flow, letdown flow,
charging flow, and turbine first stage pressure. In lieu of this comparison, the licensee
utilized an "expert panel review" to determine if the simulator operation mimics the
actual plant. When the inspectors requested the baseline data to support the analysis
documentation, the licensee was unable to provide the data. The licensee stated that
the analysis was done by a panel of experts and that the signature on the meeting
minutes constituted the required analysis and baseline data. The 1998 version of
Enclosure
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ANSI/ANS 3.5, requires that the annual simulator performance tests be conducted such
that the key parameters listed in Appendix B of this standard are recorded and that
these records be compared to actual or reference plant data (if available) or engineering
data from the FSAR. If such engineering data is not available in the FSAR, the standard
permits the use of data from subject matter expert estimates to determine acceptability
of the test.
Analysis: The inspectors determined that the failure to adhere to ANSI/ANS 3.5-1998,
as endorsed by Regulatory Guide 1.149, "Nuclear Power Plant Simulation Facilities for
Use in Operator Training and License Examinations," Revision 3, October 2001, as
committed to in the Callaway Plant Simulation certification dated March 13, 2000, was a
performance deficiency. Specifically, the simulator performance testing did not meet the
standards specified in ANSI/ANS 3.5-1998 in that: (1) all required parameters during
the simulator test were not recorded; and (2) simulator to baseline data comparisons
were unavailable.
The NRC has determined that traditional enforcement does not apply because the issue
did not have any actual safety consequence or potential for affecting the NRCs
regulatory function and did not result in any willful violation of NRC requirements or
licensee procedures. The performance deficiency is more than minor because it
affected the ability of the simulator transient tests to detect fidelity issues with the
simulator and affects the Human Performance (Human Error) attribute of the Initiating
Events and Mitigating Systems cornerstones.
Enforcement: No violation of regulatory requirements occurred. The examiners
determined that the finding did not represent a noncompliance because Callaway Plant
performed some testing even though the testing was not sufficient in scope and
because no actual events have occurred that could be attributed to a lack of simulator
fidelity testing: Finding (FIN)05000483/2005005-05, Failure to Conduct Simulator
Testing in Accordance with ANSI/ANS 3.5-1998 (CAR 200600527).
3.
Adequacy of Plant-Referenced Simulator to Conform with Simulator Requirements for
Reactivity and Control Manipulation Credits
As the result of reviewing NRC Form 398, the inspectors noted that the licensee
used the simulator to meet reactivity and control manipulation experience requirements
for initial operator and senior operator license applicants in accordance with
10 CFR 55.46(c)(2)(ii). For the manipulations, the licensee used a single page sign-off
sheet for documentation. To use the simulator for reactivity and control manipulation
credit, the regulation requires that significant control manipulations are completed
without procedural exceptions, simulator performance exceptions, or deviation from the
approved training scenario sequence. Furthermore, the ANSI standard requires that
these items be performed without offsets in the simulator and without time-compression
techniques that expected alarms are generated as required in real time with no
unexpected alarms generated during the scenario sequence. The documentation
provided could not be used to verify each of the requirements as specified in the
regulations and standards.
Enclosure
-16-
The safety significance of this issue could be more than minor due to the apparent
failure to meet the requirements of 10 CFR 55.46(c)(2)(ii) with regard to assuring
maintenance of the plant referenced simulator fidelity. Accordingly, a URI was opened
pending further review of the simulator in subsequent inspections. The licensee entered
this issue into their corrective action program as CAR 200600529: URI 05000483/
200505-06, Adequacy of Plant-Referenced Simulator to Conform with Simulator
Requirements for Reactivity and Control Manipulation Credits.
1R12
Maintenance Effectiveness (71111.12Q)
a.
Inspection Scope
The inspectors reviewed the two listed maintenance activities to: (1) verify the
appropriate handling of structures, systems, and components (SSC) performance or
condition problems; (2) verify the appropriate handling of degraded SSC functional
performance; (3) evaluate the role of work practices and common cause problems; and
(4) evaluate the handling of SSC issues reviewed under the requirements of the
maintenance rule, 10 CFR Part 50, Appendix B, and the TSs.
September 30, 2005, CAR 200507636, Missing spring in ventilation door
solenoid lock assembly
August 2, 2005, CAR 200505344, Fuel building roll-up door
Documents reviewed by the inspectors included:
Procedure EDP-ZZ-01128, Maintenance Rule Program, Revision 6
The inspectors completed two samples.
b.
Findings
No findings of significance were identified.
1R13
Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)
a.
Inspection Scope
Risk Assessment and Management of Risk
The inspectors reviewed the three listed assessment activities to verify:
(1) performance of risk assessments when required by 10 CFR 50.65 (a)(4) and
licensee procedures prior to changes in plant configuration for maintenance activities
and plant operations; (2) the accuracy, adequacy, and completeness of the information
considered in the risk assessment; (3) that the licensee recognizes, and/or enters as
Enclosure
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applicable, the appropriate licensee-established risk category according to the risk
assessment results and licensee procedures; and (4) the licensee identified and
corrected problems related to maintenance risk assessments.
October 17, 2005, Essential power, Train B, planned outage, in-office review
October 31, 2005, Spent fuel pool time-to-boil method, in-office review
November 21, 2005, Unplanned emergent maintenance on ESW inlet isolation
Valve EFHV52, in-office review
Documents reviewed by the inspectors included:
Procedure EDP-ZZ-01128, Maintenance Rule Program, Revision 6
Procedure EDP-ZZ-01129, Callaway Plant Risk Assessment, Revision 8
Procedure ODP ZZ 00001, Operations Department - Code of Conduct,
Revision 23
The inspectors completed three samples.
Emergent Work Control
The inspectors: (1) verified that the licensee performed actions to minimize the
probability of initiating events and maintained the functional capability of mitigating
systems and barrier integrity systems; (2) verified that emergent work-related activities
such as troubleshooting, work planning/scheduling, establishing plant conditions,
aligning equipment, tagging, temporary modifications (TMs), and equipment restoration
did not place the plant in an unacceptable configuration; and (3) reviewed the FSAR to
determine if the licensee identified and corrected risk assessment and emergent work
control problems.
October 17, 2005, Essential power, Train B, planned outage. The inspectors
observed compensatory risk mitigation actions from the control building and
completed an in-office review.
November 21, 2005, ESW inlet isolation Valve EFHV52. The inspectors
observed compensatory risk mitigation actions from the control building and
completed an in-office review.
Documents reviewed by the inspectors included:
Nuclear Management and Resource Council 93-01, Industry Guidelines for
Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 3
Procedure EDP-ZZ-01129, Callaway Plant Risk Assessment, Revision 9
Enclosure
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The inspectors completed two samples.
b.
Findings
No findings of significance were identified.
1R14
Personnel Performance During Nonroutine Plant Evolutions (71111.14)
a.
Inspection Scope
The inspectors: (1) reviewed operator logs, plant computer data, and/or strip charts for
the below listed evolutions to evaluate operator performance in coping with nonroutine
events and transients; (2) verified that operator actions were in accordance with the
response required by plant procedures and training; (3) attended and/or reviewed
postevent critic meetings; and (4) verified that the licensee has identified and
implemented appropriate corrective actions associated with personnel performance
problems that occurred during the nonroutine evolutions sampled.
November 7, 2005, CAR 200509143, Rapid pressurizer surge line cooldown due
to melting lead blankets
November 8, 2005, CAR 200509191, Pressurizer surge line heatup rate
exceeded
November 14, 2005, CAR 200509345, Unplanned securing of the steam dumps
and subsequent reactor coolant system (RCS) heatup with initiating RCS
temperature at 340EF
November 15, 2005, Plant cooldown to remove a shim on Steam Generator D
Documents reviewed by the inspectors included:
Procedure OTG-ZZ-00001, Plant Heatup, Cold Shutdown to Hot Standby,
Revision 46
Procedure APA-ZZ-00500, Corrective Action Program, Revision 38
Procedure OSP-BB-00007, RCS Heatup and Cooldown Limitations, Revision 9
The inspectors completed four samples.
Enclosure
-19-
b.
Finding
Failure to Follow Procedures Resulted in Violation of RCS Cooldown and Heatup Rate
Limits
Introduction. The inspectors identified a Green NCV of TS 5.4.1.a, Procedures, after
AmerenUE operations personnel failed to maintain the RCS temperature limits on two
occasions.
Description. On November 7, 2005, plant operators terminated a plant heatup and
decreased the RCS pressurizer surge line temperature 260EF in one hour. The
operators initiated the rapid cooldown by isolating pressurizer auxiliary spray, resulting in
an in-surge of cooler RCS water. The operators conducted the rapid cooldown after
several containment lead shield blanket polyvinylchloride covers in containment
unexpectedly melted. The shield blankets had not been removed from the uninsulated
pressurizer surge line prior to plant heatup due to a work scheduling error. The licensee
identified a second example of excessive surge line temperature on November 8, 2005.
