ML030280726
| ML030280726 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/27/2003 |
| From: | Collins L NRC/RGN-III/DRP/RPB6 |
| To: | Bakken A American Electric Power Co |
| References | |
| IR-02-009 | |
| Download: ML030280726 (78) | |
See also: IR 05000315/2002009
Text
January 27, 2003
Mr. A. C. Bakken III
Senior Vice President
Nuclear Generation Group
American Electric Power Company
500 Circle Drive
Buchanan, MI 49107
SUBJECT:
D.C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2
NRC INSPECTION REPORT 50-315/02-09(DRP); 50-316/02-09(DRP)
Dear Mr. Bakken:
On December 28, 2002, the NRC completed an inspection at your D. C. Cook Nuclear Power
Plant, Units 1 and 2. The enclosed report documents the inspection findings which were
discussed on January 3, 2003, with Mr. J. Pollock and other members of your staff.
This inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, six findings of very low safety significance (Green) were
identified which involved violations of NRC requirements. However, because of their very low
safety significance and because they have been entered into your corrective action program,
the NRC is treating these issues as Non-Cited Violations, in accordance with Section VI.A.1 of
the NRC Enforcement Policy. If you contest the Non-Cited Violations, you should provide a
response with the basis for your denial, within 30 days of the date of this inspection report, to
the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington,
D.C. 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of
Enforcement, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the
NRC Resident Inspector at the D. C. Cook facility.
Since the terrorist attacks on September 11, 2001, the NRC has issued two Orders (dated
February 25, 2002, and January 7, 2003) and several threat advisories to licensees of
commercial power reactors to strengthen licensee capabilities, improve security force
readiness, and enhance access authorization. The NRC also issued Temporary Instruction
2515/148 on August 28, 2002, that provided guidance to inspectors to audit and inspect
licensee implementation of the interim compensatory measures (ICMs) required by the
February 25th Order. Phase 1 of TI 2515/148 was completed at all commercial nuclear power
plants during calendar year (CY) 02, and the remaining inspections are scheduled for
completion in CY 03. Additionally, table-top security drills were conducted at several licensees
to evaluate the impact of expanded adversary characteristics and the ICMs on licensee
protection and mitigative strategies. Information gained and discrepancies identified during the
A. Bakken
-2-
audits and drills were reviewed and dispositioned by the Office of Nuclear Security and Incident
Response. For CY 03, the NRC will continue to monitor overall safeguards and security
controls, conduct inspections, and resume force-on-force exercises at selected power plants.
Should threat conditions change, the NRC may issue additional Orders, advisories, and
temporary instructions to ensure adequate safety is being maintained at all commercial power
reactors.
In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter
and its enclosures will be available electronically for public inspection in the NRC Public
Document Room or from the Publicly Available Records (PARS) component of NRC's
document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Laura Collins, Acting Chief
Branch 6
Division of Reactor Projects
Docket Nos. 50-315; 50-316
Enclosure:
Inspection Report 50-315/02-09(DRP);
50-316/02-09(DRP)
cc w/encl:
J. Pollock, Site Vice President
M. Finissi, Plant Manager
R. Whale, Michigan Public Service Commission
Michigan Department of Environmental Quality
Emergency Management Division
MI Department of State Police
D. Lochbaum, Union of Concerned Scientists
DOCUMENT NAME: C:\\ORPCheckout\\FileNET\\ML030280726.wpd
To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy
OFFICE
RIII
RIII
RIII
RIII
NAME
LCollins/trn
DATE
01/27/03
OFFICIAL RECORD COPY
A. Bakken
-3-
ADAMS Distribution:
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C. Ariano (hard copy)
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket Nos:
50-315; 50-316
License Nos:
Report No:
50-315/02-09(DRP); 50-316/02-09(DRP)
Licensee:
Indiana Michigan Power Company
Facility:
D. C. Cook Nuclear Power Plant, Units 1 and 2
Location:
1 Cook Place
Bridgman, MI 49106
Dates:
October 1, 2002 through December 28, 2002
Inspectors:
B. Kemker, Senior Resident Inspector
I. Netzel, Resident Inspector
R. Azua, Project Engineer, Region IV
M. Bielby, Operations Engineer
J. Ellegood, Resident Inspector, Perry
R. Gibbs, Senior Reactor Analyst, NRR
R. Jickling, Emergency Preparedness Inspector
R. Krsek, Resident Inspector, Palisades
J. Maynen, Physical Security Inspector
W. Poertner, Operations Engineer, NRR
R. Powell, Senior Resident Inspector, Perry
P. Prescott, Operations Engineer, NRR
R. Schmidt, Radiation Specialist
H. Walker, Senior Reactor Engineer
R. Winter, Reactor Engineer
S. Wong, Senior Reactor Analyst, NRR
D. Wrona, Operations Engineer, NRR
Approved by:
L. Collins, Acting Chief
Branch 6
Division of Reactor Projects
1
SUMMARY OF FINDINGS
IR 05000315-02-09(DRP), IR 05000316-02-09(DRP), on 10/01/2002-12/28/2002, Indiana
Michigan Power Company, D. C. Cook Nuclear Power Plant, Units 1 and 2. Maintenance
Effectiveness, Identification and Resolution of Problems, Event Follow-up.
This report covers a 13-week period of inspection by resident, regional, and headquarters
based inspectors. The inspectors identified six Green findings. The significance of most
findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual
Chapter 0609, "Significance Determination Process," (SDP). The NRCs program for
overseeing the safe operation of commercial nuclear power reactors is described in
NUREG 1649, "Reactor Oversight Process," Revision 3, dated July 2000.
A.
Inspector Identified Findings
Cornerstone: Initiating Events
Green. The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to assure that
prompt corrective actions were taken to address age-related failures of reactor
control instrumentation power supplies to prevent repetition of power supply
failures, a significant condition adverse to quality. This issue was self-revealed
on May 12, 2002, when an automatic reactor trip of Unit 2 occurred due to the
failure of redundant 24-volt direct current power supplies in reactor control
instrumentation cabinet 2-PS-CGC-16. The failure of both power supplies
caused the number 21 steam generator feedwater regulating valve to close.
Unit 2 subsequently tripped on low steam generator water level coincident with
low feedwater flow.
The inspectors assessed this finding using the Significance Determination
Process (SDP). The inspectors concluded that this issue, if left uncorrected,
would become a more significant safety concern with the likelihood of continued
failures of reactor control instrumentation power supplies and was therefore
more than a minor concern. The inspectors also concluded that this finding was
associated with the initiating events cornerstone and adversely affected the
cornerstone objective. Specifically, the failure of redundant power supplies in
reactor control instrumentation cabinets would upset plant stability (cause a
reactor trip) and challenge the function of critical safety equipment. The
inspectors performed a Phase 1 SDP review of this finding using the guidance
provided in NRC Inspection Manual Chapter (IMC) 0609, Appendix A,
"Significance Determination of Reactor Inspection Findings for At-Power
Situations." Because this finding contributes to both the likelihood of a reactor
trip and the likelihood that mitigation equipment or functions will not be available,
the inspectors determined that this finding required a Phase 2 SDP analysis.
After a review of additional information, the inspectors determined that a Phase 3
analysis was required. The Phase 3 SDP analysis, performed with the
assistance of the NRC probabilistic risk analysis staff, determined that the
resultant Core Damage Frequency and Large Early Release Frequency
2
associated with this finding were less than 1E-6 per year and 1E-7 per year,
respectively. Based on these results, this issue was determined to be of very
low safety significance. (Section 1R12)
Green. The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to take
corrective action to preclude the repetition of reactor control instrumentation
24-volt direct current power supply failures. Specifically, the licensee failed to
perform weekly verification of control group power supplies to ensure that the
"power available" status lights were lit. This corrective action was identified by
the licensee in response to the Unit 2 reactor trip on May 12, 2002, which was
caused by the failure of redundant power supplies in reactor control
instrumentation cabinet 2-PS-CGC-16. The licensee subsequently performed
this check on November 22, 2002, and discovered a failed 24-volt direct current
power supply in Unit 1 cabinet 1-PS-CGC-16.
The inspectors assessed this finding using the Significance Determination
Process (SDP). The inspectors concluded that this issue could be reasonably
viewed as a precursor to a significant event (i.e., potentially result in a reactor trip
similar to the Unit 2 trip on May 12, 2002), and was therefore more than a minor
concern. The inspectors also concluded that this finding was associated with the
initiating events cornerstone and adversely affected the cornerstone objective.
Specifically, the failure of redundant power supplies in reactor control
instrumentation cabinets would upset plant stability (cause a reactor trip) and
challenge the function of critical safety equipment. The inspectors performed a
Phase 1 SDP review of this finding using the guidance provided in NRC
Inspection Manual Chapter 0609, Appendix A, "Significance Determination of
Reactor Inspection Findings for At-Power Situations." Because this finding
contributes to both the likelihood of a reactor trip and the likelihood that
mitigation equipment or functions will not be available, the inspectors determined
that this finding required a Phase 2 SDP analysis. After a review of additional
information, the inspectors determined that a Phase 3 analysis was required.
The Phase 3 SDP analysis, performed with the assistance of the NRC
probabilistic risk analysis staff, determined that the resultant Core Damage
Frequency and Large Early Release Frequency associated with this finding were
less than 1E-6 per year and 1E-7 per year, respectively. Based on these results,
this issue was determined to be of very low safety significance. (Section 1R12)
Green. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V,
"Instructions, Procedures, and Drawings," was self-revealed. The licensee failed
to provide an appropriate procedure for testing the Unit 1 pressurizer power
operated relief valves (PORVs), causing an uncontrolled release of reactor
coolant system inventory to the pressurizer relief tank. This issue was
self-revealed on June 5, 2002, when pressurizer PORV 1-NRV-153 inadvertently
opened while testing actuation logic circuitry for pressurizer PORV 1-NRV-151.
The surveillance test procedure failed to provide adequate control of 1-NRV-151
and 1-NRV-153, which have a common automatic opening signal. The release
rate exceeded the 25 gallons-per-minute limit established for declaring an
Unusual Event in accordance with the licensees Emergency Plan.
3
The inspectors assessed this finding using the Significance Determination
Process (SDP). The inspectors concluded that this issue could be reasonably
viewed as a precursor to a significant event and was therefore more than a
minor concern. The inspectors also concluded that this finding was associated
with the initiating events cornerstone and adversely affected the cornerstone
objective. Specifically, the uncontrolled release of reactor coolant system
inventory upset plant stability and challenged the inventory control safety
function. Because Unit 1 was in a shutdown mode during this period, the
inspectors performed a Phase 1 SDP review of this issue using the guidance
provided in NRC Inspection Manual Chapter 0609, Appendix G, "Shutdown
Operations Significance Determination Process." Based on the plant conditions
at the time, the inspectors concluded that the most appropriate Appendix G
checklist to use for this issue was the checklist for "Pressurized Water Reactor
Hot Shutdown Operation - Time to core boiling less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />." Because,
operator intervention was required to manually close the affected PORV block
valve, the inspectors concluded that the unit was in a configuration where a
single active failure or personnel error could have resulted in a rapid loss of
reactor coolant system inventory as described in Section II.B.(2) of the checklist.
Consequently, the inspectors concluded that this issue increased the likelihood
of a loss of reactor coolant system inventory and therefore required a Phase 2
SDP analysis. The inspectors discussed the safety significance of this issue with
the Regional Senior Reactor Analyst (SRA). The SRA reviewed the finding and
determined that the drain path could be easily isolated, accurate reactor coolant
system level indication was available, all steam generators were available for
cooling, and all trains of standby injection were available and not impacted by the
finding. Based on these factors the finding was determined to be of very low
safety significance. (Section 4OA3.1)
Green. The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee
failed to provide appropriate instructions for conducting a planned shutdown of
Unit 2 on January 19, 2002, which resulted in unnecessarily challenging the
automatic start function of Unit 2 turbine driven auxiliary feedwater pump
(TDAFWP). This issue was self-revealed when the TDAFWP unexpectedly
started due to low steam generator levels following the manual reactor trip.
The inspectors assessed this finding using the Significance Determination
Process (SDP). The inspectors concluded that this finding was associated with
the initiating events cornerstone and adversely affected the cornerstone
objective and was therefore more than a minor concern. Specifically, the
function of critical safety equipment was challenged and plant stability was upset
during the performance of a normal plant shutdown by the automatic start of
Unit 2 TDAFWP. The inspectors performed a Phase 1 SDP review of this issue
using the guidance provided in NRC Inspection Manual Chapter 0609,
Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., "Significance Determination of Reactor Inspection Findings for
At-Power Situations." Because this finding did not cause or contribute to the
likelihood of an initiating event, the inspectors concluded that this issue was of
very low safety significance. (Section 4OA3.3)
4
Cornerstone: Mitigating Systems
Green. The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to assure that
corrective actions were taken to preclude repetition of emergency diesel
generator (EDG) starting air system relay failures, a significant condition adverse
to quality. This issue was self-revealed when the failure of a starting air system
relay for the Unit 2 AB EDG occurred on October 16, 2002, causing the engine
to roll without a valid start signal. The inspectors subsequently identified that
appropriate corrective actions to prevent repetition had not been taken following
two previous age-related EDG air start relay failures in January 1999 and
September 2000.
The inspectors assessed this finding using the Significance Determination
Process (SDP). The inspectors concluded that this issue, if left uncorrected,
would become a more significant safety concern and was therefore more than a
minor concern. The inspectors also concluded that this finding was associated
with the mitigating systems cornerstone and adversely affected the cornerstone
objective. Specifically, the repetitive EDG air start relay failures affected the
availability, reliability, and capability of systems that respond to initiating events
to prevent undesirable consequences. The inspectors performed a Phase 1
SDP review of this finding using the guidance provided in NRC Inspection
Manual Chapter 0609, Appendix A, "Significance Determination of Reactor
Inspection Findings for At-Power Situations," and determined that this finding
was a licensee performance deficiency of very low safety significance because
the finding: (1) was not a design or qualification deficiency; (2) did not represent
an actual loss of safety function of a system; (3) did not represent an actual loss
of safety function of a single train for greater than its Technical Specification
allowed outage time; (4) did not represent an actual loss of safety function of one
or more Non-Technical Specification trains of equipment designated as risk
significant; and (5) did not screen as potentially risk significant due to a seismic,
flooding, or severe weather initiating event. (Section 4OA2.1)
Cornerstone: Barrier Integrity
Green. The inspectors identified a Non-Cited Violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action." The licensee failed to identify
and take appropriate corrective actions to preclude the failure of four Unit 1
reactor coolant system pressure boundary charging line check valves (Velan
Model B10-3114B-13M), which were at risk of common cause failure due to
industry identified design and manufacturing defects, a significant condition
adverse to quality. This issue was self-revealed when the check valves were all
found to be stuck in either the full or partially open position during radiographic
nonintrusive testing in May 2002.
The inspectors assessed this finding using the Significance Determination
Process (SDP). The inspectors concluded that this finding was associated with
the barrier integrity cornerstone and adversely affected the cornerstone
5
objective, and as such it was more than a minor concern. Specifically, the
charging line check valves perform a safety-related function of limiting the
release of reactor coolant inventory should a charging line failure occur. The
failure of the valves in the open position would prohibit the performance of this
function and therefore affects the objective of the barrier integrity cornerstone.
The inspectors performed a Phase 1 SDP review of this finding using the
guidance provided in NRC Inspection Manual Chapter (IMC) 0609, Appendix A,
"Significance Determination of Reactor Inspection Findings for At-Power
Situations." Because this finding involved the integrity of the reactor coolant
system barrier, the inspectors determined that this finding required a Phase 2
SDP analysis. After consulting with the Regional Senior Reactor Analyst, the
inspectors determined that this issue was of very low safety significance because
no actual loss of safety function occurred. The inspectors concluded that no
actual loss of safety function occurred based on the reported minimal force
required to shut the valves (indicating they would have shut given the differential
pressure applied during accident conditions) and the redundancy provided by a
third check valve (1-CS-321) in the charging line. In accordance with IMC 0609,
Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., Attachment 1, Step 2.6, the SDP results were not evaluated for
potential risk contribution due to Large Early Release Frequency because the
accident sequence result was less than 1E-7 per year. (Section 4OA2.2)
B.
Licensee Identified Violations
None
6
REPORT DETAILS
Summary of Plant Status
Unit 1 operated at or near full power during this inspection period with the following exceptions:
On October 5, 2002, the licensee reduced power to approximately 90 percent of rated
thermal power to repair a steam leak on a feedwater heater. Following the repair, the
licensee returned the unit to full power on October 7, 2002.
On November 10, 2002, the licensee initiated a power reduction to approximately
30 percent of rated thermal power to enter the Containment Building and add oil to a
reactor coolant pump motor. Following the maintenance activity, the licensee returned
the unit to full power on November 12, 2002.
On December 21, 2002, the licensee initiated a power reduction to approximately
53 percent of rated thermal power to enter the Containment Building and add oil to a
reactor coolant pump motor. Following the maintenance activity, the licensee returned
the unit to full power on December 22, 2002.
On December 24, 2002, the licensee reduced power to approximately 55 percent of
rated thermal power to remove a main feedwater pump from service and repair a failed
weld on a small diameter instrument line at the discharge of the pump. Following the
repair, the licensee returned the unit to full power on December 25, 2002.
Unit 2 operated at or near full power during this inspection period.
On November 4, 2002, the licensee received approval of a Notice of Enforcement
Discretion to extend the 72-hour allowed action time of Technical Specification 3.8.1.1.b
to preclude shutting down the unit until the CD emergency diesel generator could be
restored to an operable status.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness
1R01
Adverse Weather Protection (71111.01)
a.
Inspection Scope
The inspectors reviewed the licensees procedures and preparations for cold weather
conditions. The inspectors reviewed winterization procedures, severe weather
procedures, emergency plan implementing procedures related to severe weather, and
performed general area walkdowns. Specifically:
During general pre-winterization walkdowns conducted the week of
October 14, 2002, the inspectors toured selected buildings and areas to verify
7
the licensee had identified all discrepant conditions such as damaged doors,
windows, or vent louvers. Additionally, the inspectors observed housekeeping
conditions and verified that materials capable of becoming airborne missile
hazards during high wind conditions, or impacting snow removal, were
appropriately located and restrained.
During post-winterization walkdowns conducted the week of November 18, 2002,
the inspectors verified that all items on the licensees pre-winterization checklist
were completed with appropriate corrective actions taken for identified discrepant
conditions. Additionally, the inspectors verified that outside water storage tanks
(e.g., refueling water storage tanks, primary water storage tanks, and
condensate storage tanks) and associated valve houses and piping had no
missing or damaged insulation and were serviced by operable heat trace circuits.
During post-winterization walkdowns conducted the week of December 9, 2002,
the inspectors toured plant areas to monitor the physical condition of cold
weather protection features following a period of extended freezing
temperatures. The inspectors observed insulation, heat trace circuits, space
heater operation, and weatherized enclosures to ensure operability of affected
systems.
b.