Plant operators increased the surge line temperature about 175EF in one hour during a
plant heatup.
TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, required temperature changes
of all RCS components (except the pressurizer) be limited to 100EF in one hour. The
TS Bases defined the surge line as part of the RCS. General Operating
Procedure OTG-ZZ-00001, Plant Heatup, Cold Shutdown to Hot Standby, required
operating personnel maintain greater than 5 gpm auxiliary spray and a pressurizer
outsurge. Procedure OSP-BB-00007, RCS Heatup and Cooldown Limitations,
required that RCS temperature changes not exceed 100EF in one hour during
cooldown/heatup evolutions. The inspectors identified that operations personnel failed
to recognize the applicability of the TS and apply the appropriate TS action statement.
Analysis: The performance deficiency associated with this finding involved failure of
operations personnel to follow established procedures and recognize the appropriate TS
action. This finding was greater than minor because it is associated with the reactor
safety barrier integrity cornerstone attribute of equipment performance and affects the
associated cornerstone objective to ensure reasonable assurance that the RCS piping
barrier will protect the public from radionuclide releases caused by accidents or events.
Using Manual Chapter 0609, Significance Determination Process, Appendix G,
Shutdown Operations, this finding was determined to have very low safety significance
because, based on the engineering evaluation of RCS thermal stress resulting from the
temperature transients, the condition did not significantly increase the likelihood of a
loss of RCS inventory and did not degrade the licensees ability to terminate a leak path.
The cause of this finding is related to the crosscutting element of human performance
because of personnel failure to follow procedures.
Enforcement: TS 5.4.1.a, Procedures, required that written procedures be
established, implemented, and maintained covering the activities specified in
Appendix A, Typical Procedures for Pressurized Water Reactors, of Regulatory
Guide 1.33, Quality Assurance Program Requirements (Operation), February 1978.
Enclosure
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Regulatory Guide 1.33, Appendix A, Section 2a, required general plant operating
procedures for cold shutdown to hot standby to be implemented. Entry into TS 3.4.3,
RCS Pressure and Temperature (P/T) Limits, action was required when an RCS
component temperature transient exceeded 100EF cooldown and or heatup limit within a
one-hour period. Callaway Procedure OSP-BB-00007, RCS Heatup and Cooldown
Limitations, required that RCS temperature changes shall not exceed 100EF in one
hour during cooldown or during heatup evolutions. Contrary to these requirements, on
November 7 and 8, 2005, operations personnel did not maintain the RCS temperature
rate less than 100EF within one hour. Because of the very low safety significance and
the licensees action to place this issue in their corrective action program as
CARs 200509487 and 200509143, this violation is being treated as an NCV in
accordance with Section VI.A.1 of the Enforcement Policy (NCV 50-483/2005005-07).
1R15
Operability Evaluations (71111.15)
a.
Inspection Scope
The inspectors: (1) reviewed plant status documents such as operator shift logs,
emergent work documentation, deferred modifications, and standing orders to
determine if an operability evaluation was warranted for degraded components;
(2) referred to the FSAR and design basis documents to review the technical adequacy
of licensee operability evaluations; (3) evaluated compensatory measures associated
with operability evaluations; (4) determined degraded component impact on any TSs;
(5) used the significance determination process to evaluate the risk significance of
degraded or inoperable equipment; and (6) verified that the licensee has identified and
implemented appropriate corrective actions associated with degraded components.
Operability Determination 200509277, Overpressurization of the turbine-driven
auxiliary feedwater pump (TDAFP) during the backleakage test of its discharge
Operability Determination 200509374, Pressurizer power-operated relief valve
stroke time basis
Operability Determination 200509368, Excessive stroke time of feedwater
isolation Valve AEFV0040
Operability Determination 200505062, Insufficient time to transfer ECCS and
containment spray to cold leg recirculation
Operability Determination 2005003773, Degraded containment cooler heat
removal capability
The inspectors completed five samples.
b.
Findings
No findings of significance were identified.
Enclosure
-21-
1R16
Operator Workarounds (71111.16)
a.
Inspection Scope
Selected Operator Workarounds
The inspectors reviewed the two listed operator workarounds to: (1) determine if the
functional capability of the system or human reliability in responding to an initiating event
is affected; (2) evaluate the effect of the operator workaround on the operators ability to
implement abnormal or emergency operating procedures; and (3) verify that the
licensee has identified and implemented appropriate corrective actions associated with
operator workarounds.
November 23, 2005, In-office review of the degradation of main steam line
Monitor 16
November 23, 2005, Maintenance repair of Bistable SB069 and permissive
indicating panel
Documents reviewed by the inspectors included:
December 2005, Operator Work Around and Burdens list
Procedure APA-ZZ-00018, Conduct of Operations - Quality Control, Revision 7
Procedure ODP-ZZ-00001, Operations Department - Code of Conduct,
Revision 25
The inspectors completed two samples.
Cumulative Review of the Effects of Operator Workarounds
The inspectors reviewed the cumulative effects of operator workarounds to determine:
(1) the reliability, availability, and potential for misoperation of a system; (2) if multiple
mitigating systems could be affected; (3) the ability of operators to respond in a correct
and timely manner to plant transients and accidents; and (4) if the licensee has
identified and implemented appropriate corrective actions associated with operator
workarounds.
The inspectors reviewed the Operator Workaround and Burdens List.
The inspectors completed one sample.
b.
Findings
No findings of significance were identified.
Enclosure
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1R17
Permanent Plant Modifications (71111.17)
a.
Inspection Scope
Annual Review
The inspectors reviewed key affected parameters associated with energy needs,
materials/replacement components, timing, heat removal, control signals, equipment
protection from hazards, operations, flowpaths, pressure boundary, ventilation
boundary, structural, process medium properties, licensing basis, and failure modes for
the modification listed below. The inspectors verified that: (1) modification preparation,
staging, and implementation did not impair emergency/abnormal operating procedure
actions, key safety functions, or operator response to loss of key safety functions;
(2) postmodification testing maintained the plant in a safe configuration during testing by
verifying that unintended system interactions will not occur, SSC performance
characteristics still meet the design basis, the appropriateness of modification design
assumptions, and the modification test acceptance criteria has been met; and (3) the
licensee has identified and implemented appropriate corrective actions associated with
permanent plant modifications.
November 1, 2005, Modification MP 05-3051, Containment Sump
Valves EJHV8811A and EJHV8811B. The inspectors performed an in-office
review and performed a walkdown of the affected equipment in the auxiliary
building.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b.
Findings - Use of a Nonqualified Calculation in a Safety Related Modification
Introduction. The NRC identified a Green NCV of 10 CFR Part 50, Appendix B,
Criteria V, Instructions, Procedures, and Drawings, associated with an inadequate
engineering procedure used to verify calculations. The inadequate procedure resulted
in the use of a nonqualified, nonsafety-related engineering calculation to demonstrate
the safety function of the containment recirculation sump valves following a modification.
Description: AmerenUE failed to ensure a nonsafety-related vendor supplied calculation
was qualified before use to demonstrate the design bases function of safety-related
components after a modification. AmerenUE identified that maximum postaccident
differential pressure assumed between the containment recirculation sump and RHR
system was incorrect. Based on industry operational experience (OE), engineering
determined the maximum design differential pressure the containment sump valve
operators would have to open against increased from 53 pounds per square inch
differential (psid) to 468 psid. The Engineering Department generated Modification
MP 05-3051, Containment Sump Valves EJHV8811A and EJHV8811B, to increase
valve operator opening torque. To support the modification, AmerenUE purchased
nonsafety-related Calculation KCI 330-001-DC1, Revision 0, October 18, 2005, from a
Enclosure
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vendor. The calculation used a new, realistic approach to establish the maximum valve
operator torque. Plant engineering used the operator torque developed from this
calculation to ensure the sump valves would open against the higher differential
pressure after modification. Engineering personnel used Procedure EDP-ZZ-04023,
Calculations, Revision 17, to qualify the vendor supplied calculation before approved
use in the safety-related application. Procedure EDP-ZZ-04023 provided insufficient
detail to enable engineering personnel to verify the design by either an alternate method
or suitable test program to qualify the nonsafety-related calculation.
Analysis: The performance deficiency associated with this finding involved the failure of
engineering personnel to only use qualified calculations for safety-related modifications.
This finding is greater than minor because, if left uncorrected, this finding would become
a more significant safety concern affecting other safety-related modifications. This
finding affected the mitigating systems cornerstone. Using the Manual Chapter 0609,
Significance Determination Process, Phase 1 Worksheet, this finding is determined to
have very low safety significance because this finding involves a design deficiency
confirmed not to result in loss of operability per Part 9900, Technical Guidance,
Operability Determination Process for Operability and Functional Assessment. The
cause of this finding is related to the crosscutting element of human performance in that
the procedure did not ensure the calculations were qualified to support a design basis
function of a safety-related component.
Enforcement: Title 10 of the Code of Federal Regulations, Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, required that activities affecting
quality be prescribed by documented instructions or procedures appropriate to the
circumstances. Contrary to this, Procedure EDP-ZZ-04023, required for an activity
affecting quality, was not appropriate to the circumstances. Specifically, on October 28,
2005, Procedure EDP-ZZ-04023 was not adequate to ensure the qualification of
nonsafety-related Calculation KCI 330-001-DC1, Revision 0, before use in
Calculation EJ-42, Revision 0, an activity affecting quality. Because this finding is of
very low safety significance and was entered into AmerenUE's Corrective Action
Program (CAR 200509849), this violation is being treated as an NCV, consistent with
Section VI.A of the NRC Enforcement Policy (NCV 05000483/2005005-08).
1R19
Postmaintenance Testing (71111.19)
a.
Inspection Scope
The inspectors selected the six listed postmaintenance test (PMT) activities of risk
significant systems or components. For each item, the inspectors: (1) reviewed the
applicable licensing-basis and/or design-basis documents to determine the safety
functions; (2) evaluated the safety functions that may have been affected by the
maintenance activity; and (3) reviewed the test procedure to ensure it adequately tested
the safety function that may have been affected. The inspectors either witnessed or
reviewed test data to verify that acceptance criteria were met, plant impacts were
evaluated, test equipment was calibrated, procedures were followed, jumpers were
properly controlled, the test data results were complete and accurate, the test
equipment was removed, the system was properly re-aligned, and deficiencies during
Enclosure
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testing were documented. The inspectors also reviewed the FSAR to determine if the
licensee identified and corrected problems related to postmaintenance testing.
September 29, 2005, PMT W236012/920, Containment cooler train. The
inspectors observed the PMT from the reactor building and the control room and
performed an in-office review.
October 12, 13, and 14, 2005, PMTs 222071/912, and W715936/900, ESW
Train A, motor and pump replacement. The inspectors observed the PMT from
the ESW pump room and the control room and performed an in-office review.
October 12, 2005, PMTs W236513/900, W236509/940, and P711090/900, EDG
Train A, major overhaul. The inspectors observed the PMT from the EDG room
and the control room and performed an in-office review.
November 22, 2005, PMTs 05110929/200 and 05110929/910, TDAFP discharge
Check Valve ALHV0054. The inspectors observed the PMT from the auxiliary
building and the control room and performed an in-office review.
November 1, 2005, PMT P721574/910, Overhaul of NN Inverter 14. The
inspectors performed an in-office review.
December 28, 2005, PMT 05112449/900, Ultimate heat sink electrical room fan.
The inspectors performed an in-office review.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed six samples.
b.
Findings
No findings of significance were identified.
1R20
Refueling and Outage Activities (71111.20)
a.
Inspection Scope
The inspectors reviewed the following risk significant refueling items or outage activities
to verify defense-in-depth commensurate with the outage risk control plan, compliance
with the TSs, and adherence to commitments in response to Generic Letter 88-17, Loss
of Decay Heat Removal: (1) the risk control plan; (2) tagging/clearance activities;
(3) RCS instrumentation; (4) electrical power; (5) decay heat removal; (6) spent fuel pool
cooling; (7) inventory control; (8) reactivity control; (9) containment closure; (10) reduced
inventory or midloop conditions; (11) refueling activities; (12) heatup and cooldown
activities; (13) restart activities; and (14) licensee identification and implementation of
appropriate corrective actions associated with refueling and outage activities. The
Enclosure
-25-
inspectors' containment inspections included observations of the containment sump for
damage and debris, and supports, braces, and snubbers for evidence of excessive
stress, water hammer, or aging.
October 29, 2005, Precore alterations verifications
October 30, 2005, ECCS full flow test, from control room
October 31, 2005, Fuel handling from the reactor building and control room
October 31, 2005, Spent fuel pool time-to-boil method, in-office review
November 13, 2005, Containment closure walkdown
November 17, 2005, Reactor startup from the control room and the outage
control center
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed one sample.
b.
Findings - Less Than Adequate Spent Fuel Pool Water Inventory Risk Controls
Introduction
The inspectors identified a Green finding after AmerenUE implemented less than
adequate risk management controls of the spent fuel pool water inventory following
reactor core offload.
Description: The inspectors identified that AmerenUE had not implemented shutdown
risk administrative controls on the fuel transfer tube gate valve and the associated
flange during the period the fuel was offloaded to the spent fuel pool. On
September 29, 2005, the core had been off-loaded to the spent fuel pool and the
transfer canal weir gate was removed. In this configuration, the fuel transfer tube valve,
if opened, would provide a drain path from the spent fuel pool through an open weir wall.
Valve ECV-995 was closed but not identified in the shutdown risk management system
and did not have administrative controls to protect against misalignment. The licensee
provided the inspectors a calculation during the inspection that demonstrated that
Valve ECV-995 could be opened during the period of concern. AmerenUEs risk
guidelines, specified in Procedure APA-ZZ-00150, Appendix H, Project Risk
Management Guidelines, provided for measures to be in place to avoid risk. This
finding was entered into the Corrective Action Program as CARs 200507593
and 200507693.
Analysis: The performance deficiency associated with this finding involved failure of the
licensee to identify and implement inventory risk controls associated with the spent fuel
pool. This finding is greater than minor because, if left uncorrected, this condition could
become a more significant safety concern. NRC Information Notice 2005-16, Outage
Enclosure
-26-
Planning and Scheduling - Impacts on Risk, described operating experience related to
refueling risk management. Information Notice 2005-16 emphasized that most spent
fuel pool events had a common thread of human error and involved equipment
misalignment. NRC Manual Chapter 0609, Significance Determination Process, does
not specifically address findings related to the spent fuel pool inventory. Therefore, this
issue was evaluated by NRC management with input from a senior reactor analyst. This
finding was determined to be of very low safety significance based on the fact that the
procedure used to manipulate the valve was not in use during this period and that
borated water makeup capabilities were available to the spent fuel pool.
Enforcement: No violation of regulatory requirements occurred. The inspectors
determined that this finding did not represent a noncompliance because it did not involve
a safety-related or TS required procedure (FIN 05000483/2005005-09).
1R22
Surveillance Testing (71111.22)
a.
Inspection Scope
The inspectors reviewed the FSAR, procedure requirements, and TSs to ensure that the
seven listed surveillance activities demonstrated that the SSCs tested were capable of
performing their intended safety functions. The inspectors either witnessed or reviewed
test data to verify that the following significant surveillance test attributes were
adequate: (1) preconditioning; (2) evaluation of testing impact on the plant;
(3) acceptance criteria; (4) test equipment; (5) procedures; (6) jumper/lifted lead
controls; (7) test data; (8) testing frequency and method demonstrated TS operability;
(9) test equipment removal; (10) restoration of plant systems; (11) fulfillment of
American Society of Mechanical Engineers code requirements; (12) updating of
performance indicator data; (13) engineering evaluations, root causes, and bases for
returning tested SSCs not meeting the test acceptance criteria were correct;
(14) reference setting data; and (15) annunciators and alarms setpoints. The inspectors
also verified that the licensee identified and implemented any needed corrective actions
associated with the surveillance testing.
September 28, 2005, Surveillance S724682, Boric acid walkdown. The
inspectors observed portions of the walkdown in the reactor building and
completed an in-office review of the completed test documentation.
October 30, 2005, Surveillance S72279, ECCS check valve flow test. The
inspectors observed portions of the test from the reactor building and the control
room and completed an in-office review of the completed surveillance test
package.
November 1, 2005, Surveillance S05514649, RCS flow test. The inspectors
observed portions of the test from the control room and completed an in-office
review of the completed test documentation.
Enclosure
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November 16, 2005, Surveillance 05513671/500, TDAFP inservice test. The
inspectors observed portions of the testing in the auxiliary building and
completed an in-office review of the test documentation.
November 17, 2005, Surveillance 05511199, Estimated critical rod position. The
inspectors observed portions of the testing from the control room and completed
an in-office review of the test documentation.
November 17, 2005, Surveillance 726457, Low power physics test program with
dynamic rod worth measurement. The inspectors observed portions of the
testing from the control room and completed an in-office review of the test
documentation.
November 25, 2005, Surveillance 05101397, Feedwater isolation valve tests.
The inspectors observed portions of the testing from the control room and
auxiliary building and completed an in-office review of the test documentation.
Documents reviewed by the inspectors are listed in the attachment.
The inspectors completed seven samples.
b.
Findings
No findings of significance were identified.
1R23
Temporary Plant Modifications (71111.23)
a.