Findings
No findings of significance were identified.
1R04
Equipment Alignment (71111.04)
.1
Partial System Walkdowns
a.
Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
Initiating Events Cornerstone
Unit 2 AB Emergency Diesel Generator (EDG)
Mitigating Systems Cornerstone
Unit 2 Turbine Driven and West Auxiliary Feedwater (AFW) System Trains
Unit 1 West Essential Service Water (ESW) System Train
The inspectors selected these systems based on their risk significance relative to the
reactor safety cornerstones. The inspectors reviewed operating procedures, Technical
Specification (TS) requirements, Administrative Technical Requirements, system
diagrams, and the impact of ongoing work activities on redundant trains of equipment in
order to identify conditions that could have rendered the systems incapable of
8
performing their intended functions. The inspectors also walked down accessible
portions of the systems to verify system components were aligned correctly.
In addition, the inspectors reviewed the issues that the licensee entered into its
corrective action program to verify that identified problems were being entered into the
program with the appropriate characterization and significance. The inspectors also
reviewed the licensees corrective actions for equipment alignment related issues
documented in selected condition reports (CRs).
b.
Findings
No findings of significance were identified.
1R05
Fire Protection (71111.05)
.1
Routine Resident Inspector Tours
a.
Inspection Scope
The inspectors performed fire protection walkdowns of the following risk-significant plant
areas:
Initiating Events Cornerstone
Lake Screen House (Zone 142)
Mitigating Systems Cornerstone
Fire Pump Building
Auxiliary Building North 609 Foot Elevation (Zone 44N)
Auxiliary Building South 609 Foot Elevation (Zone 44S)
Unit 1 East ESW Pump Room (Zone 29A)
Unit 1 West ESW Pump Room (Zone 29B)
Unit 1 Turbine Building Southeast (Zone 91)
Unit 2 Turbine Building Northeast (Zone 96)
Unit 1 Turbine Building Southwest (Zone 92)
Unit 2 Turbine Building Northwest (Zone 99)
Unit 1 CD EDG Room (Zone 15)
Unit 2 CD EDG Room (Zone 18)
The inspectors verified that fire zone conditions were consistent with assumptions in the
licensees Fire Hazard Analysis. The inspectors walked down fire detection and
suppression equipment, assessed the material condition of fire control equipment, and
evaluated the control of transient combustible materials.
b.
Findings
No findings of significance were identified.
9
1R11
Licensed Operator Requalification (71111.11)
.1
Resident Inspector Quarterly Review
a.
Inspection Scope
The inspectors assessed licensed operator performance and the training evaluators
critique during licensed operator annual requalification evaluations in the D. C. Cook
Plant operations training simulator on October 29, 2002. The inspectors focused on
alarm response, command and control of crew activities, communication practices,
procedural adherence, and implementation of emergency plan requirements.
b.
Findings
No findings of significance were identified.
1R12
Maintenance Effectiveness (71111.12)
a.
Inspection Scope
The inspectors evaluated the licensees handling of selected degraded performance
issues involving the following risk-significant structures, systems, and components
(SSCs):
Initiating Events Cornerstone
Control Group Power Supply Failures
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the SSCs. Specifically, the inspectors independently verified
the licensees management of handling of SSC performance or condition problems in
terms of:
appropriate work practices,
identifying and addressing common cause failures,
scoping of SSCs in accordance with 10 CFR 50.65(b),
characterizing SSC reliability issues,
tracking SSC unavailability,
trending key parameters (condition monitoring),
10 CFR 50.65(a)(1) or (a)(2) classification and reclassification, and
appropriateness of performance criteria for SSCs/functions classified (a)(2)
and/or appropriateness and adequacy of goals and corrective actions for
SSCs/functions classified (a)(1).
10
b.
Findings
b.1
Failure to Address Age-related Failures of Reactor Control Instrumentation Power
Supplies to Prevent Repetition of Power Supply Failures
The inspectors identified a finding of very low safety significance (Green) associated
with a self-revealed event. The licensee failed to assure that prompt corrective actions
were taken to address age-related failures of reactor control instrumentation power
supplies to prevent repetition of power supply failures, a significant condition adverse to
quality. The inspectors determined that this issue constituted a violation of 10 CFR 50,
Appendix B, Criterion XVI, "Corrective Action," and therefore dispositioned this finding
as a Non-Cited Violation.
Discussion
On May 12, 2002, an automatic reactor trip of Unit 2 occurred due to the failure of
redundant 24-volt direct current (DC) power supplies in reactor control instrumentation
cabinet 2-PS-CGC-16. The failure of both power supplies caused the number 21 steam
generator feedwater regulating valve to close. Unit 2 subsequently tripped on low steam
generator water level coincident with low feedwater flow.
Each reactor control instrumentation cabinet contains two separate power supplies
(originally Lambda Model LRS-57-24 or LMS-9120). The two power supplies are
interconnected through auctioneering diodes, such that the cabinet remains energized in
the event of the failure of one of the power supplies. The cabinets provide indication
and control functions for various plant systems including: steam generator feedwater
control, automatic steam generator power operated relief valve (PORV) control,
automatic steam dump control, reactor coolant system volume control tank automatic
make-up, automatic switch-over of the charging pump suction to the refueling water
storage tank on low-low volume control tank level, automatic pressurizer pressure
control using spray valves and heaters, automatic pressurizer level control, and
automatic pressurizer PORV controls. Detection of a single power supply failure was
inhibited because there was no annunciation on the loss of a single power supply.
There were "power available" status lights connected with each of the power supplies
located inside the normally closed cabinet doors; however, the licensee did not routinely
check the status lights prior to the Unit 2 reactor trip.
The reactor control instrumentation cabinet power supplies in question were originally
installed in both units in 1994 as part of a modification to replace obsolete equipment.
In 1999 and 2000, several of these power supplies failed and were sent to a vendor for
repair. Repair reports were generated by the vendor which identified the existence of
internal components that were much older than expected. However, these repair
reports were apparently not forwarded to the system engineering department when they
were received at D. C. Cook. Following the Unit 2 reactor trip, the two failed power
supplies from 2-PS-CGC-16 and several other failed 24-volt DC power supplies were
sent to the vendor for detailed analysis. All of the failures were determined to be
age-related. In all cases, capacitors with date codes as early as 1989 were found.
Hence, these power supplies were already several years old when they were first
installed and energized.
11
The licensee recognized in August 2001 that there had been a significant number of DC
power supply failures during the 24-month period prior to August 2001. The licensee
collectively documented a total of 20 power supply failures in CR 01236037, including
six reactor control instrumentation power supply failures, stating that the failures should
be investigated for a common cause. Other power supply failures were in nuclear
instrumentation, radiation monitoring instrumentation, reactor protection instrumentation,
rod control/rod position indication, steam generator PORV indication, reactor coolant
pump vibration monitoring instrumentation, and main generator hydrogen and carbon
dioxide (CO2) purity monitoring. The inspectors noted that the licensee did not complete
its evaluation of CR 01236037 until after the Unit 2 reactor trip 9 months later.
It is also noteworthy that Unit 2 was started up following the Cycle 13 refueling outage in
February 2002, with one of the two power supplies known to be failed in reactor control
instrumentation cabinet 2-PS-CGC-19. This failed power supply was discovered by
instrument technicians during a routine cleaning and inspection of the cabinet on
February 16, 2002, (12 days prior to completion of the outage). According to the
licensees root cause evaluation, replacement of the power supply was not performed
due to perceived time pressure associated with the refueling outage schedule.
Considering that a second power supply failure in that cabinet would result in a reactor
trip and that the licensee should have been aware of the power supply history based on
CR 01236037, the inspectors concluded that this decision was not conservative in that it
increased the likelihood an initiating event (i.e., a reactor trip).
The inspectors determined that the licensees failure to assure that corrective actions
were taken to preclude repetitive age-related failures of reactor control instrumentation
power supplies is a licensee performance deficiency warranting a significance
evaluation. The inspectors also concluded that this finding affected the cross-cutting
issue of problem identification and resolution.
Analysis
The inspectors assessed this finding using the Significance Determination Process
(SDP). The inspectors concluded that this issue, if left uncorrected, would become a
more significant safety concern with the likelihood of continued failures of reactor control
instrumentation power supplies and was therefore more than a minor concern. The
inspectors also concluded that this finding was associated with the initiating events
cornerstone and adversely affected the cornerstone objective. Specifically, the failure of
redundant power supplies in reactor control instrumentation cabinets would upset plant
stability (cause a reactor trip) and challenge the function of critical safety equipment.
The inspectors performed a Phase 1 SDP review of this finding using the guidance
provided in NRC Inspection Manual Chapter (IMC) 0609, Appendix A, "Significance
Determination of Reactor Inspection Findings for At-Power Situations." Because this
finding contributes to both the likelihood of a reactor trip and the likelihood that
mitigation equipment or functions will not be available, the inspectors determined that
this finding required a Phase 2 SDP analysis.
Using the current risk-informed inspection notebook for D. C. Cook (Revision 0) for the
Phase 2 SDP analysis, the inspectors determined that this finding was potentially
greater than very low safety significance. Specifically, the inspectors determined that
12
this issue caused the likelihood for transients involving a loss of the primary conversion
system (PCS) to be increased by an order of magnitude using Usage Rule 1.2 of
IMC 0609, Appendix A, Attachment 2. The initiating event likelihood was evaluated for a
greater than 30-day period because the condition had existed since the last refueling
outage which occurred in February 2002. However, after a review of additional
information the inspectors determined that a Phase 3 analysis was required.
Specifically, for transient sequences, including those involving a loss of the PCS, the
inspectors noted that additional credit that was not assumed in the risk-informed
notebook could be given to the AFW function. This additional credit involves the
operators ability to cross-tie the opposite units motor-driven AFW pumps to the affected
units AFW system. This cross-tie evolution is probabilistically limited to the operators
ability to perform the evolution and is given a failure probability of 0.1. Applying this
additional credit (i.e., one point), the inspectors determined that this finding was of very
low safety significance from a Core Damage Frequency (CDF) perspective.
The inspectors also evaluated the effect of this finding on the Large Early Release
Frequency (LERF) while factoring in the additional AFW credit discussed above. Using
IMC 0609, Appendix H, "Containment Integrity SDP," the inspectors determined that this
finding was also potentially greater than very low safety significance. Specifically, for ice
condenser plants involving transient accident sequences, the LERF result is a direct
correlation to the CDF result. However, the inspectors determined through discussions
with NRC risk analysts that due to refinements of LERF estimations, the impact from
transient sequences were no longer being considered as a direct correlation to the
LERF result. In fact, this refinement indicates that the LERF contribution from most
transient sequences for ice condenser plants is not risk significant. The refinement
indicates that only station blackout accident sequences would result in this one-to-one
correlation between the CDF result and the LERF estimation. When considering this
refinement for LERF estimations, the inspectors determined that this finding was also of
very low safety significance from a LERF perspective.
Enforcement
10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that measures
shall be established to assure that conditions adverse to quality, such as failures,
malfunctions, deficiencies, deviations, defective material and equipment, and
nonconformances are promptly identified and corrected. In the case of significant
conditions adverse to quality, the measures shall assure that the cause of the condition
is determined and corrective action taken to preclude repetition. Contrary to the above,
the licensee failed to promptly take corrective action to address age-related failures of
reactor control instrumentation power supplies to prevent repetition of power supply
failures, a significant condition adverse to quality. Specifically, for a 24-month period
prior to August 2001, the licensee documented six reactor control instrumentation power
supply failures. All of these failures were subsequently determined to be age-related.
Consequently, four additional reactor control instrumentation power supply failures have
occurred for the same cause since August 2001: (1) two redundant power supplies
failed in reactor control instrumentation cabinet 2-PS-CGC-16, which resulted in a
reactor trip and challenged the function of critical safety equipment; (2) one power
supply failed in reactor control instrumentation cabinet 2-PS-CGC-19; and (3) one power
supply failed in reactor control instrumentation cabinet 1-PS-CGC-16. Because of the
13
very low safety significance, this violation is being treated as a Non-Cited Violation
consistent with Section VI.A of the NRC Enforcement Policy
(NCV 50-316-02-09-01(DRP)). The licensee entered this violation into its corrective
action program as CR 02133001 and CR 02133002.
b.2
Failure to Implement a Corrective Action to Prevent Recurrence Associated with Reactor
Control Instrumentation Power Supply Failures
The inspectors identified a finding of very low safety significance (Green). The licensee
failed to take corrective action in response to a Unit 2 reactor trip on May 12, 2002, to
preclude the repetition of reactor control instrumentation 24-volt DC power supply
failures, a significant condition adverse to quality. Specifically, the licensee did not
perform prescribed weekly verifications of reactor control instrumentation power
supplies to identify failed power supplies in lieu of no annunciation on the loss of a single
power supply. The inspectors determined that this issue constituted a violation of
10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and therefore dispositioned
this finding as a Non-Cited Violation.
Discussion
Following the Unit 2 reactor trip on May 12, 2002, the licensee initiated the following
corrective actions to prevent recurrence:
The failed 24-volt DC power supplies in reactor control instrumentation cabinet
2-PS-CGC-16 and 2-PS-CGC-19 were replaced.
All 24-volt DC control group power supplies in Unit 2 were inspected and
components were verified to be no older than 2 years old. One power supply
was replaced as a result of the inspection.
All 24-volt DC control group power supplies in Unit 1 were replaced prior to the
Unit 1 reactor startup on June 8, 2002.
The licensee identified the need to establish a recurring task to perform a weekly
verification of control group power supplies to ensure that the "power available"
status lights were lit. This could afford the licensee an opportunity to take
compensatory measures or replace a failed power supply prior to the failure of
the redundant power supply.
During the inspectors review of the licensees corrective actions for the Unit 2 reactor
trip and in response to the inspectors questions, the licensee discovered that weekly
verifications of the control group power supply "power available" status lights were not
being performed. Verification of the power supply status lights was stipulated as a
restart action by the licensees Plant Operations Review Committee following the reactor
trip and was specified as a corrective action in the Licensee Event Report (LER) that
reported the event. The licensee subsequently performed this check on
November 22, 2002, and discovered a failed 24-volt DC power supply in Unit 1 cabinet
1-PS-CGC-16.
14
The inspectors concluded that the licensees failure to implement this corrective action
to prevent recurrence for a significant condition adverse to quality was a performance
deficiency warranting a significance evaluation. The inspectors also concluded that this
finding affected the cross-cutting issue of problem identification and resolution.
Analysis
The inspectors assessed this finding using the SDP. The inspectors concluded that this
issue could be reasonably viewed as a precursor to a significant event (i.e., potentially
result in a reactor trip similar to the Unit 2 trip on May 12, 2002), and was therefore more
than a minor concern. The inspectors also concluded that this finding was associated
with the initiating events cornerstone and adversely affected the cornerstone objective.
Specifically, the failure of redundant power supplies in reactor control instrumentation
cabinets could upset plant stability (i.e., cause a reactor trip) and challenge the function
of critical safety equipment. Consistent with the SDP evaluation performed for the
finding described in Section 1R12.b.1, this finding was determined to be of very low
safety significance.
Enforcement
10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that measures
shall be established to assure that conditions adverse to quality, such as failures,
malfunctions, deficiencies, deviations, defective material and equipment, and
nonconformances are promptly identified and corrected. In the case of significant
conditions adverse to quality, the measures shall assure that the cause of the condition
is determined and corrective action taken to preclude repetition. Contrary to the above,
the licensee failed to take corrective action to preclude the repetition of reactor control
instrumentation 24-volt DC power supply failures. Specifically, the licensee failed to
perform a weekly verification of control group power supplies to ensure that the "power
available" status lights were lit. This corrective action was identified by the licensee in
response to the Unit 2 reactor trip on May 12, 2002, which was caused by the failure of
redundant power supplies in reactor control instrumentation cabinet 2-PS-CGC-16. The
licensee subsequently performed this check on November 22, 2002, and discovered a
failed 24-volt DC power supply in Unit 1 cabinet 1-PS-CGC-16. Because of the very low
safety significance, this violation is being treated as a Non-Cited Violation consistent
with Section VI.A of the NRC Enforcement Policy (NCV 50-316-02-09-02(DRP)). The
licensee entered this violation into its corrective action program as CR 02325058.
1R13
Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)
a.
Inspection Scope
The inspectors reviewed the licensees evaluation and management of plant risk for
maintenance activities affecting the following equipment:
15
Initiating Events Cornerstone
Unit 1 AB EDG ESW Supply Valves
Unit 2 CD EDG
Mitigating Systems Cornerstone
Unit 1 East ESW Pump
Unit 1 West ESW Pump
Unit 2 East ESW Pump
Unit 2 East AFW System Train
These activities were selected based on their potential risk significance relative to the
reactor safety cornerstones. The maintenance associated with the Unit 2 CD EDG was
emergent work to replace the engines governor that was identified as failed during a
scheduled surveillance test. As applicable for each of the above activities, the
inspectors reviewed the scope of maintenance work, discussed the results of the
assessment with the licensees probabilistic risk analyst and/or shift technical advisor,
and verified that plant conditions were consistent with the risk assessment. The
inspectors also reviewed TS requirements and walked down portions of redundant
safety systems, when applicable, to verify that risk analysis assumptions were valid and
applicable requirements were met.
In addition, the inspectors reviewed the issues that the licensee entered into its
corrective action program to verify that identified problems were being entered into the
program with the appropriate characterization and significance. The inspectors also
reviewed the licensees corrective actions for maintenance risk related issues that were
documented in selected CRs.
b.
Findings
No findings of significance were identified.
1R14
Personnel Performance During Non-routine Plant Evolutions (71111.14)
.1
Unit 1 Power Reduction to Support Oil Addition to a Reactor Coolant Pump Motor
a.
Inspection Scope
On November 10, 2002, the licensee initiated a power reduction on Unit 1 to
approximately 30 percent of rated thermal power to enter the Containment Building and
add oil to the number 14 reactor coolant pump motor. Following the maintenance
activity, the licensee returned the unit to full power on November 12, 2002. The
inspectors observed portions of the power reduction and assessed operator
performance.
16
b.
Findings
No findings of significance were identified.
.2
Unit 1 Control Group Power Supply Replacement
a.
Inspection Scope
On December 3, 2002, the licensee replaced one of two redundant 24-volt DC power
supplies in reactor control instrumentation cabinet 1-PS-CGC-16. The licensee
identified the failed power supply on November 22, 2002 and installed a temporary
back-up power supply until replacement of the failed power supply could be performed.
This was the first time that replacement of a control group power supply was performed
with the unit on line. The inspectors reviewed the licensees preparations for this
evolution and assessed operator performance when the temporary back-up power
supply unexpectedly failed, causing a complete loss of power to the instrumentation
cabinet.
b.
Findings
No findings of significance were identified.
1R15
Operability Evaluations (71111.15)
a.