Inspection Scope
The inspectors reviewed the FSAR, plant drawings, procedure requirements, and TSs to
ensure that the three below listed TMs were properly implemented. The inspectors:
(1) verified that the modifications did not have an affect on system operability/availability;
(2) verified that the installation was consistent with modification documents; (3) ensured
that the postinstallation test results were satisfactory and that the impact of the
temporary modifications on permanently installed SSCs were supported by the test;
(4) verified that the modifications were identified on control room drawings and that
appropriate identification tags were placed on the affected drawings; and (5) verified that
appropriate safety evaluations were completed. The inspectors verified that licensee
identified and implemented any needed corrective actions associated with temporary
modifications.
November 15, 16, and 17, 2005, TM 05-0021, Reactor coolant pump vibration
circuit. The inspectors walked down portions of the TM in the control building
and completed an in-office review.
November 15, 16, and 17, 2005, TM ETP-SE-ST003, Reactivity computer for low
power physics testing. The inspectors walked down portions of the TM located
in the control building and completed an in-office review.
Enclosure
-28-
November 15, 16, and 17, 2005, TM ET-SE-ST003, Nuclear instrument channel
trip setpoints. The inspectors walked down portions of the TM located in the
control building and completed an in-office review.
Documents reviewed by the inspectors included:
Procedure ETP-SE-ST003, Precritical alignment/hookup of advanced digital
reactivity computer, Revision 6
Administrative Procedure APA-ZZ-00605, Temporary system modifications,
Revision 18
The inspectors completed three samples.
b.
Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level (EAL) and Emergency Plan Changes (71114.04)
a.
Inspection Scope
The inspectors performed in-office reviews of Revision 27 to the Callaway Plant
Radiological Emergency Response Plan, and Revision 33 to Procedure EIP-ZZ-00101,
Classification of Emergencies. These revisions were compared to their previous
revisions, to 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; to NEI 99-01, Methodology for Development of Emergency
Action Levels, Revision 2; and to the requirements of 10 CFR 50.47(b)(4) and 50.54(q)
to determine if the revisions decreased the effectiveness of the plan. These revisions:
Made minor administrative updates and corrections and updated titles
Clarified steam generator leakage terminology for a loss of containment in
EAL 2, Indicator 3b
Clarified verification of an earthquake in EAL 3H, Indicator 1c
Clarified the timeliness of classification with regard to validation of alarms
Clarified the definitions of steam generator leakage and faulted steam generator
as applied to fission product barriers
Revised the reactor coolant temperature threshold for a potential loss of
containment in EAL 2, Indicator 7b, based on updated engineering calculations
Enclosure
-29-
Revised the reactor vessel level threshold for potential loss of fuel cladding in
EAL 2, Indicator 6b, based on revised emergency operating procedures
Revised the description of telephone systems used in emergency response
facilities based on replacement of some phones
Added shelter as an option for recommendations of protective actions for the
general public
Added five special needs facilities in the emergency planning zone
Added descriptions of a safety significance fire to EAL 3E, and defined the time a
fire is out
The inspectors completed two samples during this inspection.
b.
Findings
Introduction: A violation of 10 CFR 50.54(q) was identified for implementation of a
decrease of effectiveness in the licensees emergency plan. The licensee implemented
a change to EAL 3E (Notification of Unusual Event) which defined a fire as having safety
significance only when it was located within 50 feet of vital areas, unless the smoke or
water stream from fighting the fire directly impacted listed safety-related equipment.
Description: The NRC identified that on June 8, 2005, the licensee implemented a
change to its EAL bases, which was an apparent decrease in effectiveness of the
licensees emergency plan, because it restricted applicability of EAL 3E, Fire within
Protected Area Boundary NOT Extinguished with 15 minutes of Verification.
Specifically, the revised bases clearly limited a plant fire adjacent to a vital area as one
that is within 50 feet of a vital area, except in cases where smoke or water from fighting
the fire directly affected safety-related equipment. The inspector determined that fires in
some plant areas, such as areas of the turbine building, which were classifiable under
EIP-ZZ-00101, Revision 32, may not have been classifiable using the revised EAL
bases.
Analysis: Implementation of changes to emergency action levels which decreased the
effectiveness of the emergency plan, was a performance deficiency. The finding had a
credible impact on the emergency preparedness cornerstone objective because a
licensee is less capable of implementing adequate measures to protect the health and
safety of the public during a radiological emergency if initiating conditions are removed
from licensee emergency action levels. This finding is more than minor because:
(1) restricting or limiting a classifiable condition in the licensee EALs has the potential to
impact safety; and (2) licensee implementation of a change to their emergency plan
which decreases the effectiveness of the plan without prior NRC approval impacts the
regulatory process. The finding also involves a violation of NRC requirements, subject
to enforcement action under the terms of the NRC Enforcement Policy.
Enclosure
-30-
Enforcement: Licensee implementation, without prior NRC approval, of an EAL change
which decreases the effectiveness of the emergency plan is a violation of
10 CFR 50.54(q), which states, in part, A licensee authorized to possess and operate a
nuclear power reactor shall follow and maintain in effect emergency plans that meet the
standards in §50.47(b) and the requirements in Appendix E of this part . . . The nuclear
power reactor licensee may make changes to these plans without Commission approval
only if the changes do not decrease the effectiveness of the plans and the plans, as
changed, continue to meet the standards of §50.47(b) and the requirements of
Appendix E to this part.
In accordance with Manual Chapter 0609, Appendix B, §2.2(e) and §4.4, the inspector
evaluated the significance of the finding using the General Statement of Policy and
Procedure for NRC Enforcement Actions (Enforcement Policy),Section IV,
Significance of Violations. The finding was determined to be a Severity Level IV
violation because: (1) a single EAL at the Notification of Unusual Event classification
level was affected, and (2) the violation was determined not to be a licensee failure to
meet or implement one emergency planning standard involving assessment or
notification.
Because this performance deficiency is of very low safety significance and has been
entered into the licensees corrective action system (CAR 200510162), this violation is
being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000483/2005005-10 (Change in Emergency Action Level 3E decreased the
effectiveness of the Emergency Plan).
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 Technical Specifications as criteria for determining compliance. The
inspector interviewed licensee personnel and reviewed:
Current 3-year rolling average collective exposure
Eight outage 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
Enclosure
-31-
Site-specific ALARA procedures
Eight work activities of highest exposure significance completed during the last
outage.
ALARA work activity evaluations, exposure estimates, and exposure mitigation
requirements
Intended versus actual work activity doses and the reasons for any
inconsistencies
Interfaces between operations, radiation protection, maintenance, maintenance
planning, scheduling, and engineering groups
Person-hour estimates provided by maintenance planning and other groups to
the radiation protection group with the actual work activity time requirements
Dose rate reduction activities in work planning
Assumptions and basis for the current annual collective exposure estimate, the
methodology for estimating work activity exposures, the intended dose outcome,
and the accuracy of dose rate and man-hour estimates
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
Exposures of individuals from selected work groups
Source-term control strategy
Declared pregnant workers during the current assessment period, monitoring
controls, and the exposure results
Self-assessments, audits, and special reports related to the ALARA program
since the last inspection
Corrective action documents related to the ALARA program and follow-up
activities such as initial problem identification, characterization, and tracking
Effectiveness of self-assessment activities with respect to identifying and
addressing repetitive deficiencies or significant individual deficiencies
Either because the conditions did not exist or an event had not occurred, no
opportunities were available to review the following items:
Enclosure
-32-
Records detailing the historical trends and current status of tracked plant source
terms and contingency plans for expected changes in the source term due to
changes in plant fuel performance issues or changes in plant primary chemistry
Radiation worker and radiation protection technician performance during work
activities in radiation areas, airborne radioactivity areas, or high radiation areas
The inspector completed 15 of the required 15 samples and 6 of the optional samples.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1
Routine Review of Identification and Resolution of Problems
The inspectors performed a daily screening of items entered into the licensee's
corrective action program. This assessment was accomplished by reviewing the daily
CAR Screening Report, Control Room Logs, and attending selected Corrective Action
Review Board and work control meetings. The inspectors: (1) verified that equipment,
human performance, and program issues were being identified by the licensee at an
appropriate threshold and that the issues were entered into the corrective action
program; (2) verified that corrective actions were commensurate with the significance of
the issue; and (3) identified conditions that might warrant additional follow-up through
other baseline inspection procedures.
.2
Selected Issue Follow-up Inspection
In addition to the routine review, the inspectors selected the two below listed issues for a
more in-depth review. The inspectors considered the following during the review of the
licensee's actions: (1) complete and accurate identification of the problem in a timely
manner; (2) evaluation and disposition of operability/reportability issues;
(3) consideration of extent of condition, generic implications, common cause, and
previous occurrences; (4) classification and prioritization of the resolution of the
problem; (5) identification of root and contributing causes of the problem;
(6) identification of corrective actions; and (7) completion of corrective actions in a timely
manner.
October 19, 2005, CAR 200508393, Tin whiskers: untimely corrective actions for
November 11, 2005, CAR 200509277, Unplanned pressurization and failure of
The inspectors completed two samples.
Enclosure
-33-
.3
Exposure Tracking, Higher than Planned Exposure Levels, and Radiation Worker
Practices
Section 2OS2 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 inspectors reviewed the corrective action
documents listed in the attachment against the licensees problem identification and
resolution program requirements.
.4
Semiannual Trend Review
a.
Inspection Scope
The inspectors completed a semiannual trend review of repetitive or closely related
issues that were documented in plant trend reports, problem lists, performance
indicators, system health reports, QA audit reports, corrective action documents, and
corrective maintenance documents to identify trends that might indicate the existence of
more safety significant issues. The inspectors' review consisted of the 6-month period
of July through December 2005. When warranted, some of the samples expanded
beyond those dates to fully assess the issue. The inspectors also reviewed corrective
action program items listed in the attachment. The inspectors compared and contrasted
their results with the results contained in the licensee's quarterly trend reports.
Corrective actions associated with a sample of the issues identified in the licensee's
trend report were reviewed for adequacy.
Documents reviewed by the inspectors are listed in the attachment.
b.
Findings and Observations
1.
Adverse Trend in Human Performance
The NRC identified an adverse human performance trend in December 2004 (Inspection
Report 05000483/2004005). The NRC subsequently identified a substantive
crosscutting issue in the area of human performance during the 2004 end-of-cycle
assessment. The substantive crosscutting issue was based on seven NRC findings
specifically related to personnel errors that occurred during 2004 and affected the
initiating events, mitigating systems, and barrier integrity cornerstones. AmerenUE
completed a stream analysis of the human performance events to identify commonality
and root causes in April 2005. In June 2005, the NRC and AmerenUE concluded that
the adverse trend continued during the first two calendar quarters in 2005 (NRC
Inspection Report 05000483/2005003). AmerenUE implemented the following
corrective actions in August 2005 to address the root causes of poor human
performance (CAR 200501425):
Established the Event Prevention Steering Committee
Enhanced the plant observation process by establishing metrics and
accountability
Enclosure
-34-
Identified and addressed deficiencies in the station root cause analysis
processes
Implemented a station focus of defense-in-depth error prevention tool/activities
The inspectors concluded that the adverse human performance trend continued during
the third and fourth quarters 2005. On November 28, 2005, the Callaway Event
Prevention Steering Committee also identified an adverse trend associated with station
noncompliance with written instructions (CAR 200509697). Examples used by the
licensee to identify the trend included:
CAR 200507092, September 20, 2005, Valve repositioned without the
appropriate procedure
CAR 200507699, October 2, 2005, 28 wire strand jack cables dropped from the
containment polar crane to the cavity deck
CAR 200508510, October 22, 2005, Failure to re-terminate 480 volt energized
CAR 200508753, October 28, 2005, Adverse trend of falling objects
CAR 200509404, November 15, 2005, Partial reactor trip due to failure to follow
lock and tag notes
2.
Adverse Trend in Corrective Action
The inspectors identified an adverse trend associated with ineffective corrective actions.
The inspectors considered the following examples of corrective actions that failed to
prevent recurrence of previously identified problems. The inspectors screened the
examples using Manual Chapter 0612, Appendix B, Issue Screening, and concluded
each example had only minor safety significance:
CAR 200509345, Unplanned main steam dump closure during reactor trip
breaker testing
CAR 200509474, Removal of the reactivity computer test leads out of sequence
caused a false pressurizer low level signal and charging system flow reduction
CAR 2005007860, Condensate storage tank wiper seal repeat cracking
CAR 200207808, Inadequate procedure resulted in the overpressurization of the
TDAFP suction piping and lube oil cooler
On December 1, 2005, AmerenUE also identified an adverse trend in corrective actions
resulting in a Red corrective action program system health indicator.
Enclosure
-35-
4OA5 Other Activities
.1
Temporary Instruction 2515/160, Pressurizer Penetration Nozzles and Steam Space
Piping Connections in U.S. Pressurized Water Reactors (NRC Bulletin 2004-01)
a. Inspection Scope
Industry OE has demonstrated that Alloy 82/182/600 materials exposed to primary
coolant water (or steam) at the normal operating conditions of pressurized water reactor
plants have cracked due to primary water stress corrosion cracking. The NRC issued
Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials Used in the Fabrication of
Pressurizer Penetrations and Steam Space Piping Connections at Pressurized-water
Reactors, was issued to alert licensee's to the susceptibility of Alloy 82/182/600
materials to cracking. The Callaway RCS has five pressurizer connections that were
applicable to the vulnerabilities described in NRC Bulletin 2004-01. The inspectors
compared the AmerenUE examinations of these five Alloy 82/182/600 pressurizer piping
connections with the licensees commitments documented in ULNRC-05031, Response
to NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials used in the
Fabrication of Pressurizer Penetrations and Steam Space Piping Connections at
Pressurized-Water Reactors, July 27, 2004. The inspectors performed this comparison
to verify that the examinations were consistent with the AmerenUE response to the
bulletin.
The inspectors reviewed records and examination procedures for visual examinations
(listed in the attachment) conducted during Refueling Outage 13 (Spring 2004) and
Refueling Outage 14 (Fall 2005). The inspectors performed this review to verify that the
bare metal examinations were adequate to detect the presence of boric acid crystals.
The inspectors used the guidance in Inspection Procedure 57050, Visual Testing
Examination, as acceptance criteria for this review. The inspectors reviewed volumetric
examinations conducted during 1992 and 1996. The inspectors used the guidance in
Inspection Procedure 57080, Ultrasonic Testing Examination, as acceptance criteria
for this review. The inspectors also reviewed the qualifications and certifications of the
personnel performing the examination and assessed the techniques used to detect
small boric acid deposits on the subject locations.
b.
Findings
No findings of significance were identified. The inspectors concluded that the
inspections conducted by AmerenUE were consistent with the licensees response to
NRC Bulletin 2004-01. The inspectors concluded:
Personnel performing the examination were qualified, knowledgeable, and
certified as visual examination Level 2 inspectors. Each inspector also received
additional training for identification of boric acid deposits.
The examinations were performed in accordance with station procedures and
were capable of identifying leakage in pressurizer penetration nozzle or steam
space piping components, as discussed in NRC Bulletin 2004-01.
Enclosure
-36-
The inspectors reviewed the photographic record of the examination and verified
that the physical condition of the penetration nozzles and steam space piping
components were good, without debris, insulation, dirt, or boron from other
sources during the visual examination.
The visual examination covered a 360° circumference of all the affected nozzles.
The examination was sufficient to identify and characterize small boron deposits,
as described in NRC Bulletin 2004-01.
AmerenUE did not identify any material deficiencies, cracks, or corrosion. No
indications of boric acid leaks from pressure-retaining components were identified
during the examinations and volumetric or surface examination techniques were not
used to augment the inspections.
.2
(Closed Unresolved Item 05000483/2005004-03) Potential Failure of the RHR
Containment Recirculation Sump Valves During Certain Design Bases Events
AmerenUE evaluated the containment recirculation sump valve operator torque needed
to open against the maximum calculated differential pressure that could be experienced
across the valve. AmerenUE determined that the valves were required to operate
against a 53 psid (Calculation RFR 05353, Revision F, October 31, 1989). AmerenUE
evaluated OE from the Catawba and McGuire plants (CAR 200504370) 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 30 minutes in the minimum flow configuration
and observed 189 psid across the sump valve. AmerenUE concluded that the maximum
differential pressure the valve actuator would be required to open against was 189 psid.
Engineering personnel concluded valve operability based on a linear extrapolation of the
actuator torque to the new conditions.
Subsequently, AmerenUE evaluated additional OE from the Wolf Creek plant on
September 21, 2005 (CAR 200507150). This OE alerted the industry to the potential of
additional differential pressure that could develop across the RHR sump valves while in
the minimum flow mode. AmerenUE reevaluated the RHR valves and determined that
the maximum differential pressure the valves had to open against could be 468 psid.
AmerenUE verified past sump valve operability using actual valve factors and a realistic
lock-rotor valve operator torque. AmerenUE modified the operators to provide higher
opening torque to ensure future RHR valve operability. The failure of Ameren to ensure
suitability of the RHR containment suction valves function to open under all safety-
related design bases conditions was a licensee-identified violation of 10 CFR Part 50,
Appendix B, Criteria III, Design Control. The enforcement aspects of this violation are
discussed in Section 4OA7 of this report.