Inspection Scope
The inspectors reviewed the following CRs to ensure that either: (1) the condition did
not render the involved equipment inoperable or result in an unrecognized increase in
plant risk, or (2) the licensee appropriately applied TS limitations and appropriately
returned the affected equipment to an operable status.
Mitigating Systems Cornerstone
CR 02136008
Wire Found Disconnected in 1-RPS-A
CR 02131018
Review Operability and Reportability Issues for Two Items
Dealing with Feedwater Pressure Indication and the Plant
Process Computer Calorimetric Program
CR 02290012
Steam Generator PORV Actuator Capability Calculation
Revealed Negative Calculated Margin for Full Stroke
Capability
CR 02339016
Ultrasonic Examination on Unit 1 ESW to West Motor
Driven AFW Pump Piping Found Some Silt/Sand in the
Piping
Barrier Integrity Cornerstone
CR 02135049
1-CCR-462 Leaking Excessively During Local Leak Rate
Testing
17
CR 02300002
Unit 2 Control Room Access Door 2-DR-AUX411B Latch
Has Broken and Door Will Not Shut
In addition, the inspectors reviewed the issues that the licensee entered into its
corrective action program to verify that identified problems were being entered into the
program with the appropriate characterization and significance. The inspectors also
reviewed the licensees corrective actions for issues potentially affecting the operability
of safety-related SSCs that were documented in selected CRs.
b.
Findings
No findings of significance were identified.
1R16
Operator Workarounds (71111.16)
.1
Review of Selected Operator Workarounds
a.
Inspection Scope
The inspectors evaluated the operator work-arounds (OWAs) listed below to identify any
potential affect on the functionality of mitigating systems or on the operators response
to initiating events:
OWA 01-02
Feedwater Preheat Valves Cause Cooldown During a Reactor
Trip
OWA 02-03
Need to Open Condenser Vacuum Breakers in Unit 1 to Control
Main Turbine Vibration Post Trip
The inspectors selected OWA 01-02 to review the potential affect that leakby past the
feedwater preheat control valves has on contributing to excessive plant cooldowns
following reactor trips from low power. The inspectors selected OWA 02-03 to review
the potential for loss of the secondary heat sink with operation of the main turbine
vacuum breakers. The inspectors interviewed operating and engineering department
personnel and reviewed selected procedures and documents.
b.
Findings
No findings of significance were identified.
.2
Semiannual Review of the Cumulative Effect of Operator Workarounds
a.
Inspection Scope
The inspectors reviewed the cumulative effect of OWAs, control room deficiencies, and
degraded conditions on equipment availability, initiating event frequency, and the ability
of the operators to implement abnormal or emergency operating procedures. During
this review the inspectors considered the cumulative effects of OWAs on the following:
the reliability, availability and potential for mis-operation of a system;
18
the ability of operators to respond to plant transients or accidents in a correct
and timely manner; and
the potential to increase an initiating event frequency or affect multiple mitigating
systems.
In addition, the inspectors reviewed the issues that the licensee entered into its
corrective action program to verify that identified problems were being entered into the
program with the appropriate characterization and significance. The inspectors also
reviewed the licensees corrective actions for issues potentially affecting the functionality
of mitigating systems or on the operators response to initiating events that were
documented in selected CRs.
b.
Findings
No findings of significance were identified.
1R19
Post Maintenance Testing (71111.19)
a.
Inspection Scope
The inspectors reviewed the post maintenance testing associated with the following
scheduled maintenance activities:
Barrier Integrity Cornerstone
Unit 1 Leak Test of Post Accident Containment Hydrogen Monitoring System
Mitigating Systems Cornerstone
Unit 2 CD EDG Governor Replacement
Unit 2 East AFW Pump Maintenance
Unit 2 West ESW Pump Maintenance
The inspectors selected these post maintenance testing activities because the systems
were identified as risk significant in the licensees risk analysis. The inspectors reviewed
the scope of the work performed and evaluated the adequacy of the specified post
maintenance testing. The inspectors verified that the post maintenance testing was
performed in accordance with approved procedures, that the procedures clearly stated
acceptance criteria, and that the acceptance criteria were met. During this inspection,
the inspectors interviewed operations, maintenance, and engineering department
personnel and reviewed the completed post maintenance testing documentation.
b.
Findings
No findings of significance were identified.
19
1R22
Surveillance Testing (71111.22)
a.
Inspection Scope
For the surveillance test procedures listed below, the inspectors observed selected
portions of the surveillance test and/or reviewed the test results to determine whether
risk significant systems and equipment were capable of performing their intended safety
functions and to verify that testing was conducted in accordance with applicable
procedural and TS requirements:
Barrier Integrity Cornerstone
01-OHP-4030-STP-011, "Containment Isolation and Inservice Inspection Valve
Operability Test"
02-IHP-4030-234-001, "Unit 2 Distributed Ignition System Surveillance and
Baseline Testing"
12-IHP-4030-046-227, "Unit 1 and 2 Personnel Airlock Door Seal Leak Rate
Surveillance"
Mitigating Systems Cornerstone
01-EHP-4030-ATR-225-020, "Unit 1 Auxiliary Cable Vault CO2 Fire Suppression
Test"
02-IHP-4030-SMP-219, "Steam Generator 1 & 2 Steam/Feed Flow Mismatch
and Steam Pressure Protection Set I Functional Test and Calibration"
02-IHP-4030-SMP-222, "Steam Generator 2 & 4 Steam/Feed Flow Mismatch
and Steam Pressure Protection Set II Functional Test and Calibration"
02-IHP-4030-SMP-227, "Steam Pressure Protection Set III Functional Test and
Calibration"
02-IHP-4030-SMP-228, "Steam Pressure Protection Set IV Functional Test and
Calibration"
12-EHP-5030-CAR-001, "Characterization Testing Program"
The inspectors reviewed the test methodology and test results in order to verify that
equipment performance was consistent with safety analysis and design basis
assumptions.
In addition, the inspectors reviewed the issues that the licensee entered into its
corrective action program to verify that identified problems were being entered into the
program with the appropriate characterization and significance. The inspectors also
reviewed the licensees corrective actions for surveillance testing related issues
documented in selected CRs.
b.
Findings
No findings of significance were identified.
20
1R23
Temporary Plant Modifications (71111.23)
a.
Inspection Scope
The inspectors reviewed the temporary modifications listed below to verify that the
installations were consistent with design modification documents and that the
modifications did not adversely impact system operability or availability:
Barrier Integrity Cornerstone
12-EHP-5040-EMP-006, "Disable Bridge East Travel Limit Switch on East
Auxiliary Building Crane 12-QM-3E"
Mitigating Systems Cornerstone
1-TM-02-85-R0, "Install Backup Power Supply for Control Group 1"
12-TM-00-61-R2, "Winterization/De-Winterization Temporary Modification to
Support 12-IHP-5040-EMP-004"
The first temporary modification disabled the end of travel limit switch to allow the
Auxiliary Building crane to move far enough to allow resin shipping casks to be lowered
into the drumming room. The second temporary modification installed a backup power
supply in reactor control instrumentation cabinet 1-PS-CGC-16 until a permanent power
supply replacement could be performed for one of the two redundant 24-volt DC power
supplies. The third temporary modification installed ventilation system covers and other
cold weather system protection measures for the upcoming winter season.
The inspectors verified that configuration control of the modifications were correct by
reviewing design modification documents and confirmed that appropriate
post-installation testing was accomplished. The inspectors interviewed engineering and
operations department personnel and reviewed the design modification documents
against the applicable portions of the Updated Final Safety Analysis Report (UFSAR).
b.
Findings
No findings of significance were identified.
1EP2
Alert and Notification System (ANS) Testing (71114.02)
a.
Inspection Scope
The inspectors discussed with Emergency Preparedness (EP) staff the design,
equipment, and periodic testing of the public ANS for the D. C. Cook reactor facility
emergency planning zone to verify that the system was properly tested and maintained.
The inspectors also reviewed procedures and records for a 24-month period ending
September 2002 related to ANS testing, annual preventive maintenance, and
non-scheduled maintenance. The inspectors reviewed the licensees documentation for
determining whether each model of siren installed in the emergency planning zone
would perform as expected if fully activated. Records used to document and trend
21
component failures for each model of installed siren were also reviewed to ensure that
corrective actions were taken for test failures or system anomalies.
b.
Findings
No findings of significance were identified.
1EP3
Emergency Response Organization (ERO) Augmentation Testing (71114.03)
a.
Inspection Scope
The inspectors reviewed the licensees ERO augmentation testing to verify that the
licensee maintained and tested its ability to staff the ERO during an emergency in a
timely manner. Specifically, the inspectors reviewed quarterly, off-hours staff
augmentation test procedures, dated December 14, 2001, August 23, 2001,
March 14, 2002, April 16, 2002, and July 17, 2002 drill records, primary and backup
provisions for off-hours notification of the D. C. Cook reactor facility emergency
responders, and the current ERO rosters for D. C. Cook. The inspectors reviewed and
discussed the facility EP staff's provisions for maintaining ERO call out lists.
b.
Findings
No findings of significance were identified.
1EP5
Correction of Emergency Preparedness Weaknesses and Deficiencies (71114.05)
a.
Inspection Scope
The inspectors reviewed the Performance Assurance staff's 2001 - 2002 audits to
ensure that these audits complied with the requirements of 10 CFR 50.54(t) and that the
licensee adequately identified and corrected deficiencies. The inspectors also reviewed
the EP staff's 2001 and 2002 self-assessments, and critiques to evaluate the EP staff's
efforts to identify and correct weaknesses and deficiencies. Additionally, the inspectors
reviewed a sample of EP items, CRs, and action requests related to the facility's EP
program to determine whether corrective actions were acceptably completed.
b.
Findings
No findings of significance were identified.
1EP6
Drill Evaluation (71114.06)
a.
Inspection Scope
The inspectors observed the conduct of the licensee's annual announced emergency
training exercise that was conducted in the licensee's control room simulator and
emergency response facilities on October 23, 2002. The inspection effort was focused
on evaluation of the licensee's classifications, notifications, and protective action
recommendations for the simulated event. The inspectors also evaluated the licensee's
22
conduct of the training evolution, including the licensees critique of performance to
identify weaknesses and deficiencies.
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1
Plant Walkdowns, Radiological Boundary Verifications, and Radiation Work Permit
Reviews
a.
Inspection Scope
The inspectors conducted walkdowns of the radiologically protected area to verify the
adequacy of radiological area boundaries and postings. Specifically, the inspectors
walked down radiologically significant work area boundaries (radiation, high and locked
high radiation areas) in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor,
as well as the Unit 2 Containment Building. The inspectors performed confirmatory
radiation surveys in selected portions of these areas to verify that these areas were
properly posted and controlled in accordance with 10 CFR 20, licensee procedures, and
TSs. The inspectors also examined the radiological conditions of work areas within
those radiation, high and locked high radiation areas to assess the adequacy of licensee
implemented contamination controls. Additionally, the inspectors reviewed radiation
work permits (RWPs) for general tours, access to locked high radiation areas for work
on spent fuel pool demineralizers, drumming room clean-up activities, an at power entry
into Unit 1 for work on a reactor coolant pump; and for another at power entry into Unit 2
for work on a safety injection system accumulator. The RWPs were evaluated for
protective clothing requirements, respiratory protection concerns, electronic dosimetry
alarm set points, use of remote telemetry dosimetry, radiation protection hold points,
and As-Low-As-Reasonably-Achievable considerations, to verify that work instructions
and controls had been adequately specified and that electronic dosimeter set points
were in conformity with survey results.
b.
Findings
No findings of significance were identified.
23
.2
Job-In-Progress Reviews, Observations of Radiation Worker Performance, and
Radiation Protection Technician Proficiency
a.
Inspection Scope
The inspectors observed selected portions of the following radiologically significant work
activities performed during the inspection and evaluated the licensees use of
radiological controls:
number 21 safety injection system accumulator level indicator repair, and
preparations for spent fuel pool demineralizer work.
The inspectors reviewed the pre-job briefing package for the work evolutions, reviewed
the radiological requirements for the activities and assessed the licensees performance
with respect to those requirements. The inspectors reviewed survey records, including
radiation, contamination, and airborne surveys, to verify that appropriate radiological
controls were effectively utilized. The inspectors also reviewed in-process surveys and
applicable postings and barricades to verify their accuracy. The inspectors observed
radiation protection technician (RPT) and worker performance during the work evolution
at the job sites to verify that the technicians and workers were aware of the significance
of the radiological conditions in their workplace and RWP controls/limits, and that they
were performing adequately given the radiological hazards present and the level of their
training.
b.
Findings
No findings of significance were identified.
.3
Identification and Resolution of Problems
a.
Inspection Scope
The inspectors reviewed licensee CRs written since the last inspection (July 2002) to the
date of the current inspection, which focused on access control to radiologically
significant areas (i.e., problems concerning activities in high radiation areas, radiation
protection technician performance, and radiation worker practices). The inspectors also
reviewed the recently revised "High, Locked High, and Very High Radiation Area
Access" procedure, which addressed new requirements for specific locking devices for
these areas. The inspectors reviewed these documents to assess the licensee's ability
to identify repetitive problems, contributing causes, and the extent of conditions, and
then implement corrective actions in order to achieve lasting results.
b.
Findings
No findings of significance were identified.
24
2OS3 Radiation Monitoring Instrumentation (71121.03)
.1
Walkdowns of Radiation Monitoring Instrumentation
a.
Inspection Scope
The inspectors reviewed the UFSAR and performed walkdowns of continuous air
monitors in the Auxiliary Building, radwaste area, spent fuel pool/refuel floor,
Radioactive Material Building, and one area radiation monitor (ARM) in the Unit 2
Containment Building. Additionally, the inspectors examined a representative number of
portable radiation survey instruments staged throughout the licensees facility to verify
that those instruments had current calibrations, were operable, and in good physical
condition. The inspectors also reviewed the status of repair or troubleshooting activities
associated with selected radiation monitoring instruments (i.e., small article monitors
and portal monitors that had work request tags) to verify that instrumentation problems
were being addressed in an appropriate and timely manner.
b.
Findings
No findings of significance were identified.
.2
Calibration, Operability, and Alarm Set Points of Radiation Monitoring Instrumentation
a.
Inspection Scope
The inspectors examined radiological instrumentation associated with monitoring
transient high and/or very high radiation areas to verify that the instrumentation was
operating consistent with industry standards and in accordance with station procedures.
Specifically, the inspectors assessed the operability of the following instrumentation:
Unit 2 In-Core Instrumentation Room ARM.
The inspectors reviewed the licensees alarm set point for this specific ARM to verify that
the set point was established consistent with the UFSAR, TSs, and the licensees
The inspectors discussed surveillance practices with licensee personnel and reviewed
calender year 2001 - 2002 calibration records and procedures for selected radiation
monitors used for assessment of internal exposure. The inspectors also reviewed
calibration records and procedures for those instruments utilized for surveys of
personnel and equipment prior to egress from the radiologically controlled area. These
instruments included:
AMS-4 Air Monitoring System,
APTEC PMW-3 Personnel Monitor, and
Gamma 40/60 Portal Monitor.
25
Additionally, the alarm set points for these instruments were reviewed to verify that they
were established at levels consistent with industry standards and regulatory guidance
provided in Health Physics Positions 72 and 250 of NUREG/CR-5569.
The inspectors evaluated the calibration procedures and calibration records for selected
portable radiation survey instruments to verify that they had been properly calibrated
consistent with the licensees procedures. Specifically, the inspectors reviewed the
calibrations of the following instruments:
Emergency Plan designated RO-7 ion chamber, and
Smart Radiation Monitor general area dose rate meter.
The inspectors also assessed periodic performance tests completed for selected
portable radiation survey instruments to verify that they had been tested consistent with
the licensees procedures. Specifically, the inspectors observed the performance testing
of the following instruments:
extender instruments, and
bicron RSO survey instruments.
b.
Findings
No findings of significance were identified.
.3
Radiation Protection Technician Instrument Use
a.
Inspection Scope
The inspectors observed RPTs performing in-field source checks of portable radiation
survey instruments to verify that those source checks were adequately completed using
appropriate radiation sources and station procedures. The inspectors assessed the
RPTs use of radiation/contamination detection instruments as they provided radiological
job coverage for risk significant work (e.g., the safety injection system accumulator
repair work in the Unit 2 Containment Building), as well as routine work, to ensure that
the RPTs were utilizing the appropriate instruments. The inspectors monitored RPTs
performing functional tests of selected contamination monitors, portal monitors, and
small article monitors (i.e., for surveys of personnel and equipment prior to unconditional
release from the radiologically controlled area) to verify that they were source tested and
calibrated as required by station procedures and industry standards.
b.
Findings
No findings of significance were identified.
26
.4
Problem Identification and Resolution
a.
Inspection Scope
The inspectors reviewed calendar year 2001-2002 CRs that addressed radiation
monitoring instrument deficiencies to determine if any significant radiological incidents
involving instrument deficiencies had occurred. The inspectors examined the results of
a self-assessment (i.e., the Summary Report for performance Assurance Audit
PA-0206, "Radiation Protection") that focused on the licensees CR database and
several individual CRs related to radiation monitoring instrumentation generated during
the current assessment period. The inspectors also interviewed plant staff and
examined closed CRs to verify that radiological instrumentation related issues were
adequately addressed by the licensee. The inspectors evaluated these documents to
verify the licensees ability to identify repetitive problems, contributing causes, extent of
conditions, and the implementation of corrective actions to achieve lasting results.
b.
Findings
No findings of significance were identified.
3.
SAFEGUARDS
Cornerstone: Physical Protection
3PP3
Response to Contingency Events (71130.03)
a.
Inspection Scope
The inspectors reviewed the status of security operations and assessed licensee
implementation of the protective measures in place as a result of the current, elevated
threat environment.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1
Safety System Functional Failures
a.
Inspection Scope
Mitigating Systems Cornerstone
The inspectors verified the Safety System Functional Failures performance indicator for
both units. The inspectors reviewed each LER from October 2001 to September 2002,
27
determined the number of safety system functional failures that occurred, evaluated
each LER against the performance indicator definitions, and verified the number of
safety system functional failures reported.
b.
Findings
No findings of significance were identified.
.2
Reactor Coolant System Leakage
a.
Inspection Scope
Barrier Integrity Cornerstone
The inspectors verified the Reactor Coolant System Leakage performance indicator for
both units. The inspectors reviewed operating logs and the results of reactor coolant
system water inventory balance calculations performed from October 2001 through
September 2002 and verified the licensees calculation of reactor coolant system
leakage for both units.
b.
Findings
No findings of significance were identified.
.3
Reactor Coolant System Specific Activity
a.
Inspection Scope
Barrier Integrity Cornerstone
The inspectors verified the Reactor Coolant System Specific Activity performance
indicator for both units. The inspectors reviewed specific activity results reported from
October 2001 through September 2002 and verified the licensees calculation of reactor
coolant system activity for both units. In addition, the inspectors observed staff
chemistry technicians collecting reactor coolant system samples to verify that the
technicians had complied with applicable procedures during the collection and
processing of the samples.
b.