Enclosure
-37-
.3
(Closed Apparent Violation 05000483/2005004-01) Failure to Maintain Cold
Overpressure Mitigation Measures as Required by TSs
a. Inspection Scope
A senior reactor analyst performed a Phase 3 significance determination of apparent
violation 05000483/2005004-01. The inspectors evaluated this finding using the
guidance in Manual Chapter 0612, Power Reactor Inspection Reports, dated
September 30, 2005, for determining whether a violation is licensee-identified because
this finding had not been closed prior to the revised guidance being issued. This
apparent violation is closed as a licensee-identified violation of very low safety
significance. The violation is documented in Section 4OA7 of this report.
b.
Findings
Introduction: The senior reactor analysts completed the significance determination of
the apparent violation documented in NRC Inspection Report 05000483/2005004. The
apparent violation involved the failure of AmerenUE operations personnel to ensure no
more than one centrifugal charging pump was capable of injecting into the reactor
vessel while in Mode 5, as required by TS 3.4.12.
Analysis: The performance deficiency associated with this finding involved the
licensees failure to establish and follow adequate procedures. This finding is greater
than minor because it would have become more significant, if left uncorrected, in that
inadvertent starting of the charging pump could have challenged the piping integrity of
the RCS system. The inspectors used Appendix G, Shutdown Operations Significance
Determination Process, of Manual Chapter 0609, Significance Determination Process,
to determine the significance of this finding. Unplanned entry into cold
overpressurization conditions represented additional risk incurred above the planned
outage risk. The additional risk associated with the ability of the centrifugal charging
pump to inject into the RCS constituted additional risk above the planned outage risk.
Phase 1 screening of this finding was performed using Appendix G and the
Attachment 1 checklists. Management review determined that significance
determination process Phase 3 analysis was needed for this finding.
The senior reactor analysts review of the Callaway cold overpressure mitigation
(COMS) precursor involved having both centrifugal charging pumps capable of RCS
injection. This condition lasted approximately 20 minutes.
The following conditions existed at the time of the event:
Pressurizer level was at 5 percent
There was a high pressurizer level alarm at 90 percent
There was an alarm at 5 percent level above program increase
RCS level was not being changed at the time of the event
Enclosure
-38-
No testing was being performed on or in systems connected to the RCS that
could perturb RCS level
No work was being performed on RCS level indication other than adding an
additional, alternate reactor vessel level indication with separate tap locations
The safety injection pumps were in pull-to-lock
The accumulators were isolated and vented
Each RHR train had a suction relief valve with a lift setpoint of 450 pounds per
square inch gauge (psig) (986 gpm discharge capacity)
Both trains of RHR were aligned to the RCS with one train providing decay heat
removal. Therefore, both RHR suction relief valves were available to relieve a
postulated cold overpressure challenge
Two power-operated relief valves were available for COMS, the low power-
operated relief valve setpoint was at 500 psig, the high power-operated relief
valve setpoint was at 525 psig
The pressurizer was vented to atmosphere (via a 3/4-inch manual vent valve)
To assess the risk of the event required an estimate of the likelihood that the operators
would have initiated RCS injection, resulting in a solid RCS. Based on the above
information, it appears that no plant operations were being performed at the time that
had the potential to trigger the operators to initiate RCS injection. Additionally, the alarm
indicating 5 percent above program level provides additional assurance that the
likelihood of overfilling the RCS during the 20-minute time period was small.
If a postulated RCS pressure challenge were to occur, the dominant core damage
scenario involves both RHR suction relief valves failing to reseat after an RCS pressure
challenge. In this design, the RHR suction relief valves have a lower relief setpoint than
the pressure-operated relief valves. Should one RHR relief valve fail to reseat, the
operators could isolate the valve and use the alternate train of RHR for decay heat
removal. If both relief valves were to fail to reseat, the operators would be directed to
increase charging and isolate the leak. In this plant condition, steam generator cooling
is not anticipated to match decay heat; therefore, the RCS may re-pressurize until steam
generator cooling can remove decay heat. For this situation, both pressure-operated
relief valves would be available should RCS pressure increase to the COMS setpoint.
In summary, combining the small likelihood of having an RCS pressure challenge during
the 20-minute period, the likelihood of having both RHR relief valves stick open after a
challenge, and the failure of both pressure-operated relief valves to relieve pressure, the
core damage frequency delta for this finding is estimated to be less than 1E-6.
Therefore, this finding can be characterized in the significance determination process as
Enclosure
-39-
Green. It is important to note that the licensee's robust COMS mitigation capability (the
availability of both RHR suction relief valves and the pressure-operated relief valves)
was significant in reducing the risk of this finding.
The review of the licensee's analysis only considered the likelihood of the COMS system
failing to provide RCS pressure relief following a demand. The licensee did not consider
that an RCS pressure demand may result in the RHR suction relief valve lifting and not
reseating. This scenario results in a loss of coolant accident in the RHR system as
described above.
This finding affected the barrier integrity cornerstone and the configuration control,
procedure quality, and human performance attributes of maintaining functionality of the
RCS. The senior reactor analyst determined that this finding is only of very low
significance.
Enforcement: The enforcement aspects of this finding are discussed in Section 4OA7 of
this report.
4OA6 Management Meetings
Exit Meeting Summary
On December 14, 2005, the health physics inspector presented the ALARA inspection
results to Mr. A. Heflin, Vice President, and other members of his staff who
acknowledged the findings.
On January 6, 2006, 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 emergency preparedness inspector conducted a telephonic exit interview on
January 12, 2006, to present the inspection results to Mr. M. Reidmeyer, Supervisor,
Regional Regulatory Affairs, and other members of his staff who acknowledged the
findings.
The operations branch inspectors conducted an exit meeting on June 9, 2005, regarding
the on-site portion of the inspection with Mr. R. Roselius and other members of the
licensee's staff. On December 15, 2005, the inspectors discussed biennial written
requalification examination issues with the licensee. After NRC management review of
the biennial written requalification examination observations, the inspectors again
discussed the unresolved item identified during the review of the written biennial
requalification exams with the licensee during a teleconference on January 23, 2006.
The inspectors verified that no proprietary information was provided during the
inspection.
Enclosure
-40-
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the
licensee and are violations of NRC requirements which meet the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.
Title 10 CFR Part 50, Appendix B, Criteria III, Design Control, required that measures
be established for the selection and suitability of application of equipment essential to
the safety-related functions of the SSCs. Contrary to this, on October 31, 1989, and
October 5, 2005, the selection and suitability of application for the RHR containment
sump valve operators was inadequate to ensure all safety-related functions. AmerenUE
had established an insufficient maximum differential pressure design that the sump
valves would have to open against during certain design bases events. This was
identified in the licensees corrective action program as CAR 200504370. This finding is
of very low safety significance because it does not represent a design or qualification
deficiency confirmed not to result in loss of operability per Part 9900, Technical
Guidance, Operability Determination Process for Operability and Functional
Assessment.
Title 10 CFR 55.49 requires examination integrity to be maintained. The regulation
further defines an examination compromise as any activity, regardless of intent, that
affected or could have affected the equitable and consistent administration of an
examination.
During a review of CARs, the inspectors noted that two events occurred that had the
potential to effect the integrity of the requalification examinations. The first event
occurred on May 26, 2005, and involved leaving data on the simulator's "white board"
from the previous scenario training crew. The data displayed provided information that
could be used by the oncoming training crew to assist them with the scenario (since the
same scenario was to be run). This compromise was identified by the licensee's
oncoming training crew. As a result, the oncoming crew was given a different scenario.
The second event occurred on June 8, 2005, and involved the accidental observation of
some pages out of a written examination by a candidate assigned to the training staff.
This candidate was scheduled to take the same specific examination. When the
licensee identified this compromise, the candidate was rescheduled to take a different
written examination.
These findings are greater than minor because a compromise of the integrity of the
annual requalification examinations could lead to operators (who would normally have
failed the examination) with deficient knowledge and skills to remain on shift. Allowing
operators with deficient knowledge and skills to remain on shift increases the likelihood
that a human performance error could initiate a reactor safety event or inhibit the
appropriate mitigating response to such an event. Contrary to the above, the licensee
failed to adequately assure that examination security was maintained during the
administration of examinations. The finding is of very low safety significance because
Enclosure
-41-
the potential for examination compromise was extremely low. These findings have been
entered into the corrective action program as CARs 200503988 and 200503985,
respectively.
TS 5.4.1.a, Procedures, and Regulatory Guide 1.33, Appendix A, required procedures
for shutdown to be implemented. Procedure OSP-BG-00002, Verify One Centrifugal
Charging Pump Incapable of Injection into RCS, required the licensee to ensure only
one centrifugal charging pump was capable of injecting to the RCS during Mode 5
operations with limited RCS vent path. Contrary to the above, on September 20, 2005,
the licensee failed to ensure only one centrifugal charging pump was capable of
injecting to the RCS. This finding is greater than minor because it would have become
more significant, if left uncorrected, in that inadvertent starting of the charging pump
could have challenged the piping integrity of the RCS system. This finding was
determined to be very low significance after completion of a Phase 3 SDP by the senior
reactor analyst as documented in Section 40A5 of this report. This finding was identified
in the licensees corrective action program as CAR 200507092.