Findings
No findings of significance were identified.
.4
ANS, Drill and Exercise Performance (DEP), and ERO Drill Participation
a.
Inspection Scope
Emergency Preparedness Cornerstone
28
The inspectors verified that the licensee had accurately reported the ANS, DEP, and
ERO Drill Participation performance indicators for both units. Specifically, the inspectors
reviewed the licensees performance indicator records, data reported to the NRC, and
CRs for the period July 2001 through September 2002 to identify any occurrences that
were not identified by the licensee. Records of relevant control room simulator training
sessions, periodic ANS tests, and excerpts of drill and exercise scenarios and
evaluations were also reviewed.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1
EDG Starting Air Relay Failures
a.
Inspection Scope
A failure of a starting air system relay for the Unit 2 AB EDG occurred on
October 16, 2002, causing the engine to roll without a valid start signal. The inspectors
reviewed previous failures associated with starting air system relays for the EDGs. In
addition, the inspectors reviewed the root cause evaluation for the following CR:
CR P-99-01279, "Unit 2 AB EDG Rolled with Air by Itself. An Auxiliary
Equipment Operator Was Dispatched Who Reported the Engine Rolling with Air.
No Indication of a Start Signal Was Detected Locally or in the Control Room.
Starting Air Continued to Blow Down Engine Until Air Depleted."
The inspectors verified the following attributes during their review of the licensees
corrective actions for the above CR and several other related CRs:
consideration of the extent of condition, generic implications, common cause and
previous occurrences;
classification and prioritization of the resolution of the problem, commensurate
with safety significance;
identification of the root and contributing causes of the problem; and
identification of corrective actions which were appropriately focused to correct
the problem.
The inspectors discussed the corrective actions and associated CR evaluations with site
personnel including the CR evaluators and system engineers.
b.
Findings
The inspectors identified a finding of very low safety significance (Green) associated
with this self-revealed event. The licensee failed to assure that corrective actions were
taken to preclude repetition of EDG starting air system relay failures, a significant
condition adverse to quality. The inspectors determined that this issue constituted a
29
violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and therefore
dispositioned this finding as a Non-Cited Violation.
Description
On October 16, 2002, one of the two Unit 2 AB EDG air receivers (AB2) depressurized
when starting air valve 2-XRV-222 inadvertently opened and consequently air rolled the
engine. The second safety-related air receiver for the EDG remained pressurized and
auxiliary equipment operators manually isolated 2-XRV-222, which re-pressurized the
AB2 air receiver. A subsequent licensee investigation determined that starting air
system relay 2-19-DGAB had failed causing the starting air valve to open.
The starting air system provides the motive force for starting the EDGs. The inspectors
noted that each of the four plant EDGs have two of the same model relays installed in
the starting air system, which are designated as relays 19 and 19-1. The inspectors
determined that while the failure of relay 19 resulted in depressurization of only one of
the two air start receivers for an EDG, the failure of relay 19-1 resulted in
depressurization of both air start receivers for an EDG.
The inspectors reviewed work order history and the corrective action program database
to determine whether or not previous failures had occurred on starting air system relays.
The inspectors found two recent occurrences:
in January 1999, relay 19-1 failed on the Unit 2 AB EDG starting air system,
which resulted in depressurization of the starting air system; and
in September 2000, relay 19 failed on the Unit 1 CD EDG starting air system,
which resulted in depressurization of one air start receiver.
The inspectors reviewed the licensees assessment of CR P-99-01279 associated with
the January 1999 failure and noted the following:
the root cause evaluation concluded that the most likely failure scenario of the
relay was long term overheating of the continuously energized coil in relay 19-1;
the root cause evaluation concluded that while the relay was designed and rated
for 250-volt DC, equalization of the station batteries at the plant had raised DC
bus voltages to 280-volt DC. The evaluator determined that from a thermal
deterioration perspective, this would increase the heat generated in the relay by
approximately 25 percent;
the root cause evaluation concluded that given the root cause of the failure, the
population of similarly aged relays that were continuously energized should be
considered suspect and candidates for failure;
the root cause evaluation concluded that similar relays should be expected to
function for an equivalent duration given similar operating conditions; and
the root cause evaluation recognized that the recommendations in the evaluation
were based on only one data point and that prior to a wholesale changeout of
relays, the analysis of additional relays of the same type in the same
configuration should be considered.
30
The inspectors determined that the licensees root cause evaluation addressed the
potential extent of condition and the generic implications of the relay failure. However,
the only corrective action taken for this significant condition adverse to quality was the
replacement of the failed relay 19-1 for the Unit 2 AB EDG. The licensee did not
evaluate the condition of other relays of the same type in the same configuration, nor
investigate the need for the implementation of a preventive maintenance program or
replacement program for this type of relay used in the EDG starting air system.
The inspectors also reviewed CR 00266004, which documented the failure of relay 19
for the Unit 1 CD EDG in September 2000. The inspectors noted that neither an
apparent cause evaluation nor a root cause evaluation was performed and that the
licensee did not evaluate the potential extent of condition considering the previous
failure in January 1999. The licensee concluded that the failure was age related and the
only corrective action completed was replacement of the failed relay.
The inspectors determined that the licensees failure to assure that corrective actions
were taken to preclude repetition of starting air system relay failures for the EDGs is a
licensee performance deficiency warranting a significance evaluation. The inspectors
also concluded that this finding affected the cross-cutting issue of problem identification
and resolution.
Analysis
The inspectors assessed this finding using the SDP. The inspectors concluded that this
issue, if left uncorrected, would become a more significant safety concern with the
continued failures of the starting air system relays and was therefore more than a minor
concern. The inspectors also concluded that this finding was associated with the
mitigating systems cornerstone and adversely affected the cornerstone objective.
Specifically, the repetitive EDG air start relay failures affected the availability, reliability,
and capability of systems that respond to initiating events to prevent undesirable
consequences. The inspectors performed a Phase 1 SDP review of this finding using
the guidance provided in NRC Inspection Manual Chapter 0609, Appendix A,
"Significance Determination of Reactor Inspection Findings for At-Power Situations,"
and determined that this finding was a licensee performance deficiency of very low
safety significance because the finding: (1) was not a design or qualification deficiency;
(2) did not represent an actual loss of safety function of a system; (3) did not represent
an actual loss of safety function of a single train for greater than its TS allowed outage
time; (4) did not represent an actual loss of safety function of one or more non-TS trains
of equipment designated as risk significant; and (5) did not screen as potentially risk
significant due to a seismic, flooding, or severe weather initiating event.
Enforcement
10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that measures
shall be established to assure that conditions adverse to quality, such as failures,
malfunctions, deficiencies, deviations, defective material and equipment, and
nonconformances are promptly identified and corrected. In the case of significant
conditions adverse to quality, the measures shall assure that the cause of the condition
is determined and corrective action taken to preclude repetition. Contrary to the above,
31
the licensee failed to take corrective action to prevent repetitive failures of EDG starting
air system relays 19 and 19-1, a significant condition adverse to quality. Specifically, in
January 1999, relay 19-1 failed on the Unit 2 AB EDG starting air system, which resulted
in depressurization of the starting air system. The licensee failed to promptly perform
corrective actions to preclude a repetition of starting air system relay failures for the
EDGs. Consequently, two additional starting air system relay failures have occurred:
(1) in September 2000, relay 19 failed on the Unit 1 CD EDG starting air system, which
resulted in depressurization of one air start receiver; and (2) in October 2002, relay 19
failed on the Unit 2 AB EDG starting air system, which resulted in depressurization of
one air start receiver. Because of the very low safety significance, this violation is being
treated as a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement
Policy (NCV 50-315/316-02-09-03(DRP)). The licensee entered this violation into its
corrective action program as CR 02289033.
.2
Common Cause Failure of Four Unit 1 Charging System Check Valves
a.
Inspection Scope
During the Unit 1 Cycle 18 refueling outage in May 2002, the licensee discovered the
failure of four reactor coolant system pressure boundary charging line check valves.
The inspectors reviewed the circumstances relating to the common cause failure of
these valves documented in the root cause evaluation for the following CR:
CR 02134021, "Check Valves 1-CS-328-L1, 1-CS-328-L4, 1-CS-329-L1, and
1-CS-329-L4 Were Found Open During Radiographic Nonintrusive Testing."
The inspectors verified the following attributes during their review of the licensees
corrective actions for the above CR and several other related CRs:
consideration of the extent of condition, generic implications, common cause,
and previous opportunities to identify and correct the condition;
classification and prioritization of the resolution of the problem, commensurate
with safety significance;
identification of the root and contributing causes of the problem; and
identification of corrective actions which were appropriately focused to correct
the problem.
The inspectors also reviewed the corrective actions and associated CR evaluations with
applicable site personnel including the CR evaluators and system engineers.
b.
Findings
The inspectors identified a finding of very low safety significance (Green) associated
with this self-revealed event. The licensee failed to identify and take appropriate
corrective actions to preclude the failure of reactor coolant system pressure boundary
charging line check valves (Velan model B10-3114B-13M), which were at risk of
common cause failure due to industry identified design and manufacturing defects, a
significant condition adverse to quality. The inspectors determined that this issue
32
constituted a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," and
therefore dispositioned this finding as a Non-Cited Violation.
Description
On May 12, 2002, the licensee conducted radiographic nonintrusive testing on check
valve 1-CS-329-L1 in accordance with procedure 12-EP-4030-001-001, "Check Valve
Examination Surveillance." Results of the examination indicated that the valves disk
was stuck in the open position. Subsequent radiography of the remaining three identical
charging line check valves, 1-CS-329-L4, 1-CS-328-L1, and 1-CS-328-L4, identified that
all three were stuck in either the full or partially open position. The licensees
examination of the valves identified several as-found condition discrepancies including:
discs binding against the valve body due to oversize discs;
improper bushing positioning; and
problems with disc to arm clearances.
The licensee documented the condition and entered it into its corrective action program
as CR 02134021. In response to the CR, the licensee conducted a thorough root cause
analysis, which the inspectors concluded properly identified substantial industry
operating experience documenting similar failures. Further, the licensees evaluation
identified instances of D. C. Cook operating experience involving similar failures which
were not considered when evaluating operating experience. Specifically:
The Velan Valve Corporation published Service Bulletin 104 on October 10,
1990. The licensee received the bulletin on October 12, 1990 and entered it into
the its corrective action program as problem report 90-1503. The licensee took
corrective action including one-time vendor training on the identified valve
deficiencies and the return of on-site spare parts to the vendor for examination.
No positive actions, however, such as specific dimensional checks, were taken
to determine if the condition existed in the in-service components.
On February 10, 1992, the licensee received correspondence from
Westinghouse Electric Corporation informing them of a 10 CFR Part 21 Report
filed by Velan. The correspondence was entered into the licensees corrective
action program as problem report 92-157. The issue was closed with no action
taken based on the previous evaluation of problem report 90-1503.
On January 24, 1996, Operating Experience 7640 was received by the licensee
which documented Sequoias discovery of four Velan Model B10-3114B-13MS
3-inch charging injection check valves in the stuck open position. The OE was
entered into the licensees corrective action program as CR 96-0094. The
licensees evaluation concluded that although similar valves were installed at
D. C. Cook, similar problems had not been observed and therefore no corrective
action was required. Again, no specific dimensional checks of the installed
valves were performed.
The inspectors determined that the licensees failure to assure that corrective actions
were taken to preclude the failure of reactor coolant system pressure boundary charging
33
line check valves due to industry identified design and manufacturing defects was a
licensee performance deficiency warranting a significance evaluation. The inspectors
also concluded that this finding affected the cross-cutting issue of problem identification
and resolution. The inspectors noted that the majority of the instances of missed
opportunities to identify and correct potential valve deficiencies occurred in the early to
mid 1990s time frame.
Analysis
The inspectors assessed this finding using the SDP. The inspectors concluded that this
finding was associated with the barrier integrity cornerstone and adversely affected the
cornerstone objective, and as such it was more than a minor concern. Specifically, the
charging line check valves perform a safety-related function of limiting the release of
reactor coolant inventory should a charging line failure occur. The failure of the valves
in the open position would prohibit the performance of this function and therefore affects
the objective of the barrier integrity cornerstone. The inspectors performed a Phase 1
SDP review of this finding using the guidance provided in NRC Inspection Manual
Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings
for At-Power Situations." Because this finding involved the integrity of the reactor
coolant system pressure boundary, the inspectors determined that this finding required
a Phase 2 SDP analysis. After consulting with the Regional SRA, the inspectors
determined that this issue was of very low safety significance because no actual loss of
safety function occurred. The inspectors concluded that no actual loss of safety function
occurred based on the reported minimal force required to shut the valves (indicating that
they would have shut given the differential pressure applied during accident conditions)
and the redundancy provided by a third check valve (1-CS-321) in the charging line. In
accordance with IMC 0609, Appendix A, Attachment 1, Step 2.6, the SDP results were
not evaluated for potential risk contribution due to LERF because the accident sequence
result was less than 1E-7 per year.
Enforcement
10 CFR 50, Appendix B, Criterion XVI, "Corrective Action," states, in part, that measures
shall be established to assure that conditions adverse to quality, such as failures,
malfunctions, deficiencies, deviations, defective material and equipment, and
nonconformances are promptly identified and corrected. Contrary to the above, the
licensee failed to take corrective action to preclude the failure of reactor coolant system
pressure boundary charging line check valves (Velan model B10-3114B-13M), which
were at risk of common cause failure due to industry identified design and
manufacturing defects, a condition adverse to quality. Specifically, industry operating
experience was published in October 1990, February 1992, and January 1996 and
subsequently entered into the licensees corrective action program. However, the
licensee took no positive actions, such as specific dimensional checks, to determine if
the condition existed in the in-service components. Consequently, during the Unit 1
Cycle 18 refueling outage in May 2002, the licensee discovered the failure of four
reactor coolant system pressure boundary charging line check valves (1-CS-329-L1,
1-CS-329-L4, 1-CS-328-L1, and 1-CS-328-L4), which were the result of conditions
identified in the industry operating experience. Because of the very low safety
significance, this violation is being treated as a Non-Cited Violation consistent with
34
Section VI.A of the NRC Enforcement Policy (NCV 50-315-02-09-04(DRP)). The
licensee entered this violation into its corrective action program as CR 02134021.
4OA3 Event Follow-up (71153)
.1
(Closed) Unresolved Item (URI) 50-315-02-06-01(DRP): "Pressurizer Power Operated
Relief Valve (PORV) Inadvertently Opened During Testing Resulting in a Loss of
Reactor Coolant System Inventory and an Unusual Event."
a.
Inspection Scope
On June 5, 2002, with Unit 1 in Mode 4 (Hot Shutdown), pressurizer PORV 1-NRV-153
inadvertently opened while testing actuation logic circuitry for pressurizer PORV
1-NRV-151. Approximately 100 gallons of reactor coolant was released to the
pressurizer relief tank. The release rate exceeded the 25 gallons-per-minute (gpm) limit
established for declaring an Unusual Event in accordance with the licensees
Emergency Plan. The inspectors reviewed the circumstances associated with this
event, including the root cause determination, operator response during the event, and
corrective actions.
b.
Findings
A finding of very low safety significance (Green) was self-revealed. The licensee failed
to provide an appropriate procedure for testing the Unit 1 pressurizer PORVs, causing
an uncontrolled release of reactor coolant system inventory to the pressurizer relief tank.
This finding was dispositioned as a Non-Cited Violation of 10 CFR 50, Appendix B,
Criterion V, "Instructions, Procedures, and Drawings."
Discussion
As discussed in NRC Inspection Report 50-315/316-02-06(DRP), the inspectors
originally documented this finding as an Unresolved Item pending a final safety
significance determination. The inspectors referred this finding to the Regional SRA to
perform the additional analysis.
Analysis
The inspectors assessed this finding using the SDP. The inspectors concluded that this
issue could be reasonably viewed as a precursor to a significant event and was
therefore more than a minor concern. The inspectors also concluded that this finding
was associated with the initiating events cornerstone and adversely affected the
cornerstone objective. Specifically, the uncontrolled release of reactor coolant system
inventory upset plant stability and challenged the inventory control safety function.
Because Unit 1 was in a shutdown mode during this period, the inspectors performed a
Phase 1 SDP review of this issue using the guidance provided in NRC Inspection
Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination
Process." Based on the above information, the inspectors concluded that the most
appropriate Appendix G checklist to use for this issue was the checklist for "Pressurized
Water Reactor Hot Shutdown Operation - Time to core boiling less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />."
35
Because operator intervention was required to manually close the affected PORV block
valve, the inspectors concluded that the unit was in a configuration where a single active
failure or personnel error could have resulted in a rapid loss of reactor coolant system
inventory as described in Section II.B.(2) of the checklist. Consequently, the inspectors
concluded that this issue increased the likelihood of a loss of reactor coolant system
inventory and therefore required a Phase 2 SDP analysis. The inspectors discussed the
safety significance of this issue with the Regional SRA. The SRA reviewed the finding
and determined that the drain path could be easily isolated, accurate reactor coolant
system level indication was available, all steam generators were available for cooling,
and all trains of standby injection were available and not impacted by the finding. Based
on these factors the finding is characterized as having very low safety significance.
Enforcement
10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires,
in part, that activities affecting quality shall be prescribed by documented instructions,
procedures, or drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, or drawings. Contrary
to the above, the licensee failed to provide a procedure of a type appropriate to the
circumstances for testing the Unit 1 pressurizer PORVs, which is an activity affecting
quality. Specifically, the instructions contained in 1-IHP-4030-102-017, "Pressurizer
Power Operated Relief Valve (PORV) Actuation Channel Calibration with Valve
Operation (for Modes 1, 2, and 3)," Revision 1, failed to provide adequate control of
pressurizer PORVs 1-NRV-151 and 1-NRV-153, which have a common automatic
opening signal. This issue was self-revealed on June 5, 2002, when pressurizer PORV
1-NRV-153 inadvertently opened while testing actuation logic circuitry for 1-NRV-151,
causing an uncontrolled release of reactor coolant system inventory to the pressurizer
relief tank. Because of the very low safety significance, this violation is being treated as
a Non-Cited Violation consistent with Section VI.A of the NRC Enforcement Policy
(NCV 50-316-02-09-05(DRP)). The licensee entered this violation into its corrective
action program as CR 02157039.
.2
(Closed) LER 50-315-1999-010-01: "Reactor Coolant System Leak Detection System
Sensitivity Not in Accordance with TS Basis," Supplement 1. On May 3, 1999, the
licensee documented that the Unit 1 and Unit 2 lower containment sump level and flow
monitoring capabilities were not consistent with the recommendations of Regulatory
Guide 1.45 as stated in TS Basis 3/4.4.6.1. Specifically, the containment sump level
and flow monitoring systems were not able to detect a 1 gpm reactor coolant system
leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The licensees review of this issue determined that, other than the
TS Basis statement that the containment sump level and flow leak detection systems
are consistent with the recommendations of Regulatory Guide 1.45, the most restrictive
requirement was that the leak detection system be able to detect a 1 gpm leak within
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to meet Generic Letter 84-04 requirements. The licensee has changed the TS
Basis statement using the 10 CFR 50.59 process and documented this corrective action
in Supplement 1 to LER 50-315-1999-010-00. The inspectors concluded that this
change was appropriate. This event did not constitute a violation of NRC requirements.