A-1
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
W. Arbour, Senior Operations Training Supervisor
S. Aufdemberge, Operating Supervisor
K. Bruckerhoff, Supervisor, Emergency Preparedness
F. Diya, Manager, Engineering Services
R. Farnam, General Supervisor, Radiation Protection
S. Ganz, Operating Supervisor
J. Geyer, Health Physicist, Radiation Protection
K. Gilliam, ALARA Coordinator, Radiation Protection
S. Halverson, General Supervisor, Simulator
A. Heflin, Site Vice President
T. Herrmann, Vice President, Engineering
J. Hiller, Regulatory Affairs, Engineer
G. Hurla, Supervisor, Radiation Protection
M. Jennings, Operating Supervisor
L. Kanuckel, Manager, Quality Assurance
S. Kochert, Operating Supervisor
V. Miller, ALARA Specialist, Radiation Protection
R. Moody, Operating Supervisor
T. Moser, Manager, Plant Engineering
C. Naslund, Senior Vice President and Chief Nuclear Officer
R. Nelson, Shift Supervisor
D. Neterer, Manager, Operations
M. Reidmeyer, Supervisor, Regional Regulatory Affairs
R. Roselius, Superintendent, Training
K. Young, Manager, Regulatory Affairs
LIST OF ITEMS OPENED AND CLOSED
Opened
Indeterminate Containment Cooler Operability and Heat
Removal Capability (Section 1R07)05000483/2005005-04
Adequacy of the Biennial Requalification Written
Examination (Section 1R11)05000483/2005005-06
Adequacy of Plant-Referenced Simulator to Conform with
Simulator Requirements for Reactivity and Control
Manipulation Credits (Section 1R11)
A-2
Attachment
Opened and Closed
Minimum gap size exceeded for containment recirculation
sump (Section 1R04)
Seven examples of inadequately performed continuous fire
watches (Section 1R05)05000483/2005005-05
Failure to Conduct Simulator Testing in Accordance with
ANSI/ANS 3.5-1998 (Section 1R11)05000483/2005005-07
Failure to Follow Procedures Resulted in Violation of RCS
Cooldown and Heatup Rate Limits (Section 1R14)05000483/2005005-08
Use of a Nonqualified Calculation in a Safety-Related
Modification (Section 1R17)05000483/2005005-09
Less Than Adequate Spent Fuel Pool Water Inventory
Risk Controls (Section 1R20)05000483/2005005-10
Change in EAL 3E decreased the effectiveness of the
Emergency Plan (Section 1EP4)
Closed
Potential Failure of the RHR Containment Suction Valves
During Certain Design Bases Events (Section 4OA5)05000483/2005004-01
Failure to Maintain Cold Overpressure Mitigation Measures
as Required by TSs (Section 4OA5)
DOCUMENTS REVIEWED
Section 1R04: Equipment Alignment
Drawings
E-23KJ01A, Revision 14, Diesel General KKJ01 Engine Control (Start / Stop) Circuit
M22-BG03, Chemical and Volume Control System
M22-BG05, Chemical and Volume Control System
M22-EJ01, Residual Heat Removal System
Miscellaneous
Callaway Action Request 200509189
Procedure OSP-EJ-00003, Containment Recirculation Sump Inspection, Revision 5
A-3
Attachment
FSAR Table 6.2.2-1, Comparison of the Recirculation Sump Design with each of the Positions
of Regulatory 1.82
NRC Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation
at Pressurized-Water Reactors
Work package W229952, Recirculation Sump Inspection
ULNRC-04966, Callaway Plant, Union Electric Co. Supplement to Response to NRC
Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at
Pressurized Water Reactors
Section 1R05: Fire Protection
Procedures
APA-ZZ-00743, Fire Team Organization and Duties, Revision 18
EIP-ZZ-00226, Fire Response Procedure for Callaway Plant, Revision 11
SDP-KC-00001, Requirements for and Duties of Compensatory Fire Watches, Revision 5
Requests for Resolution
RFR 15704, Electrical Safety Equipment Lockers, Revisions A and B
RFR 18572, Allowed Storage of D/G Tool Boxes and Barring Device, Revision A
RFR 3487, Breaker Test Area in NB01 Switchgear Room 3301, Revision B
Miscellaneous
Information Notice 97-48, Inadequate or Inappropriate Interim Fire Protection Compensatory
Measures
Section 1R07: Heat Sink Performance
Callaway Action Requests
CAR 200503773, Containment cooler heat removal surveillance requirements
CAR 200502534, Incorrect component cooling water heat exchanger indication
Drawings
M-22EF02, Essential Service Water System
M-22EF08, Essential Service Water Containment Air Coolers
Procedures
ESP-EF-002A, Essential Service Water Train A Flow Verification, Revision 0
OSP-EF-P001A, ESW Train A Inservice Test, Revision 43
A-4
Attachment
Miscellaneous
SGN01A ETP-ZZ-03001, Heat Exchange Inspection Report, Revision 5, completed on
September 23, 2005
Surveillance 05515092, Essential Service Water, performed on October 12, 2005
Work Package W 236012/920, Containment Cooler Unit A PMT
Section 1R11: Licensed Operator Requalification
Procedures
TDP-IS-00002, Simulator Configuration Management, Revision 4
TDP-IS-00001, Simulator Operation and Maintenance, Revision 3
SRO-RER02C113J(TC), Emergency Event Classification, Revision 20040710
URO-SEG02C21J, Shift Non-essential CCW Supply Loops, Revision 20050604
URO-AEO05045J, Locally Close Valves for a CIS [Containment Isolation Signal]-B,
Revision 20050508
URO-SGN02C27J, Secure D Containment Cooler Fan, Revision 20050421
URO-AEO15016J, Local Manual Start of NE02, Revision 20050422
URO-SBB04C67J(A), Pressurizer Level Channel Failure, Revision 20050421
SRO-RER02C143J(TC), Emergency Event Classification, Revision 20050413
URO-SAB04C61J(A), Place Steam Dumps in Steam Pressure Mode, Revision 20050413
URO-SEF02C03J, Manually Operate an ESW Train, Revision 20050413
EOP-SBG06014J, Shift and Vent CVCS Seal Water Injection Filters, Revision 20050314
URO-AEO05001j(A), Locally Start (NE01) Emergency Diesel, Revision 20050413
URO-SBG02C04J, Swap From B CCP to NCP, Revision 20050413
SRO-RER02C118(TC), Emergency Event Classification, Revision 20050323
URO-SGN04C71J(A), Start A Containment Cooler Fan, Revision 20050820
EOS-SNN03011J, Shift an Instrument Bus to Backup Power Supply, Revision 20050323
A-5
Attachment
URO-Paralleling Diesel Generator A to XNB01, Revision 20050625
URO-SSF01C05J, Perform Control Rod Partial Movement Test, Revision 20050215
URO-AEO05PA023J, Locally Close Valves for a CIS-A, Revision 20050323
URO-SSP03C15J, Radiation Monitors Source Check, Revision 20050328
URO-AEO15029J, Locally Isolate a MSIV, Revision 20050314
URO-SBG02C01J, Placing Excess Letdown in Service, Revision 20050314
SRO-RER02C45J, Emergency Event Classification, Revision 20050414
URO-AEO01C151J(A), Emergency Boration Per /ES-0.1/Addendum 4, Revision 20050502
EOS-SNK01051J, Place NK22 in Service to Bus NK02, Revision 20050314
Scenarios
DS-07, Small Break LOCA With Failure of CPIS [Containment Purge Isolation System] and
CCP [Component Cooling Pump]/Loss of NB01, Revision 20050520
DS-32, Faulted-Ruptured S/G, Revision 20050520
DS-14, Separate Faulted and Ruptured S/Gs, Revision 20050310
DS-24, Loss of Letdown, ATWS with Stuck Open Pressurizer Safety Valve, Revision 20050311
DS-15, Load Increase with Multiple Rod Drop/Pressurizer Steam Space Leak,
Revision 20050507
DS-40, Faulted/Ruptured S/G, Revision 20050507
DS-04, Loss of Heat Sink without Bleed and Feed Required, Revision 20050514
DS-05, Faulted/Ruptured S/G, Revision 20050514
DS-01, ATWS, Revision 20050308
DS-26, Large LOCA and Transfer to Cold Leg Recirculation, Revision 20050310
DS-08, Feedline Break Inside Containment with CCP and SLIS Failures, Revision 20050414
DS-37, Station Blackout due to Seismic Conditions, Revision 20050329
DS-18, SGTR Without Pressurizer Pressure Control, Revision 20050422
A-6
Attachment
DS-19, Turbine Trip Failure with Loss of Heat Sink, Revision 20050422
Written Examinations
T61.0810 8, LOCT Cycle 05-4 Biennial Exam, SRO Week 1