This LER is closed.
36
.3
Unanticipated Start of the Unit 2 Turbine Driven Auxiliary Feedwater Pump (TDAFWP)
During a Normal Plant Shutdown for Refueling Outage
a.
Inspection Scope
On January 19, 2002, in preparation for a Unit 2 refueling outage, operators initiated a
planned manual reactor trip of Unit 2 from 22 percent power. Shortly thereafter, an
automatic start of the TDAFWP occurred due to a valid low level condition in two of the
four steam generators. The inspectors reviewed the circumstances associated with this
event, including the root cause determination, operator response leading to and during
the event, and corrective actions.
b.
Findings
(Closed) LER 50-316-2002-004-00: "Unanticipated Start of the Turbine Drive Auxiliary
Feedwater Pump."
(Closed) LER 50-316-2002-004-01: "Unanticipated Start of the Turbine Drive Auxiliary
Feedwater Pump," Supplement 1.
(Closed) LER 50-316-2002-004-02: "Unanticipated Start of the Turbine Drive Auxiliary
Feedwater Pump," Supplement 2.
The inspectors identified a finding of very low safety significance (Green) associated
with this self-revealed event. The licensee failed to provide instructions of a type
appropriate to the circumstances for a planned shutdown of Unit 2, which resulted in
unnecessarily challenging the automatic start function of Unit 2 TDAFWP. The
inspectors determined that this issue constituted a violation of 10 CFR 50, Appendix B,
Criterion V, "Instructions, Procedures, and Drawings," and therefore dispositioned this
finding as a Non-Cited Violation.
Discussion
On January 19, 2002, an automatic start of the Unit 2 TDAFWP occurred due to a valid
low level condition in two of the four steam generators following a planned manual
reactor trip from 22 percent power in preparation for a refueling outage. The licensee
had recently revised its plant shutdown procedure (02-OHP-4021-001-003, "Power
Reduction," Revision 15) to trip the reactor from less than 22 percent power in order to
enter the refueling outage more expeditiously. The licensee had previously performed
the manual reactor trip between 1 percent and 4 percent power. Water levels in each of
the four steam generators were within the program band prior to the reactor trip and
rapidly shrank to below the TDAFWP auto-start actuation setpoint. Following the
reactor trip and TDAFWP start, steam generator water levels rapidly recovered with
three AFW pumps supplying water from the condensate storage tank and cooled the
reactor coolant system below the system "no-load" temperature of 547 degrees
Fahrenheit. Pressurizer level also dropped to below 17 percent as a result of the
cooldown, which resulted in reactor coolant system letdown isolation. The automatic
start of the TDAFWP and reactor coolant system letdown isolation were both
37
unexpected occurrences for a normal plant shutdown and unnecessarily challenged the
operators.
The inspectors previously reviewed the licensees apparent cause evaluation for this
event. The NRC concluded in NRC Inspection Report 50-315/316-02-04(DRP) that the
licensees ability to consistently identify reasonable causes for conditions adverse to
quality in apparent cause evaluations performed for Category 3 CRs was inadequate
and documented a finding (FIN 50-315/316-02-04-03). The licensees apparent cause
evaluation for this event was one of the four examples included in that finding. The
inspectors concluded that the licensees apparent cause evaluation failed to adequately
address the cause for the unexpected TDAFWP start. The inspectors noted that the
evaluation was limited to the 10 CFR 50.73 reportability aspects of the unexpected
actuation of an engineered safety features component. The licensee subsequently
wrote CR 02107016 to evaluate the operational aspects of the unexpected automatic
pump start and to identify appropriate corrective actions.
The inspectors reviewed the licensees evaluation documented in CR 02107016 and
concurred with the licensees conclusion that a planned shutdown should not challenge
critical safety equipment to automatically start. The licensee submitted Supplement 1 to
LER 50-316-2002-004-00 to provide this conclusion and the corrective actions. The
licensee subsequently revised the plant shutdown procedure to initiate the reactor trip
from less than 17 percent power and has successfully performed the procedure on both
units without challenging the automatic start function of an TDAFWP. The licensee
submitted Supplement 2 to LER 50-316-2002-004-00 to identify the cause for the
engineered safety features component actuation and to clarify statements made in
earlier revisions of the LER. The inspectors concluded that 02-OHP-4021-001-003,
"Power Reduction," Revision 15, was not appropriate to the circumstances because
initiating the reactor trip from 22 percent power unnecessarily challenged the automatic
start function of Unit 2 TDAFWP.
Analysis
The inspectors assessed this finding using the SDP. The inspectors concluded that this
finding was associated with the initiating events cornerstone and adversely affected the
cornerstone objective and was therefore more than a minor concern. Specifically, the
function of critical safety equipment was challenged and plant stability was upset during
the performance of a normal plant shutdown by the automatic start of Unit 2 TDAFWP.
The inspectors performed a Phase 1 SDP review of this issue using the guidance
provided in NRC Inspection Manual Chapter 0609, Appendix A, "Significance
Determination of Reactor Inspection Findings for At-Power Situations." Because this
finding did not cause or contribute to the likelihood of an initiating event, the inspectors
concluded that this issue was of very low safety significance.
Enforcement
10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires,
in part, that activities affecting quality shall be prescribed by documented instructions,
procedures, or drawings, of a type appropriate to the circumstances and shall be
accomplished in accordance with these instructions, procedures, or drawings. Contrary
38
to the above, the licensee failed to provide instructions of a type appropriate to the
circumstances for conducting the Unit 2 plant shutdown on January 19, 2002, which is
an activity affecting quality. Specifically, the instructions contained in
02-OHP-4021-001-003, "Power Reduction," Revision 15, failed to ensure that the
automatic start function of Unit 2 TDAFWP would not be unnecessarily challenged
during a normal plant shutdown. Because of the very low safety significance, this
violation is being treated as a Non-Cited Violation consistent with Section VI.A of the
NRC Enforcement Policy (NCV 50-316-02-09-06(DRP)). The licensee entered this
violation into its corrective action program as CR 02019036 and CR 02107016.
.4
(Closed) LER 50-316-2002-005-00: "Unit 2 Trip Due to Instrument Rack 24-Volt DC
Power Supply Failure." The event described in this LER was discussed in Section 1R12
of this report. The inspectors concluded that the licensees failure to assure that prompt
corrective actions were taken to address age-related failures of reactor control
instrumentation power supplies to prevent repetition of power supply failures was a
finding of very low safety significance and a violation of 10 CFR 50, Appendix B,
Criterion XVI, "Corrective Action". The licensee reported this event as a condition that
resulted in an automatic actuation of the reactor protection system in accordance with
10 CFR 50.73(a)(2)(iv)(A). This LER is closed.
.5
(Closed) LER 50-316-1997-004-02: "Analysis Demonstrates Design Basis Impact of
Inadequate Refueling Outage Safety Evaluation Was Negligible," Supplement 2. The
licensee submitted Supplement 2 to LER 50-316-1997-004-00 to provide additional
information concerning the analysis of this event. The inspectors determined that the
information provided in Supplement 2 to LER 50-316-1997-004-00 did not raise any new
issues or change the conclusion of previous NRC reviews documented in Inspection
Reports 50-315/316-97-02(DRP), 50-315/316-98-09(DRS), 50-315/316-99-029(DRS),
and 50-315/316-00-01(DRP). This LER is closed.
.6
(Closed) LER 50-315-1997-005-00: "Reactor Coolant Pump Fire Protection Inoperable
for Extended Period Without Compensatory Actions Due to Improperly Fabricated
Gasket in Spray Header Line." The closure of Supplement 1 to this LER is discussed
below in Section 4OA3.7. This LER is closed.
.7
(Closed) LER 50-315-1997-005-01: "Reactor Coolant Pump Fire Protection Inoperable
for Extended Period Without Compensatory Actions Due to Improperly Fabricated
Gasket in Spray Header Line," Supplement 1. On March 5, 1997, it was discovered that
gaskets in the fire protection water spray system for the number 13 reactor coolant
pump had not been properly fabricated prior to installation during the 1995 refueling
outage. The gaskets for a spectacle flange were fabricated out of a sheet of red rubber,
without the center removed to provide a flow path for the water spray. The center area,
which should have been cut away during gasket fabrication, was found to be torn. The
torn gasket was attributed to the pressure of the system supervisory air. The licensee
reported this event in accordance with 10 CFR 50.73(a)(2)(i)(B) as an operation
prohibited by the plants TSs. Fire Protection requirements were subsequently removed
from the TSs in March of 1996. However, at the time the gaskets were installed, the
Fire Protection TSs were still in effect. During review of this event, the licensee
identified that personnel error was the root cause of the event. The personnel involved
did not properly incorporate the information contained in the job order activity regarding
39
fabrication of the gaskets into their actions. At the time of discovery, the licensee
corrected the problem and entered the issue into its corrective action program as
CR 97-0586. The failure to enter previous TS 3.7.9.2, Table 3.7-5B for an inoperable
reactor coolant pump sprinkler system whenever the reactor coolant pump was operable
was a violation TS 3.7.9.2. This finding constitutes a violation of minor significance that
is not subject to enforcement action in accordance with Section VI of the NRC
Enforcement Policy. This LER is closed.
.8
(Closed) LER 50-315-1998-056-01: "Inadequate Control and Processing of Design
Information Results in Unanalyzed Hot Leg Recirculation Switchover," Supplement 1.
On December 11, 1998, the licensee identified an unanalyzed condition related to the
post-loss-of-coolant accident emergency core cooling system hot leg switchover
sub-criticality analysis. The inspectors reviewed the original LER in NRC Inspection
Report 50-315/316-99-29(DRS) and concluded that this was a minor issue. The
licensee submitted Supplement 1 to LER 50-315-1998-056-00 to provide new
information concerning the analysis of the event and corrective actions. The cause of
this event was the licensees failure to adequately control design basis calculations and
supporting documentation, which is a violation of 10 CFR 50, Appendix B, Criteria III,
"Design Control." The licensee entered this event into its corrective action program as
CR 98-7848. This finding constitutes a violation of minor significance that is not subject
to enforcement action in accordance with Section VI of the NRC Enforcement Policy.
This LER is closed.
.9
(Closed) LER 50-315-1998-029-01: "Fuel Handling Area Ventilation System Inoperable
Due to Original Design Deficiency," Supplement 1. The inspectors reviewed the original
LER in NRC Inspection Report 50-315/316-99-29(DRS) and concluded that this was a
minor issue. The licensee submitted Supplement 1 to LER 50-315-1998-029-00 to
provide additional information concerning the analysis of the event, the cause, and the
corrective actions. The inspectors determined that the information provided in
Supplement 1 to LER 50-315-1998-029-00 did not raise any new issues or change the
conclusion of the initial review. This LER is closed.
.10
(Closed) LER 50-315-1999-003-00: "Control Room Pressurization System Surveillance
Test Does Not Test System in Normal Operating Condition." On January 7, 1999, the
licensee identified that the TS surveillance test procedure for testing the control room
pressurization function (12 EHP 4030 STP.229, "Control Room Emergency Ventilation
Test," Revision 3,) did not test the control room pressurization system in the normal
operating configuration. Specifically, one of the prerequisites prior to performing the
surveillance test was to verify that the pressure boundary door that separated the Unit 1
and Unit 2 control rooms (12DR-AUX415) was closed. However, it was identified that
the door was normally maintained open to facilitate access and egress between the two
control rooms with no procedural guidance to close the door during an event. It was
identified that a failure to recognize the door as part of both units control room pressure
boundary design resulted in the door being maintained open since initial plant start-up.
The inspectors concluded that this constitutes a violation of 10 CFR 50, Appendix B,
Criterion V, "Instructions, Procedures, and Drawings". The bases for TS Surveillance
Requirement 4.7.5.1.e.3 was to demonstrate control room operability such that radiation
exposure to personnel occupying the control room would be limited to 5 rem or less
whole body, or its equivalent. The licensee as part of its corrective actions closed and
40
labeled pressure boundary door 12DR-AUZ415. In addition, the licensee performed a
tracer gas test under several conditions, including one in which door 12DR-AUX415 was
left open between the Unit 1 and Unit 2 control rooms with one of the units control room
ventilation systems treated as inoperable. The test showed that the amount of unfiltered
in-leakage was not highly dependent on pressurization and the dose consequences of
having the door open during a postulated accident would remain within 10 CFR 50,
Appendix A, General Design Criteria 19 allowable limits. Therefore, this issue is
considered to be of minor safety significance. The licensee entered this event into its
corrective action program as CR 99-0275. This finding constitutes a violation of minor
significance that is not subject to enforcement action in accordance with Section VI of
the NRC Enforcement Policy. This LER is closed.
.11
(Closed) LER 50-315-1999-003-01: "Control Room Pressurization System Surveillance
Test Does Not Test System in Normal Operating Condition," Supplement 1. The
licensee submitted Supplement 1 to LER 50-315-1999-003-00 to provide additional
information concerning the analysis of the event, the cause, and the corrective actions.
The inspectors determined that the information provided in Supplement 1 to LER
50-315-1999-003-00 did not raise any new issues or change the conclusion of the initial
review which was documented above in Section 4OA3.10. This LER is closed.
.12
(Closed) LER 50-315-2000-004-00: "Circuit Design Could Result in Failure of
Emergency Diesel Generators to Load Properly After Loss of Offsite Power." On
July 19, 1999, an unanalyzed condition was identified by the Expanded System
Readiness Review Teams wherein a sneak electrical circuit existed that could cause
improper EDG load sequencing of equipment onto the vital buses following a loss-of-
coolant accident concurrent with a loss of offsite power. The licensee reported this
event as a condition that was outside the design basis in accordance with 10 CFR
50.73(a)(2)(ii). The licensee implemented a design change in both Unit 1 and Unit 2 to
eliminate the possibility of the sneak circuit. This event did not constitute a violation of
NRC requirements. This LER is closed.
4OA5 Other
.1
(Open) URI 50-316-02-09-07(DRP): "Review of NOED-02-3-058 Regarding D. C. Cook,
Unit 2, Compliance With Technical Specification 3.8.1.1." By letter dated
November 6, 2002, the licensee requested that the NRC exercise discretion not to
enforce compliance with the actions of TS 3.8.1.1 regarding operability of the Unit 2 CD
EDG. The inspectors opened URI 50-316-02-09-06 to track documentation of the root
cause for the Notice of Enforcement Discretion (NOED) request, review the NOED
approval basis, and verify licensee activities associated with NOED implementation.
.2
Completion of Appendix A to Temporary Instruction 2515/148, Revision 1
The inspectors completed the pre-inspection audit for interim compensatory measures
at nuclear power plants, dated September 13, 2002.
.3
(Closed) Inspector Follow-up Item (IFI) 50-315/316-99-29-01: "Review and Approval of
Dose Calculation for General Design Criteria 19 Control Room Habitability Issue." The
inspectors reviewed calculation RD-01-05, "Adjusted Dose Consequences for Changes
41
to Control Room," Revision 1. The calculation stated that the control room dose
consequences for all events would be below the acceptance criteria required by
10 CFR 50.67. No findings of significance were identified. This item is closed.
.4
(Closed) IFI 50-316-00-07-03: "Failure to Perform Post Modification Checks to Verify
Adequate Clearance Between the Pressurizer Surge Line Whip Restraints and the
Surge Line Under Hot Plant Conditions." The inspectors performed a limited review of
calculation SD-990825-001, "HELB [High Energy Line Break]: Structural Evaluation of
Surge Line Pipe Whip Restraints," Revision 3. This calculation included determination
of necessary clearances for the pressurizer surge line whip restraints. Additional design
engineering documents were reviewed to verify that the calculation results were properly
incorporated into the plant design. No findings of significance were identified. This item
is closed.
.5
(Closed) URI 50-315/316-00-16-04: "Determine Whether the Latent Failure of a Test
Relay Should Be Treated Under the Category of a Single Failure." The NRC staff
reviewed this issue and determined that the failure of a K-800 relay would not prohibit
the proper operation of an engineered safety features actuation circuit in response to a
valid actuation signal. No findings of significance were identified. This item is closed.
.6
(Closed) URI 50-315/316-01-15-01: "A Change Was Made to the UFSAR Without a
10 CFR 50.59 Evaluation." The licensee changed the UFSAR and inappropriately used
10 CFR 50.71 (e) rather than 10 CFR 50.59 to evaluate the UFSAR change. The
inspectors reviewed CR 01291058 which described this issue. The licensee corrected
this problem by performing a 10 CFR 50.59 screening and concluded that the change
did not require NRC approval prior to implementation. No findings of significance were
identified. This item is closed.
4OA6 Meetings
.1
Interim Exits
The results of the Emergency Preparedness Program Inspection were presented to
Mr. J. Molden and other members of licensee management at the conclusion of the
inspection on December 6, 2002. The licensee acknowledged the findings presented.
The inspector asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
The results of the Radiological Protection Instrumentation and Access Control
Inspection were presented to Mr. J. Molden and other members of licensee
management at the conclusion of the inspection on December 6, 2002. The licensee
acknowledged the findings presented. The inspector asked the licensee whether any
materials examined during the inspection should be considered proprietary. No
proprietary information was identified.
.2
Resident Inspectors Exit
The inspectors presented the inspection results to Mr. J. Pollock and other members of
licensee management at the conclusion of the inspection on January 3, 2003. The
42
licensee acknowledged the findings presented. The inspectors asked the licensee
whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identified.
43
KEY POINTS OF CONTACT
Licensee
J. Gebbie, Plant Engineering Assistant Director
G. Gibson, Site Protective Services
C. Graffenius, Emergency Planner
S. Greenlee, Nuclear Technical Services Director
G. Harland, Work Control/Maintenance Director
R. Hershberger, Chemistry Supervisor
R. LaBurn, Radiation Protection General Supervisor
E. Larson, Operations Director
B. McIntyre, Regulatory Assurance Manager
R. Meister, Regulatory Affairs Specialist
J. Molden, Acting Plant Manager
D. Moul, Operation Work Control Manager
T. Noonan, Performance Assurance Director
S. Partin, Emergency Planning Manager
J. Pollock, Site Vice President
B. Robinson, Radiation Protection Superintendent
M. Scarpello, Regulatory Compliance Supervisor
S. Simpson, Operations Staff Manager
D. Wood, Radiation / Environmental Manager
44
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-316-02-09-01
Failure to assure that prompt corrective actions were taken to
address age-related failures of reactor control instrumentation
power supplies to prevent repetition of power supply failures
(Section 1R12)
50-316-02-09-02
Failure to implement a corrective action to prevent recurrence
associated with reactor control instrumentation power supply
failures (Section 1R12)
50-315/316-02-09-03
Failure to assure that corrective actions were taken to
preclude repetition of EDG starting air system relay failures
(Section 4OA2.1)
50-315-02-09-04
Failure to identify and take appropriate corrective actions to
preclude the failure of reactor coolant system pressure
boundary charging line check valves which were at risk of
common cause failure due to industry identified design and
manufacturing defects (Section 4OA2.2)
50-315-02-09-05
Failure to provide an appropriate procedure for testing the
Unit 1 pressurizer power operated relief valves causing an
uncontrolled release of reactor coolant system inventory to
the pressurizer relief tank (Section 4OA3.1)
50-316-02-09-06
Failure to provide appropriate instructions for a planned
shutdown of Unit 2 which resulted in unnecessarily
challenging the automatic start function of Unit 2 turbine
auxiliary feedwater pump (Section 4OA3.3)
50-316-02-09-07
Review of NOED-02-3-058 regarding D. C. Cook, Unit 2,
compliance with Technical Specification 3.8.1.1
(Section 4OA5.1)
Closed
50-316-02-09-01
Failure to assure that prompt corrective actions were taken to
address age-related failures of reactor control instrumentation
power supplies to prevent repetition of power supply failures
(Section 1R12)
50-316-02-09-02
Failure to implement a corrective action to prevent recurrence
associated with reactor control instrumentation power supply
failures (Section 1R12)
45
50-315/316-02-09-03
Failure to assure that corrective actions were taken to
preclude repetition of EDG starting air system relay failures
(Section 4OA2.1)
50-315-02-09-04
Failure to identify and take appropriate corrective actions to
preclude the failure of reactor coolant system pressure
boundary charging line check valves which were at risk of
common cause failure due to industry identified design and
manufacturing defects (Section 4OA2.2)
50-315-02-06-01
Pressurizer power operated relief valve inadvertently opened
during testing resulting in a loss of reactor coolant system
inventory and an Unusual Event (Section 4OA3.1)
50-315-02-09-05
Failure to provide an appropriate procedure for testing the
Unit 1 pressurizer power operated relief valves causing an
uncontrolled release of reactor coolant system inventory to
the pressurizer relief tank (Section 4OA3.1)
50-315-1999-010-01
LER
Reactor coolant system leak detection system sensitivity not
in accordance with TS [Technical Specification] Basis
(Section 4OA3.2)
50-316-02-09-06
Failure to provide appropriate instructions for a planned
shutdown of Unit 2 which resulted in unnecessarily
challenging the automatic start function of Unit 2 turbine
auxiliary feedwater pump (Section 4OA3.3)
50-316-2002-04-00
LER
Unanticipated start of the turbine drive auxiliary feedwater
pump (Section 4OA3.3)
50-316-2002-04-01
LER
Unanticipated start of the turbine drive auxiliary feedwater
pump (Section 4OA3.3)
50-316-2002-04-02
LER
Unanticipated start of the turbine drive auxiliary feedwater
pump (Section 4OA3.3)
50-316-2002-05-00
LER
Unit 2 trip due to instrument rack 24-volt DC [direct current]
power supply failure (Section 4OA3.4)
50-316-1997-004-02
LER
Analysis demonstrates design basis impact of inadequate
refueling outage safety evaluation was negligible
(Section 4OA3.5)
50-315-1997-005-00
LER
Reactor coolant pump fire protection inoperable for extended
period without compensatory actions due to improperly
fabricated gasket in spray header line (Section 4OA3.6)
50-315-1997-005-01
LER
Reactor coolant pump fire protection inoperable for extended
period without compensatory actions due to improperly
fabricated gasket in spray header line (Section 4OA3.7)
46
50-315-1998-056-01
LER
Inadequate control and processing of design information
results in unanalyzed hot leg recirculation switchover
(Section 4OA3.8)
50-315-1998-029-01
LER
Fuel handling area ventilation system inoperable due to
original design deficiency (Section 4OA3.9)
50-315-1999-003-00
LER
Control room pressurization system surveillance test does not
test system in normal operating condition (Section 4OA3.10)
50-315-1999-003-01
LER
Control room pressurization system surveillance test does not
test system in normal operating condition (Section 4OA3.11)
50-315-2000-004-00
LER
Circuit design could result in failure of emergency diesel
generators to load properly after loss of offsite power
(Section 4OA3.12)
50-315/316-99-29-01
IFI
Review and approval of dose calculation for General Design
Criteria 19 control room habitability issue (Section 4OA5.3)
50-316-00-07-03
IFI
Failure to Perform post modification checks to verify
adequate clearance between the pressurizer surge line whip
restraints and the surge line under hot plant conditions
(Section 4OA5.4)
50-315/316-00-16-04
Determine whether the latent failure of a test relay should be
treated under the category of a single failure
(Section 4OA5.5)
50-315/316-01-15-01
A change was made to the UFSAR without a 10 CFR 50.59
evaluation (Section 4OA5.6)
Discussed
50-315/316-02-04-03
Green finding regarding the failure to consistently identify a
reasonable apparent cause for conditions adverse to quality
(Section 4OA3.1)
47
LIST OF ACRONYMS USED
Agency-wide Documents and Management System
As-Low-As-Reasonably-Achievable
Alert and Notification System
Core Damage Frequency
CFR
Code of Federal Regulations
Carbon Dioxide
CR
Condition Report
CY
Calender Year
Direct Current
Drill and Exercise Performance
Division of Reactor Projects
Division of Reactor Safety
EHP
Electrical Maintenance Head Procedure
Emergency Response Organization
Essential Service Water
Early Warning System
Finding
IFI
Inspector Follow-up Item
IHP
Instrument Maintenance Head Procedure
IMC
Inspection Manual Chapter
LER
Licensee Event Report
Larger Early Release Frequency
MHP
Maintenance Head Procedure
Non-Cited Violation
NEI
Nuclear Energy Institute
Notice of Enforcement Discretion
NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
OA
Other Activities
OHP
Operations Head Procedure
Operator Workaround
Publically Available Records
Power Conversion System
Performance Indicator
PMI
Plant Managers Instruction
Plant Managers Procedure
Power Operated Relief Valve
Pressurized Water Reactor
Radiation Protection Technician
Radiation Work Permit
Significance Determination Process
48
SPP
Special Plant Procedure
Senior Reactor Analyst
Structures, Systems, and Components
Surveillance Test Procedure
TDAFWP
Turbine Driven Auxiliary Feedwater Pump
TS
Technical Specification
Updated Final Safety Analysis Report
Unresolved Item
49
LIST OF DOCUMENTS REVIEWED
The following is a list of licensee documents reviewed during the inspection, including
documents prepared by others for the licensee. Inclusion on this list does not imply the NRC
inspectors reviewed the documents in their entirety, but rather, that selected sections or
portions of the documents were evaluated as part of the overall inspection effort. Inclusion of a
document in this list does not imply NRC acceptance of the document, unless specifically stated
in the inspection report.
1R01
Adverse Weather Protection
PMP 5055-001-001
Winterization/Summerization
Revision 0
12-OHP 4022.001.010
Severe Weather
Revision 1
12-IHP 4022.001.009
Plant Winterization and De-Winterization
Revision 0
CR P-96-00229
Coils Were Found Frozen and Ruptured
on Various Air Handling Units
February 14, 1996
CR P-98-06318
There is No Freeze Protection for the
Condensate Storage Tanks
October 29, 1998
CR P-99-01516
Plant Does Not Have adequate
Winterization Policies
January 26, 1999
CR P-99-16338
Winter Storm Damage to Intake
Structures
June 10, 1999
CR P-00-01638
Unit 1 and 2 Screenhouse Water Level
Sensing Lines Freezing
January 28, 2000
CR 02087029
Reviewed Winterization Program Per
Condition Report 01318056, Action 1,
and Determined That Two Procedure
Enhancements Are Necessary to Make
the Program Successful in the Future
March 28, 2002
1R04
Equipment Alignment
1R04.1
Partial System Walkdowns
Unit 2 Turbine Driven and West Auxiliary Feedwater (AFW) System Trains
02-OHP-4021-056-001
Filling and Venting Auxiliary Feedwater
System
Revision 15
12-PMP-4030.001.001
Impact of Safety Related Ventilation in
the Operability of Technical Specification
Equipment
Revision 5
50
02-OHP-4030-STP-017W
West Motor Driven Auxiliary Feedwater
System Test
Revision 11
02-OHP-4030-STP-017T
Turbine Driven Auxiliary Feedwater
System Test
Revision 15
OP-2-5106A
Flow Diagrams - Auxiliary Feedwater
Revision 45
DB-12-AFWS
Design Basis Document - Auxiliary
Feedwater System
Revision 0
D. C. Cook Nuclear Plant Updated Final
Safety Analysis Report (UFSAR), Section
10.5.2, "Auxiliary Feedwater System"
Revision 17
CR 02296004 (1)
2-FW-244-2 West Motor Driven AFW
Pump Suction Strainer OME-32W South
Basket Vent Valve Leaks at a Rate of
Less than One Drop per Minute
October 23, 2002
CR 02296006 (1)
2-FW-244-1 West Motor Driven AFW
Pump Suction Strainer OME-32W North
Basket Vent Valve Leaks at a Rate of
Less than One Drop per Minute
October 23, 2002
CR 02298053 (1)
2-HV-AFP-FD-4B, Fire Damper for the
U2 West Motor Driven Aux Feed Pump
Room, Was Found to Have a Small
Plastic Security Sign Laying in the Fire
Damper Track
October 25, 2002
Unit 2 AB Emergency Diesel Generator (EDG)
02-OHP-4021-032-008AB
Operating DG2AB Subsystems
Revision 2
OP-2-5251B-59
Flow Diagram Emergency Diesel
Generator AB Unit 2
Revision 59
OP-2-5151A-51
Flow Diagram Emergency Diesel
Generator AB Unit 2
Revision 51
CR 02308032 (1)
2-DG-136A (2AB EDG Starting Air
Receiver Number 2 to Flywheel Air Jack
Connection Shutoff Valve) Was Found
Open During a Procedure Walkdown
Being Performed By an NRC Inspector
November 4, 2002
51
Unit 1 West Essential Service Water (ESW) System Train
12-OHP-4021-019-001
Operation of the Essential Service Water
System
Revision 25
01-OHP-4030-066-4025
Unit 1 Appendix R and Ventilation
Requirements for Unit 2
Revision 3
OP-1-5113-74
Flow Diagram Essential Service Water
Revision 74
Miscellaneous Condition Reports
CR 022700009 (1)
Tell Tale Drains From Unit 1 Steam
Generator 1 and 4 Safety Valves Have
Signs of Leakage
September 26, 2002
1R05
Fire Protection
1R05.1
Routine Resident Inspector Tours
D. C. Cook Nuclear Plant UFSAR,
Section 9.8.1, "Fire Protection System"
Revision 17
D. C. Cook Nuclear Plant Fire Hazards
Analysis, Units 1 and 2
Revision 8
D. C. Cook Nuclear Plant Units 1 and 2
Probabilistic Risk Assessment, Fire
Analysis Notebook
February 1995
D. C. Cook Nuclear Plant Administrative
Technical Requirements Manual,
Sections 1-FP-7 and 2-FP-7, "Fire Rated
Assemblies"
PMP 2270.CCM.001
Control of Combustible Materials
Revision 1
PMP 2270.FIRE.002
Responsibilities for Cook Plant Fire
Protection Program Document Updates
Revision 0
PMP 2270.WBG.001
Welding, Burning and Grinding Activities
Revision 0
PMP 5020.RTM.001
Restraint of Transient Material
Revision 1
PMI 2270
Fire Protection
Revision 26
12-PPP-2270-066-001
Portable Fire Extinguisher Inspections
Revision 0a
Job Order R0216014
18 Month Surveillance of Fire Dampers in
Accordance with 12-PPP-4030-066-021
June 9, 2002
52
12-PPP-4030-066-021
Inspection of Fire Dampers Protecting
Safety-Related Areas
Revision 1
Job Order R0234423
Perform 6 Month Surveillance of Fire
Detection Circuits in Accordance with
12-IHP-4030-STP-206
November 1, 2002
12-IHP-4030-STP-206
Fire Detection Instrumentation Channel
Functional Test
Revision 3
02-IHP-4030-266-052
Unit 2 Control Rod Drive, Transformer,
Switchgear Room Carbon Dioxide Fire
Suppression Test
Revision 1
Miscellaneous Condition Reports
CR 02305079 (1)
Misinterpretation of the Administrative
Technical Requirement Surveillance
Requirements for Fire Damper Closure
Testing
November 1, 2002
CR 02317181 (1)
Tracking CR to Add Fire Pump House to
Fire Hazard Analysis Fire Zone
Designations
November 13, 2002
1R11
Licensed Operator Requalification
1R11.1
Resident Inspector Quarterly Review
Licensed Operator Requalification
Training Simulator Evaluation Scenario
for October 29, 2002
1R12
Maintenance Effectiveness
PMP-5035-MRP-001
Administration
Revision 4
PMI-5035
Revision 9
53
Control Group Power Supply Failures
Maintenance Rule (a)(1) Action Plan
Reactor Control 7 Instrumentation
System - Control Group Power Supplies
Revision 0
Maintenance Rule Scoping Document
Reactor Control System
Revision 1
LER 316-2002-005-00
Unit 2 Trip Due to Instrument Rack
24-Volt DC (Direct Current) Power Supply
Failure
July 10, 2002
Inadequate Maintenance of
Uninterruptible Power Supplies and
Inverters
March 24, 1994
95-10, Supplement 2
Potential for Loss of Automatic
Engineered Safety Features Actuation
August 11, 1995
D. C. Cook Unit 2 Tripped From Full
Power Due to an Instrument Rack Power
Supply Failure
May 12, 2002
38915, Revised
D. C. Cook Unit 2 Tripped From Full
Power Due to an Instrument Rack Power
Supply Failure
May 14, 2002
PMP 4010.TRP.001
Unit Two Reactor Trip Review Report
May 12, 2002
Unit 2 Control Room Logs
May 12, 2002
through
May 13, 2002
CR 01236037
There Have Been a Significant Number
of Electronic DC Power Supply Failures
During the Past 24 Months
August 24, 2001
CR 02047020
2-CG-2-19 Power Supply PS2 Power
Available Lamp Is Off
February 16, 2002
CR 02133001
Both 24-Volt DC Power Supplies in
Control Group 1 for Rack 16 Failed
May 12, 2002
CR 02133002
Unit 2 Trip From 100 Percent Power
Level Due to Low Feedwater Flow
Coincident With Low Steam Generator
Level on Loop 1
May 12, 2002
54
CR 02133035
After Unit 2 Trip, No Auto-makeup to
Volume Control Tank, 2-QRV-303 Went
to Full Divert, and No Refueling Water
Sequence Due to 2-QLC-451
May 12, 2002
CR 02133058
The Procedure for Volume Control Tank
Instrument Malfunction Directs the
Bistable to Be Tripped Which Does Not
Result in a Conservative Condition
May 12, 2002
CR 02134014 (1)
TS 3.0.3 Was Entered Erroneously
Following the Unit 2 Reactor Trip on May
12, 2002
May 13, 2002
CR 02137004
This CR Written at Plant Operations
Review Committee Chairmans Request
to Drive a Design Engineering Evaluation
of the Instrument Control Power Electrical
Distribution Based on the Equipment
Operability During the Unit 2 Reactor Trip
on May 12, 2002
May 17, 2002
CR 02138002
24-Volt DC Power Supplies in Control
Group 1 Are Only Production 4 Volts DC
May 18, 2002
CR 02139034
80-Volt DC Power Supply in Control
Group 3 1-PS-CGC-20 PS-1 Is Dead
May 19, 2002
CR 02142030
Unit 2 Tripped Due to Loss of Control
Room Control Group 1 Power Supplies
May22, 2002
CR 02325058 (1)
Weekly Recurring Tasks to Walkdown
Taylor Mod 30 Power Supplies - No
Documented Performance of Walkdown
Since September 30, 2002
November 21, 2002
CR 02326025
PS2 Available Lamp Not Lit
November 22, 2002
1R13
Maintenance Risk Assessments and Emergent Work Evaluation
PMP-2291-OLR-001
On-Line Risk Management
Revision 2
PMP-2291-OLR-001
On-Line Risk Management
Revision 3
Industry Guideline for Monitoring the
Effectiveness of Maintenance at Nuclear
Power Plants, Section 11, "Assessment
of Risk Resulting From Performance of
Maintenance Activities"
Revision 2
55
PMP-2291-OLR-001
Data Sheet 1
On-Line Risk Management Work
Schedule Review and Approval Form
Cycle 43, Week 3, with Revisions
October 6, 2002
through
October 12, 2002
Unit 2 Control Room Logs
October 8, 2002
through
October 10, 2002
Unit 2 Supervisors Turnover Logs
October 8, 2002
through
October 10, 2002
Unit 2 Abnormal Position Log
October 8, 2002
through
October 10, 2002
Online Integrated Work Schedule
October 8, 2002
through
October 10, 2002
Unit 1 AB EDG ESW Supply Valves
Daily Shift Managers Logs
December 6, 2002
PMP-2291-OLR-001,
Data Sheet 1
On-Line Risk Management Work
Schedule Review and Approval Form
Cycle 43, Week 11
December 1, 2002
through
December 7, 2002
Unit 2 CD EDG
Daily Shift Managers Logs
November 4, 2002
through
November 5, 2002
PMP-2291-OLR-001
Data Sheet 1
On-Line Risk Management Work
Schedule Review and Approval Form
Cycle 43, Week 7
November 3, 2002
through
November 9, 2002
Unit 1 East ESW Pump
Daily Shift Managers Logs
December 15, 2002
through
December 19, 2002
PMP-2291-OLR-001
Data Sheet 1
On-Line Risk Management Work
Schedule Review and Approval Form
Cycle 44, Week 1
December 15, 2002
through
December 21, 2002
56
NRC Letter to A. Christopher Bakken III,
Subject: "Donald C. Cook Nuclear Plant,
Units 1 and 2 - Issuance of Amendments
(TAC NOS. MB5729 and MB5730)"
September 9, 2002
Unit 1 West ESW Pump
Daily Shift Managers Logs
October 27, 2002
through
October 31, 2002
PMP-2291-OLR-001
Data Sheet 1
On-Line Risk Management Work
Schedule Review and Approval Form
Cycle 43, Week 6
October 27, 2002
through
November 2, 2002
Unit 2 East ESW Pump
Daily Shift Managers Logs
November 17, 2002
through
November 23, 2002
PMP-2291-OLR-001
Data Sheet 1
On-Line Risk Management Work
Schedule Review and Approval Form
Cycle 43, Week 9
November 17, 2002
through
November 23, 2002
NRC Letter to A. Christopher Bakken III,
Subject: "Donald C. Cook Nuclear Plant,
Units 1 and 2 - Issuance of Amendments
(TAC NOS. MB5729 and MB5730)"
September 9, 2002
D. C. Cook Nuclear Plant UFSAR,
Section 9.8.3, "Service Water Systems"
Revision 17
CR 02324116 (1)
Training for the Sky Jack Fork Truck Was
Observed Being Performed in Close
Proximity of the Unit 1 Station
Transformer
November 20, 2002
Unit 2 East AFW System Train
PMP-2291-OLR-001
Data Sheet 1
On-Line Risk Management Work
Schedule Review and Approval Form
Cycle 43, Week 5, with Revisions
October 20, 2002
through
October 26, 2002
Unit 2 Control Room Logs
October 20, 2002
through
October 26, 2002
57
Unit 2 Supervisors Turnover Logs
October 20, 2002
through
October 26, 2002
Unit 2 Abnormal Position Log
October 20, 2002
through
October 26, 2002
Online Integrated Work Schedule
October 20, 2002
through
October 26, 2002
Miscellaneous Condition Reports
CR 02344008
1-HE-8 Turbine Auxiliary Cooling Water
Heat Exchanger Became Air Bound on
the Circulating Water Side of the Heat
Exchanger Resulting in a Secondary
Transient in Unit 1
December 10, 2002
CR 02346017
There Have Been Repeat Instances of
Inconsistencies in Regards to Application
of Cascading TS
December 12, 2002
CR 02350015
Unit 2 Containment Hydrogen
Recombiner Number 2 Maintenance Was
Scheduled During the 1E ESW Pump
Replacement
December 16, 2002
1R14
Personnel Performance During Non-routine Plant Evolutions