T61.0810 8, LOCT Cycle 05-4 Biennial Exam, URO Week 2
Miscellaneous
2003-2005 Continuing Sample Plan
Job-Duty-Task by Job for URO [Unit Reactor Operator] dated 3/17/05
Job-Duty-Task by Job for SRO dated 4/14/05
Written Summary of Simulator Testing Topic Public Meeting with Industry Focus Group (FG) on
Operator Licensing Issues (DRAFT)
Response to April 7, 2004 Public Meeting Minutes Attachment 6
Callaway Plant Simulator White Paper showing how all parameters are demonstrated, June 8,
2005
Simulator Annual Performance Test Book
Simulator "Differences" List, May 16, 2005
Section 1R17: Permanent Plant Modifications
Calculations
330-001-DC1, Motor terminal voltage and nominal torque output, Revision 0
EJ-42, MOV sizing for EJHV8811A and EJHV8811B, Revision 0
Westinghouse Calculation SCP-05-69, Valve Factors for Valve Location 8811A and 8811B,
October 28, 2005
Callaway Action Requests
200507150
200509849
200505194
Miscellaneous
Predictive Performance Report, E170.0197, CA 1527, May 10, 1990
A-7
Attachment
Modification MP 05-3051
Section 1R19: Postmaintenance Testing
Procedures
OSP-SF-00005, Estimated Critical Rod Position Calculation ST-13002, Revision 16
ETP-ZZ-ST010, Low Power Physics Test Program with Dynamic Rod Worth Measurement,
Revision 8
OSP-BG-0001A, Boron Injection Flowpaths, Revision 14
APA-ZZ-00500, Corrective Action Program, Revision 38
OSP-AL-P0002, Turbine-Driven Auxiliary Feedwater Pump Inservice Test, Revision 49
Miscellaneous
PM0826213, Overhaul of NN Inverter, PMB Charging-1-5.2-4, Revision 0
Section 1R20: Refueling and Outage Activities
Procedures
APA-ZZ-00150, Outage Preparation and Execution, Revision 12
EDP-ZZ-1129, Callaway Plant Risk Assessment, Revision 8
OSP-SF-00003, Pre-Core Alteration Verifications, Revision 12
OSP-SF-00003, Pre-Core Alterations Verifications, Revision 15
OSP-ZZ-00001, Control Room Shift and Daily Log Readings and Channel Checks, Revision 39
OTG-ZZ-00001, Plant Heatup Cold Shutdown to Hot Standby, Revision 45
OTG-ZZ-00006, Plant Cooldown Hot Standby to Cold Shutdown, Revision 6
OTO-KE-00001, Fuel Handling Accident, Revision 7
QCP-ZZ-05048, Boric Acid Walkdown for RCS Pressure Boundary, Revision 2
Miscellaneous
Curve Book, Figure 8-6, RCS Pressure-Temperature Limitations
Nuclear Utility Management and Resource Council 91-06, Guidelines for Industry Actions to
Assess Shutdown Management
Quality Assurance Surveillance Reports
SP05-028, December 12, 2005, Assess lifting, removal and placement of the reactor vessel
head and upper internals
A-8
Attachment
SP05-047, December 9, 2005, Reduced inventory control, risk assessment, outage technical
specifications
Callaway Action Requests
200002070
200202540
200302806
200307232
200307247
200307844
200402256
200500720
200500756
200501092
200501407
200501837
200501990
200502420
200502438
200502548
200503439
200503622
200503773
200504591
200504950
200505062
200505368
200505716
200506244
200507150
200507278
200508169
200507593
200507693
Section 1R22: Surveillance Testing
Procedures
OSP-EM-V0003, ECCS Check Valve Inservice Test IPTE, Revision 21
OSP-BB-00006, Reactor Coolant Circulation, Revision 7
OSP-BG-0001A Boron Injection Flowpaths modes 4 through 8
Audits and Self-Assessments
Quality Assurance Surveillance Report SP05-027, November 6, 2005, Assess effectiveness of
fuel movement, compliance to TSs and procedures applicable to fuel movement
Quality Assurance Surveillance Report SP05-037, November 18, 2005, Assess implementation
of the steam generator replacement project
Section 71152: Identification and Resolution of Problems
Procedures
APA-ZZ-00500, Corrective Action Program, Revision 38
Callaway Action Requests
200507092
200507699
200508510
200508753
200509404
A-9
Attachment
Miscellaneous
Callaway Plant Quarterly Performance Analysis Report Third Quarter
Event Review Team Meeting Summaries
AUCA 05-040, October 2, 2005, Strand wires were dropped from the TLD on the polar crane to
the cavity deck
AUCA 05-047, October 15, 2005, Corrosion discovered on the new B low pressure turbine rotor
AUCA 05-049, October 17, 2005, Employee falls in containment while wearing fall protection
AUCA 05-050, October 19, 2005, Pit at VBS checkpoint lowered prematurely
AUCA 05-057, October 29, 2005, Leaking head gaskets during KKJ01B maintenance run
Surveillance Reports
SP05-034, September 24, 2005, Postmodification test planning for CMP 03-1014 - EP8818A-D
valve replacements
SP05-045, September 29, 2005, Bottom mounted instrumentation inspection and cleaning
SP05-026, September 30, 2005, Assess various areas during plant shutdown
SP05-061, October 28, 2005, Refuel 14 worker practices
SP05-070, November 5, 2005, QA walkdowns to assure appropriate combustible loadings and
housekeeping, and operable fire doors and halon systems
SP05-074, November 15, 2005, Assess interim compensatory actions in response to NRC
Bulletin 20003-1 and Generic Letter 2004-2
SP05-044, November 23, 2005, Refuel 14 work activities on the TDAFP
SP05-056, November 29, 2005, Review of the tin whisker inspections
SP05-068, November 30, 2005, Assessment of Operating License Amendment 1248`
SP05-029, December 6, 2005, Assess effectiveness of control room personnel from Mode 3
ascending to Mode 1
SP05-058, December 14, 2005, QA assessment of Refueling Outage 14 mode change
restraints
SP05-063, December 15, 2005, ESW strainer replacement activities
A-10
Attachment
SP05-071, December 8, 2005, Review control logs and verify CARs were written when
appropriate
SP05-078, November 30, 2005, Main feedwater regulation valve and bypass regulation valve
testing in Refuel 14
Callaway Plant Quarterly Performance Analysis Report First Quarter
Callaway Plant Quarterly Performance Analysis Report Second Quarter
Quality Assurance Audits
AP05-010, October 5, 2005, Problem resolution, adverse trends, OQAM audit
requirements/other commitments, review of self-assessments, organization, special nuclear
material program, special nuclear material inventory, source control, and software management
Section 4OA5: Other Activities
Surveillances
S724682, Task 150, Inspection of pressurizer surge nozzle welds for boron
S724682, Leakage examination of the RCS, September 25, 2005
S714761, Leakage examination of the RCS, April 29, 2004
Procedures
QCP-ZZ-05048, Boric Acid Walkdown for RCS Pressure Boundary, Revision 2
QCP-ZZ-05048, Boric Acid Walkdown for RCS Pressure Boundary, Revision 1
Miscellaneous
Letter to the NRC from AmerenUE, ULNRC-05031, July 27, 2004, Response to NRC Bulletin 2004-01, Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer
Penetrations and Steam Space Piping Connections at Pressurized Water Reactors
UT Data Sheet 1021-02, Examination BB 2TBB03-1-w, April 18, 1992
UT Data Sheet 1021-01, Examination BB 2TBB03-2-w, April 17, 1992
UT Data Sheet 6276-95-01, Examination BB 2TBB03-3-A-w, October 30, 1996
UT Data Sheet 6276-95-02, Examination BB 2TBB03-3-B-w, October 30, 1996
UT Data Sheet 6276-001, Examination BB 2TBB03-3-C-w, October 27, 1996
UT Data Sheet 6276-002, Examination BB 2TBB03-4-w, October 27, 1996
CAR 200507515, Boric acid walkdown for Refuel 14
A-11
Attachment
LIST OF ACRONYMS
as low as is reasonably achievable
Callaway Action Request
cold overpressure mitigation
emergency action level
essential service water
finding
Final Safety Analysis Report
noncited violation
operational experience
psid
pounds per square inch differential
psig
pounds per square inch gauge
postmaintenance test
structures, systems, and components
turbine-driven auxiliary feedwater pump
TMs
TSs
Technical Specifications
unresolved item