1R14.1
Unit 2 Power Reduction to Support Oil Addition to a Reactor Coolant Pump
Motor
02-OHP-4021-001-003
Power Reduction
Revision 15
Daily Shift Managers Logs
November 10, 2002
through
November 11, 2002
CR 02315078 (1)
Rising Pressure Indications on
1-IPI-260/-265, Safety Injection Pump
Discharge Pressures
November 11, 2002
1R14.2
Unit 1 Control Group Power Supply Replacement
PMI-4090
Infrequently Performed Test or Evolution
Briefing Guide for Replacement of PS2
Power Supply at 1-PS-CGC-16
December 3, 2002
58
D. C. Cook Nuclear Plant Unit 1
Technical Specifications
Daily Shift Managers Logs
December 3, 2002
1R15
Operability Evaluations
D. C. Cook Nuclear Plant Unit 1 and 2
Technical Specifications
D. C. Cook Nuclear Plant UFSAR
Revision 17
Information to Licensees Regarding NRC
Inspection Manual Section on Resolution
of Degraded and Nonconforming
Conditions
Revision 1
PMP-7030-ORP-001
Revision 9
R219215-02
Ultrasonic Test Report - Unit 2 Essential
Service Water Line to West Motor Driven
Auxiliary Feedwater Pump
December 4, 2002
CR 01101066
Flush Unit 2 ESW Line to West Motor
Driven AFW Pump due to Silt/Sand in
Piping
April 11, 2001
CR 02136008
While Performing Inspection of Termi-
point Connection, a Wire Found
Disconnected in 1-RPS-A
May 15, 2002
CR 02131018
Review Operability and Reportability
Issues for Two Items Dealing with
Feedwater Pressure Indication and the
Plant Process Computer Calorimetric
Program
May 11, 2002
CR 02135049
1-CCR-462 Leaking Excessively During
May 15, 2002
CR 02290012
Steam Generator PORV Actuator
Capability Calculation Revealed Negative
Calculated Margin for Full Stroke
Capability
October 17, 2002
CR 02300002
Unit 2 Control Room Access Door 2-DR-
AUX411B Latch Has Broken and Door
Will Not Shut
October 27, 2002
59
CR 02339016
Ultrasonic Examination on Unit 1 ESW to
West Motor Driven AFW Pump Piping
Found Some Silt/Sand in the Piping
December 5, 2002
1R16
Operator Workarounds
PMP-4010-OWA-001
Oversight and Control of Operator
Workarounds
Revision 1
NRC Inspection Manual
Temporary Instruction
2515/138
Evaluation of the Cumulative Effect of
Operator Workarounds
Work Around Review Board Meeting
Agenda
October 23, 2002
Work Around Review Board Meeting
Minutes
October 23, 2002
Unit 1 Operator Workarounds
October 23, 2002
Unit 2 Operator Workarounds
October 23, 2002
Unit 1 Operator Workarounds
Contingency Actions for Reactor
Operators
October 23, 2002
Unit 1 Operator Workarounds
Compensatory Actions for Auxiliary
Equipment Operators
October 23, 2002
Unit 2 Operator Workarounds
Contingency Actions for Reactor
Operators
October 23, 2002
Unit 2 Operator Workarounds
Compensatory Actions for Auxiliary
Equipment Operators
October 23, 2002
CR 01048019
Unit 1 Main Turbine Was Deliberately
Slowed Due to High Vibration Using the
Vacuum Breakers
February 17, 2002
CR 01280031
Bypass Steam to Feedwater Heater
Valves Leak and Must be Manually
Isolated
October 7, 2001
CR 02023083
Feedwater Preheating Valves Will Not
Isolate Sufficiently to Prevent Cooldown
January 23, 2002
60
CR 02312059 (1)
Update Operations Lesson Plan RQ-C-
KNOW. Lesson Plan Contains Incorrect
Information
November 8, 2002
1R19
Post Maintenance Testing
Unit 2 CD EDG Governor Replacement
02-OHP-4030-STP-
027CD
CD Diesel Generator Operability Test
(Train A)
Revision 20
PMP-2291-PNT-001
Post Maintenance Testing
Revision 3
12-MHP-5021-032-017
Attachment 1, Guideline To Run Diesel
At Slow Speed
Revision 4
CR 02306005
CD EDG Exhibited 150 Kilowatt
Oscillations at Full Load During
Surveillance Testing
November 2, 2002
Unit 2 East AFW Pump Maintenance
Job Order 02228012-01
2-PP-3E-MTR - Unit 2 East Motor Driven
AFW Pump Motor - Check For Soft Foot
and Perform Alignment
August 16, 2002
Job Order 02136108-01
2-FRV-255 Adjust Packing and Perform
Diagnostic Testing
May 16, 2002
02-OHP-4030-STP-017E
East Motor Driven Auxiliary Feedwater
System Test
Revision 10
02-OHP-4021-056-002,
Attachment 2
Auxiliary Feed Pump Operation - East
Motor Driven Auxiliary Feedwater Pump
Long-Term Minimum Flow
Revision 13
02-OHP-4021-056-002,
Attachment 9
Auxiliary Feed Pump Operation - East
Motor Driven Auxiliary Feedwater Pump
Operation
Revision 13
Unit 2 East Motor Driven AFW Pump
Historical Vibration Data
January 2001
through
October 2002
OHI-4030
Removal and Restoration of Technical
Specification Related Equipment - Unit 2
East Motor Driven AFW Pump
October 23, 2002
61
CR 02296051 (1)
The Motor Mounting Bolts on the Unit 2
East Motor Driven AFW Pump Were Not
Tightened per the Torque Selection
Procedure 12-MHP-5021-001-009
October 23, 2002
Unit 2 West ESW Pump Maintenance
Job Order 02269031-01
2-PP-7W - Uncouple-Inspect Coupling
Gap on Pump
September 26, 2002
Job Order 02269031-02
2-PP-7W Run Pump/Perform Operability
Test
September 26, 2002
02-OHP-4030-219-022W
West Essential Service Water System
Test
Revision @
Unit 1 Post Accident Containment Hydrogen Monitor
12-EHP-4030-STP-236-
010
Leak Test of Unit 1 and Unit 2 Post
Accident Containment Hydrogen
Monitoring System
Revision 1
OP-1-5141D-19
Flow Diagram Post-Accident Sampling
Containment Hydrogen Unit No. 1
Revision 19
OP-2-5141D-14
Flow Diagram Post-Accident Sampling
Containment Hydrogen Unit No. 2
Revision 14
1R22
Surveillance Testing
D. C. Cook Nuclear Plant UFSAR
Revision 17
D. C. Cook Nuclear Plant Unit 1 and 2
Technical Specifications
Unit 1 Auxiliary Cable Vault CO2 [Carbon Dioxide] Fire Suppression Test
01-EHP-4030-ATR-225-
020
Unit 1 Auxiliary Cable Vault CO2 Fire
Suppression Test
Revision 0
Administrative Technical
Requirements
Section 1-FP-5, Low Pressure CO2
Systems
Revision 25
CR 02345019
Procedure Needs Enhancement to Avoid
Unwanted CO2 Discharge
December 11, 2002
CR 02345020
Procedure Needs Fixes
December 11, 2002
62
Unit 2 Distributed Ignition System Surveillance and Baseline Testing
02-IHP-4030-234-001
Unit 2 Distributed Ignition System
Surveillance and Baseline Testing
Revision 0
D. C. Cook Nuclear Plant UFSAR,
Section 14.3.6.6, "Distributed Ignition
System"
Revision 17
Job Order R232934-01
Perform Quarterly Distributed Ignition
System Surveillance Unit 2
Steam Generator Steam/Feed Flow Mismatch and Steam Pressure Protection Functional
Testing
02-IHP-4030-SMP-219
Steam Generator 1&2 Steam/Feed Flow
Mismatch and Steam Pressure Protection
Set I Functional Test and Calibration
Revision 6
02-IHP-4030-SMP-222
Steam Generator 2&4 Steam/Feed Flow
Mismatch and Steam Pressure Protection
Set II Functional Test and Calibration
Revision 4
Job Order R0235115-01
Perform 2IHP-4030-SMP-219
PMI-4030 Performance
Review and Acceptance
Sheet
Performance Review and Acceptance
Sheet for Job Order R0235115-01
Job Order R0235114-01
Perform 02-IHP-4030-SMP-222
OP-2-99012
Steam Generator 1 & 2 Mismatch
Channel 1 Functional Diagram
Revision 1
D. C. Cook Nuclear Plant UFSAR,
Chapter 7, "Instrumentation and Control"
Revision 17
Containment Isolation and In-service Inspection Valve Operability Testing
01-OHP-4030-STP-011
Containment Isolation and In-service
Inspection Valve Operability Test
Revision 23
Steam Pressure Protection Functional Testing
02-IHP-4030-SMP-227
02-IHP-4030-SMP-227 Steam Pressure
Protection Set III Functional Test and
Calibration
Revision 2
63
02-IHP-4030-SMP-228
02-IHP-4030-SMP-228 Steam Pressure
Protection Set IV Functional Test and
Calibration
Revision 2
Unit 2 East AFW Pump Characterization Testing
12-EHP-5030-CAR-001
Characterization Testing Program
Revision 0
Motor Analysis Report,
Sequence 24
Motor-Driven AFW Pump, 1-PP-3W-MTR
December 4, 2002
Unit 1 and 2 Personnel Airlock Door Seal Leak Rate Surveillance Testing
12-IHP-4030-046-227
Unit 1 and Unit 2 Personnel Airlock Door
Seal Leak Rate Surveillance
Revision 0
Nuclear Energy Institute
(NEI) 94-01
Industry Guideline for Implementing
Performance-Based Option of 10 CFR
Revision 0
Performance-Based Leak-Test Program
September 1995
Performance-Based Containment Leak-
Test Program
September 1995
Miscellaneous Condition Reports
CR 02269002
Unit 2 Main Turbine "B" Control Valve
Opened Unexpectedly From 50 Percent
to 75 Percent, Causing an Unintended
Power Rise and Reactor Coolant System
Temperature Reduction and Automatic
Control Rod Withdrawal
September 25, 2002
1R23
Temporary Plant Modifications
D. C. Cook Nuclear Plant UFSAR
Revision 17
12-EHP-5040-MOD-001
Revision 9
Disable East Travel Limit Switch on East Auxiliary Building Crane
12-EHP-5040-EMP-006
Disable Bridge East Travel Limit Switch
on East Auxiliary Building Crane
12-QM-3E
Revision 0
64
2002-1065-00
10 CFR 50.59 Applicability Determination
for 12-IHP-5040-EMP-006, Revision 0,
"Disable Bridge East Travel Limit Switch
on East Auxiliary Building Crane"
12-TC-02-64-R0
Design Packet for Temporary Condition
to Disable East Travel Limit Switch on
East Auxiliary Building Crane
PMP-4050-CHL-001
Revision 1
D. C. Cook Nuclear Plant UFSAR,
Section 9.7, "Reactor Components and
Fuel Handling System"
Revision 17
D. C. Cook Nuclear Plant UFSAR,
Section 12.2.1, "Control of Heavy Loads"
Revision 17
Plant Winterization
12-TM-00-61-R2
Winterization/De-Winterization
Temporary Modification to Support
12-IHP-5040-EMP-004
December 30, 2000
Job Order R0235054
Plant Winterization, Perform PM Task 30
September 27, 2002
RPA005058
Winterization of Tank Vents
12-EHP-5040-MOD-001
Revision 9
12-IHP-5040-EMP-004
Plant Winterization and De-winterization
Revision 3
D. C. Cook Nuclear Plant UFSAR,
Section 10.5, "Condensate and
Feedwater Systems"
Revision 17
Install Backup Power Supply for Control Group 1
1-TM-02-85-R0
Install Backup Power Supply for Control
Group 1
November 23, 2002
2002-1637-00
10 CFR 50.59 Applicability Determination
for 1-TM-02-85-R0, "Install Backup
Power Supply for Control Group 1"
November 23, 2002
Job Order 02326025-01
Install Temporary Modification
1-TM-02-85-R0
November 23, 2002
65
1EP2
Alert and Notification System (ANS) Testing
Berrien County Early Warning System
(EWS) Operations Manual
December 12, 2001
D. C. Cook Sirens and Contour Maps
January 2002
through
September 2002
Berrien County Monthly EWS Test
Reports
April 2002
through
October 2002
1EP3
Emergency Response Organization (ERO) Augmentation Testing
D. C. Cook Emergency Plan, Section E
Revision 17
D. C. Cook Emergency Plan, Section N
Revision 17
PMP-2080-EPP-100
Emergency Response
Revision 0
PMP-2080-EPP-107
Notification
Revision 16
SA-2001-SPS-014
Unannounced Drill
December 14, 2001
SA-2001-SPS-032
Semi-Annual Unannounced Drill
August 23, 2001
SA-2002-SPS-026
Unannounced Drill
March 14, 2002
SA-2002-SPS-027
Unannounced Drill
April 16, 2002
SA-2002-SPS-028
On-Shift Unannounced Drill
July 17, 2002
CR 02032031
December 14, 2001 Drill, Emergency
Operations Facility and Technical
Support Center Did Not Activate Within
60 Minutes
February 1, 2001
1EP5
Correction of Emergency Preparedness Weakness and Deficiencies
PA-01-18
PA Audit - Emergency Planning
February 25, 2002
PA-02-15
PA Audit - Emergency Planning
November 22, 2002
PMP-7030-CAP-001
Corrective Action Program Process Flow
Revision 13
SA-2001-SPS-026
Self-Assessment - Emergency Plan
Graded Exercise
July 25, 2001
66
SA-2001-SPS-036
Self-Assessment - 4th Quarter 2001 ERO
Drill
December 27, 2001
SA-2001-SPS-037
Self-Assessment - On-Shift Emergency
Planning Staffing Survey
February 28, 2002
SA-2002-SPS-013
Self-Assessment - 1th Quarter 2002
Accountability Drill
March 27, 2002
SA-2002-SPS-021
Self-Assessment - Emergency Plan Drill
August 7, 2002
SA-2002-SPS-022
Self-Assessment - Emergency Plan Drill
September 24, 2002
SA-2002-SPS-031
Self-Assessment - Review of NRC IN
October 21, 2002
CR 01247001
During The ESW Flow Restriction Event
on 8/29/01, a More Conservative
Decision Regarding Emergency Plan
Entry Would Have Been Appropriate
September 3, 2001
CR 02157101
Evaluate Timeliness of NRC Notification
For Unusual Event Based On PORV
Opening On June 5, 2002
June 6, 2002
CR 02163045
Catastrophic Failure Resulting in a Loss
Of Offsite Power Sources Supplied to
Reserve Feed
June 12, 2002
CR 02010029
Personnel Manning the New
Maintenance Contractor Building Did Not
Report a Failure with the Public Address
System to the Emergency Plan System
Engineer During the Accountability Drill
on January 8, 2002
January 10, 2002
CR 02204003
ERO Dialogic System Failed to Perform
as Required During a Forced Outage
Initiation
July 22, 2002
CR 02214013
Power Was Lost at the Buchanan Office
Building Due to a Storm Resulting in a
Loss of Power to the Emergency
Operations Facility
August 2, 2002
67
CR 02268022
The Process for Ensuring Qualified
Individuals Report for ERO Duties Failed
and as a Result the Operations Support
Center Was Activated During the 9/18/02
Drill with Unqualified Individuals and
Individuals That Had Not Been Confirmed
to be ERO Qualified
September 25, 2002
CR 02276062
Error in Minimum Staffing Requirement
Found in Table 1, Revision 17 of the
October 3, 2002
CR 02284015
Acceptable Interim Actions Have Not
Been Initiated and a Request for Project
Authorization Has Not Been Implemented
to Correct 20 Plant Locations Where it
Has Been Reported That the Public
Address System Is Not in Regulatory
Compliance
October 11, 2002
CR 02340005
Document Errors Reported in 2nd and 3rd
Quarter 2002 NRC DEP Performance
Indicator Data
December 5, 2002
1EP6
Drill Evaluation
Exercise Scope and Objectives for
October 24, 2002 Annual Exercise
Emergency Notification Forms
Completed During Annual Exercise
October 24, 2002
Desktop Guide for Emergency Planning
Performance Indicators
Revision 2
PMP-2080-EPP-107
Notifications
Revision 16
2OS1 Access Control to Radiologically Significant Areas
PMP-6010-RPP-003
High, Locked High, and Very High
Radiation Area Access
Revision 11
PMP-6010-RPP-006
Radiation Work Permit Program
Revision 7a
RP 014-01
Total Effective Dose Equivalent
Evaluation Worksheet for Work at 587
Foot Drumming Room Clean-up
September 23, 2002
68
RP 014-01
Total Effective Dose Equivalent
Evaluation Worksheet for Work on Spent
Fuel Pool Demineralizer High Pressure
Spray of the Inlet Retention Element
October 23, 2002
RP 014-01
Total Effective Dose Equivalent
Evaluation Worksheet for Unit 2 at Power
Entry to Work on Number 1, Safety
Injection Accumulator
November 8, 2002
RWP 020504
Restricted Area NRC Tours and
Inspections
Revision 10
RWP 021016
Resin Sluice activities - Locked High
Radiation Areas
Revision 6
RWP 021037; 617
617 Foot Demineralizer Locked High
Radiation Area Work Activities
Revision 3
RWP 021046; 587
Drumming Room Activities
Revision 1
RWP 021052
Unit 2 At Power Entry
Revision 2
RWP 02-1037
Radiation Protection ALARA [As-Low-As-
Reasonably-Achievable] Plan for Work
on Spent Fuel Pool Demineralizer High
Pressure Spray of the Inlet Retention
Element
Revision 1
RWP 02-1046
Radiation Protection ALARA Plan for
Work at 587 Foot Drumming Room
Clean-up
Revision 0
RWP 02-1052
Radiation Protection ALARA Plan for
Work on Reactor Coolant Pumps 11 and
14
Revision 1
12-THP-6010-RPP-006
Radiation Work Permit Processing
Revision 17
12-THP-6010-RPP-401
Performance of Radiation and
Contamination Surveys
Revision 10
12-THP-6010-RPP-418
Radiological Postings
Revision 9
CR 02217009
Modifications to Radiological Posting
Program
August 5, 2002
CR 02226075
Unnecessary Locked High Radiation
Areas
August 14, 2002
CR 02308023
Valve Released to Unrestricted Area
November 4, 2002
69
CR 02337041
Improper Receipt of Package Containing
Radioactive Source
November 27, 2002
2OS3 Radiation Monitoring Instrumentation
PA-02-06
Performance Assurance Audit, "Radiation
Protection"
April 16, 2002
12-THP06010-RPI-500
Instrument Issue and Operation Testing
Revision 13
12-THP06010-RPI-500
Instrument Issue and Operation Testing;
Data from Portal Monitor Operational
Checks Performed on December 5, 2002
Revision 13
12-THP06010-RPC.512
Calibration of the Eberline Smart Portable
Survey Meter(s)
Revision 5
12-THP06010-RPC.512
Calibration of the Eberline Smart Portable
Survey Meter(s); Data Sheet from
December 3, 2002
Revision 5
12-THP06010-RPC-513
Calibration of the Eberline Model R0-7
Survey Meter
Revision 2
12-THP06010-RPC-513
Calibration of the Eberline Model R0-7
Survey Meter; Data Sheet from
December 3, 2002
Revision 2
ALARA Radiation Protection Daily Dose
Report and Schedule
December 2, 2002
CR 02246017
Foot and Hand Monitor Found Out of
Service
September 3, 2002
CR 02249037
Instrument Missing from Work Area
September 6, 2002
CR 02287056
Discrepancies Between Laboratory
Cross-check Program
September 30, 2002
CR 02304023
Failure to Follow Procedural
Requirements for Instrument
Accountability
October 31, 2002
Stations Radiation Protection Instrumentation
Blitz Team Bulletin; Weekly Station
Performance Bulletin
December 3, 2002
70
Calibration Packages from a Selection of
Stations Radiation Protection
Instruments
December 2001
through
December 2002
Online Quality Control Schedule
November 27, 2002
Radiation Protection Instrument Use
History Analysis Forms; Selections from
Number 858-953
January 2002
through
December 2002
Report of Instruments Due for Calibration
December 31, 2002
4OA1 Performance Indicator (PI) Verification
Regulatory Assessment Performance
Indicator Guideline
Revision 2
SPP-2060-SFI-101
PI Data Gathering
Revision 0
PMP-7110.PIP.001
Regulatory Oversight Program PI
Revision 1
Letter from J. Pollock, American Electric
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 4Q2001 -- PI Data
Elements (QR and CR)"
January 17, 2002
Letter from J. Pollock, American Electric
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 1Q2002 -- PI Data
Elements (QR and CR)"
April 19, 2002
Letter from J. Pollock, American Electric
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 2Q2002 -- PI Data
Elements (QR and CR)"
July 18, 2002
Letter from J. Pollock, American Electric
Power, to the US NRC, Subject: "Cook
Unit 1 and 2 -- 3Q2002 -- PI Data
Elements (QR and CR)"
October 21, 2002
Administrative Technical Requirements
Units 1 and 2, Reactor Coolant System,
Supplemental Operational and
Surveillance Requirements
Revision 18
Results of Gamma Spectrometry Count
of Units 1 and 2 Reactor Coolant System
Specific Activity Samples
December 4, 2002
71
OHI-4032
Leakage Monitoring Program
Revision 2
12-THP-6020-CHM-101
Revision 14
12-EHP-5030-001-008
Recirculation Loop Total Leak Rate
Revision 3
12-THP-6020-INS-026
Gamma Spectrometry System
Revision 1
Licensee Event Reports
October 1, 2001
through
September 30, 2002
Non conservative Reactor Coolant
System Leakage Calculation
June 20, 1994
CR 01201019
Enhancements May Be Needed in the
Documentation of Test Results of 4 Alert
and Notification System Sirens in the
Two State Parks
July 20, 2001
CR 01325066
Resident Inspector Observations of Cook
Operations Training with Regard to EP
Performance Indicator Data Gathering
November 21, 2001
CR 02193022
Declining Trend in DEP Emergency
Planning NRC Performance Indicator
July 12, 2002
CR 2009038
Declining Trend in DEP Emergency
Planning NRC Performance Indicator
January 9, 2002
CR 02019069
Exceeding Limits for Hard Gammas in
Reactor Coolant System Filtrate Isotopic
Mixture
January 19, 2002
CR 02219004
E-BAR Determinations Found to Be
Slightly Erroneous
August 6, 2002
CR P-00-29181
Control Room Operability Evaluation, with
Subsequent Lowering of TS for Reactor
Coolant System Specific Activity
December 15, 1999
4OA2 Identification and Resolution of Problems
4OA2.1
EDG Starting Air Relay Failures
JO 00266004
Job Order - Unit 1 CD Diesel Failed to
Stop (Suspect 1-19-DGCD)
October 6, 2000
72
CR P-99-01279
Unit 2 AB EDG Rolled With Air by Itself.
No indication of a Start Signal was
Detected Locally or the Control Room.
Starting Air Continued to Blow Down
Engine Until Air Depleted.
January 21, 1999
CR P-99-01336
Found a Relay (EDG - Start Failure Relay
2-19-DGAB) in a De-energized State.
Voltage was Present at the Coil
Terminals Which Should Have Kept the
Relay in an Energized State
January 22, 1999
CR 00266004
Unit 1 CD Diesel Generator Failed to
Stop from the Control Room
September 22, 2000
CR 02289033
Diesel Generator 2AB 2-OME-150-AB
Starting Rolling Unexpectedly on Starting
Air
October 16, 2002
CR 02296037
Replace 1-19-1-DGCD. There Have
Been Three 19/19-1 Relay Failures Since
January 1999. They Are Original
Installation and Normally Energized.
October 23, 2002
4OA2.2
Common Cause Failure of Four Unit 1 Charging System Check Valves
Industry Operating Experience
Rev. 5
Part 21 Notification, Swing Check Valves
- Forged Steel
January 18, 1991
Problem Report 90-1503
Velan Valve Designs
October 12, 1990
Problem Report 92-157
Part 21 Issue Concerning Velan Valves
February 20, 1992
CR 96-0094
Operating Experience 7640, "Charging
Injection Valves Found Stuck Open"
January 24, 1996
CR 02132050
Disc on Valve 1-CS-329-L1 Was Found
in the Open Position
May 12, 2002
CR 02134021
Check Valves 1-CS-328-L1, 1-CS-328-
L4, 1-CS-329-L1, and 1-CS-329-L4 Were
Found Open During Radiographic
Nonintrusive Testing
May 14, 2002
CR 02205061
10 CFR 21 Evaluation Required for
Manufacturing Defects Discovered on
Velan 3-inch Check Valves
July 24, 2002
CR P-00-07039
NRC Information Notice 2000-08
May 16, 2000
73
CR 00278072
Operating Experience Number 11420
October 4, 2000
CR 01067004
Operating Experience 11950
March 1, 2001
CR 01136027
NRC Information Notice 2001-06
May 16, 2001
CR 01198006
Operating Experience 12454
July 17, 2001
CR 01198020
Operating Experience 12451
July 17, 2001
CR 01206021
Operating Experience 12454
July 25, 2001
CR 01362008
NRC Information Notice 2001-14
December 28, 2001
CR 02002020
NRC Information Notice 2001-19
January 2, 2002
74
4OA3 Event Follow-up
4OA3.1
Pressurizer Power Operated Relief Valve (PORV) Inadvertently Opened
During Testing Resulting in a Loss of Reactor Coolant System Inventory and
an Unusual Event
Unit 1 Declared an Unusual Event Due to
Reactor Coolant System Leakage
Greater Than 25 Gallons-Per-Minute
During Surveillance Testing
June 6, 2002
01-OHP-4030-102-017
Pressurizer PORV Actuation Channel
Calibration with Valve Operation (for
Modes 1, 2, and 3)
Revision 0
01-OHP-4030-102-017
Pressurizer PORV Actuation Channel
Calibration with Valve Operation (for
Modes 1, 2, and 3)
Revision 1
Daily Shift Managers Logs
June 5, 2002
through
June 6, 2002
CR 02157039
Pressurizer PORV 1-NRV-153 Opened
During Testing with its Block Valve Open,
Causing an Unexpected Release of
Reactor Coolant System Inventory into
the Pressurizer Relief Tank
June 6, 2002
CR 02157101
Evaluate Timeliness of NRC Notification
for Unusual Event Declared Based on
PORV Opening on June 5, 2002 at 23:00
June 6, 2002
4OA3.2
LER 315-1999-010-01, "Reactor Coolant System Leak Detection System
Sensitivity Not in Accordance with TS Basis"
LER 315-1999-010-00
Reactor Coolant System Leak Detection
System Sensitivity Not in Accordance
with TS Basis
May 3, 1999
LER 315-1999-010-01
Reactor Coolant System Leak Detection
System Sensitivity Not in Accordance
with TS Basis
March 6, 2000
Safety Evaluation of Westinghouse
Topical Reports Dealing with Elimination
of Postulated Pipe Breaks in PWR
(Pressurized Water Reactors) Primary
Main Loops
February 13, 1984
75
4OA3.3
Unanticipated Start of the Unit 2 Turbine Driven Auxiliary Feedwater Pump
(TDAFWP) During a Normal Plant Shutdown for Refueling Outage
LER 50-316-2002-04-00
Unanticipated Start of the Turbine Drive
Auxiliary Feedwater Pump
March 15, 2002
LER 50-316-2002-04-01
Unanticipated Start of the Turbine Drive
Auxiliary Feedwater Pump, Supplement 1
June 28, 2002
LER 50-316-2002-04-02
Unanticipated Start of the Turbine Drive
Auxiliary Feedwater Pump, Supplement 2
December 13, 2002
The Turbine Driven Auxiliary Feedwater
Pump Auto Started After a Scheduled
Reactor Trip From 20 Percent Power
January 19, 2002
02-OHP-4021-001-003
Power Reduction
Revision 15
2001-0985-00
10 CFR 50.59 Screening for Revision 15
to 02-OHP-4021-001-003, "Power
Reduction"
January 11, 2002
Daily Shift Managers Logs
January 19, 2002
CR 02019036
During Planned Reactor Trip the
TDAFWP Started
January 19, 2002
CR 02107016 (1)
CR 02019036 Evaluation Did Not
Address the Operational Aspects of the
TDAFWP Auto Start on the Planed
April 17, 2002
4OA3.4
LER 50-316-2002-05-00, "Unit 2 Trip Due to Instrument Rack 24-Volt DC
Power Supply Failure"
LER 50-316-2002-05-00
Unit 2 Trip Due to Instrument Rack
24-Volt DC Power Supply Failure
July 10, 2002
4OA3.5
LER 50-316-1997-004-02, "Analysis Demonstrates Design Basis Impact of
Inadequate Refueling Outage Safety Evaluation Was Negligible"
LER 316-1997-04-02
Analysis Demonstrates Design Basis
Impact of Inadequate Refueling Outage
Safety Evaluation was Negligible
January 22, 1998
LER 316-1997-04-01
Change to Component Cooling Water
Temperature Without Revision to UFSAR
November 17, 1997
76
LER 316-1997-04-00
Change to Component Cooling Water
Temperature Without Revision to UFSAR
Results in Condition Outside Design
Basis
September 22, 1997
CR 97-2342
Inadequate Safety Review Performed for
Establishment of a 90F Upper Limit for
Component Cooling Water During Unit 2
1996 Refueling Outage
August 26, 1997
Amendment Request
1202
Refueling Operations Decay Time
Technical Specification
November 16, 1994
Amendment Request
1202B
Response to Request for Additional
Information (RAI) Technical Specification
Amendment Refueling Operations Decay
Time
August 1, 1996
Amendment Request
1202A
Refueling Operations Decay Time
Updated Analysis and Response to RAI
February 1, 1996
Amendment Request
1202D
Response to RAI Regarding Refueling
Operations Decay Time
June 19, 1997
Amendment Request
1202F
Request to Withdraw the Refueling
Operations Decay Time Technical
Specification Amendment Request
January 27, 1998
Amendment Request
1146
Refueling Operations Decay Time
Technical Specification
November 30, 2001
4OA3.6
LER 50-315-1997-005-00, "Reactor Coolant Pump Fire Protection Inoperable
for Extended Period Without Compensatory Actions Due to Improperly
Fabricated Gasket in Spray Header Line"
LER 315-1997-05-00
Reactor Coolant Pump Fire Protection
Inoperable for Extended Period Without
Compensatory Actions due to Improperly
Fabricated Gasket in Spray Header Line
April 14, 1997
CR 97-0586
When Installing Blank Side of Spectacle
Flange Rubber Gasket Not Properly
Installed
March 5, 1997
77
4OA3.7
LER 50-315-1997-005-01, "Reactor Coolant Pump Fire Protection Inoperable
for Extended Period Without Compensatory Actions Due to Improperly
Fabricated Gasket in Spray Header Line"
LER 315-1997-05-00
Reactor Coolant Pump Fire Protection
Inoperable for Extended Period Without
Compensatory Actions due to Improperly
Fabricated Gasket in Spray Header Line
April 14, 1997
LER 315-1997-05-01
Reactor Coolant Pump Fire Protection
Inoperable for Extended Period Without
Compensatory Actions due to Improperly
Fabricated Gasket in Spray Header Line
October 23, 1997
CR 97-0586
When Installing Blank Side of Spectacle
Flange Rubber Gasket Not Properly
Installed
March 5, 1997
4OA3.8
LER 50-315-1998-056-01, "Inadequate Control and Processing of Design
Information Results in Unanalyzed Hot Leg Recirculation Switchover"
LER 50-315-1998-056-00
Inadequate Control and Processing of
Design Information Results in
Unanalyzed Hot Leg Recirculation
Switchover
January 6, 1999
LER 50-315-1998-056-01
Inadequate Control and Processing of
Design Information Results in
Unanalyzed Hot Leg Recirculation
Switchover
November 24, 1999
CR P-98-7848
Unanalyzed Condition Pertaining to Post-
Loss-of-Coolant Accident Emergency
Core Cooling System Hot Leg Switchover
December 11, 1998
4OA3.9
LER 50-315-1998-029-01, "Fuel Handling Area Ventilation System Inoperable
Due to Original Design Deficiency"
LER 50-315-1998-029-01
Fuel Handling Area Ventilation System
Inoperable Due to Original Design
Deficiency
August 4, 1999
Calculation No. RD-99-01
Control Room Dose Resulting from a
Fuel Handling Accident for Off-Load
Specific Conditions
Revision 1
78
CR P-98-01712
Fuel Handling Area Ventilation System
Inoperable Due to Original Design
Deficiency
April 22, 1998
4OA3.10
LER 50-315-1999-003-00, "Control Room Pressurization System Surveillance
Test Does Not Test System in Normal Operating Condition"
LER 50-315-1999-003-00
Control Room Pressurization System
Surveillance Test Does Not Test System
in Normal Operating Condition
February 24, 1999
CN-CRA-99-78
D.C. Cook TID-14844 Source Term Loss
of Coolant Accident Radiation Dose
Analysis
February 29, 2000
12 EHP 4030 STP 229
Control Room Emergency Ventilation
Test
Revision 3
American Electric Power
Purchase Order A 10342
NCS Corporation - Control Room
Envelope In-leakage Testing at D.C.
Cook Nuclear Plant 1999 - Final Report
August 11, 1999
CR P-99-00275
Control Room Pressurization System
Surveillance Test Does not Test System
in Normal Operating Condition
January 7, 1999
4OA3.11
LER 50-315-1999-003-01, "Control Room Pressurization System Surveillance
Test Does Not Test System in Normal Operating Condition"
LER 50-315-1999-003-01
Control Room Pressurization System
Surveillance Test Does Not Test System
in Normal Operating Condition
November 10, 2000
4OA3.12
LER 50-315-2000-004-00, "Circuit Design Could Result in Failure of
Emergency Diesel Generators to Load Properly After Loss of Offsite Power"
LER 50-315-2000-004-00
Circuit Design Could Result in Failure of
Emergency Diesel Generators to Load
Properly After Loss of Offsite Power
July 3, 2000
Safety Systems Response to Loss of
Coolant and Loss of Offsite Power
March 8, 1993
CR P-99-18884
Certain Automatic Safety Systems Could
Respond Inappropriately to Certain
Sequences of Loss of Coolant and Loss
of Offsite Power Events
July 19, 1999
79
4OA5.3
Inspector Follow-up Item (IFI) 50-315/316-99-29-01, "Review and Approval of
Dose Calculation for General Design Criteria 19 Control Room Habitability
Issue"
DIT-B-00069-00
Design Input for D. C. Cook Offsite and
Control Room Dose Analysis
July 21, 1999
DIT-B-00069-09
Design Input for D. C. Cook Offsite and
Control Room Dose Analysis
April 5, 2002
RD-01-05
Adjusted Dose Consequences for
Changes to Control Room
Revision 1
4OA5.4
IFI 50-316-00-07-03, "Failure to Perform Post Modification Checks to Verify
Adequate Clearance Between the Pressurizer Surge Line Whip Restraints and
the Surge Line under Hot Plant Conditions"
2-DCP-4260
Modification to Surge Line Pipe Whip
Restraints with Field Change Requests
Revision 0
SD-990825-001
HELB Structural Evaluation of Surge Line
Pipe Whip Restraints
Revision 3
CR 01089055
Calculation Impact Assessment Did Not
Adequately 03/30/2001 Address Impacts
Associated with Calculation
SD-990825-001
March 30, 2001
4OA5.5
URI 50-315/316-00-16-04, "Determine Whether the Latent Failure of a Test
Relay Should Be Treated under the Category of a Single Failure"
NRC Task Interface
Agreement No. 2000-12
Evaluation of the Engineering Safety
Features Safeguards Test Cabinet
November 11, 2000
4OA5.6
URI 50-315/316-01-15-01, "A Change Was Made to the UFSAR Without a
10 CFR 50.59 Evaluation"
CR 01291058
UFSAR Change Request Number 969,
Changed the Seismic Class of
Components Within the ESW System,
but Did Not Use CFR50.59 as a Basis for
the Change
October 18, 2001
(1)
Condition report written as a result of inspection activities.