ML040830634
| ML040830634 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/22/2004 |
| From: | Grobe J NRC/RGN-III |
| To: | Myers L FirstEnergy Nuclear Operating Co |
| References | |
| IR-04-002 | |
| Download: ML040830634 (65) | |
See also: IR 05000346/2004002
Text
March 22, 2004
Mr. Lew W. Myers
Chief Operating Officer
FirstEnergy Nuclear Operating Company
Davis-Besse Nuclear Power Station
5501 North State Route 2
Oak Harbor, OH 43449-9760
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION
NRC INTEGRATED INSPECTION REPORT 05000346/2004002
Dear Mr. Myers:
On February 14, 2004, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Davis-Besse Nuclear Power Station. The enclosed inspection report
documents the inspection findings which were discussed on January 23, February 25, 26,
and 27, 2004, with you or members of your staff.
The 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. For the entire inspection period, the Davis-Besse Nuclear Power Station was under
the Inspection Manual Chapter (IMC) 0350 Process. The Davis-Besse Oversight Panel
assessed inspection findings and other performance data to determine the required level and
focus of followup inspection activities and any other appropriate regulatory actions.
The report documents three inspection findings of very low safety significance (Green), two of
which involved violations of NRC requirements. The findings did not present any immediate
safety concerns. Because the violations were of very low safety significance and were entered
into your corrective action program, the NRC is treating these violations as Non-Cited
Violations, consistent with Section VI.A of the NRC Enforcement Policy. In addition, one issue
was reviewed under the NRC traditional enforcement process and was determined to be a
Severity Level IV violation of NRC requirements. Because this violation was non-wilful,
non-repetitive, and was entered into your corrective action program, the NRC is treating this
issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement
Policy.
This report also documents the closure of two items on the NRCs Restart Checklist. Item 2.a,
Reactor Pressure Vessel Head Replacement, was resolved through our inspection of
associated licensee activities and our independent inspections. Item 2.e, High Pressure
Injection Pump Internal Clearance and Debris Resolution, was closed based on our inspections
and evaluations of the modifications associated with the high pressure injection pumps to
resolve this issue.
L. Myers
-2-
If you contest the severity of any Non-Cited Violations, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with
copies to the Regional Administrator, Region III, 801 Warrenville Road, Lisle, IL 60532-4351;
the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington DC
20555-001; and the NRC Resident Inspector at Davis-Besse.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter
and its enclosure 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/
John A. Grobe, Chairman
Davis-Besse Oversight Panel
Docket No. 50-346
License No. NPF-3
Enclosure:
Inspection Report 05000346/2004002
w/Attachments: Supplemental Information
TIA 2003-04, Evaluation of Modifications to the High Pressure
Injection Pump
cc w/encl:
The Honorable Dennis Kucinich
G. Leidich, President - FENOC
Plant Manager
Manager - Regulatory Affairs
M. OReilly, Attorney, FirstEnergy
Ohio State Liaison Officer
R. Owen, Administrator, Ohio Department of Health
Public Utilities Commission of Ohio
President, Board of County Commissioners
Of Lucas County
C. Koebel, President, Ottawa County Board of Commissioners
D. Lochbaum, Union Of Concerned Scientists
J. Riccio, Greenpeace
P. Gunter, N.I.R.S.
DOCUMENT NAME: S:\\ADMIN\\Karen's Docs\\ML040830634.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
NAME
CLipa/trn
JGrobe
DATE
03/17/04
03/22/04
OFFICIAL RECORD COPY
L. Myers
-3-
ADAMS Distribution:
AJM
SPS1
RidsNrrDipmIipb
GEG
CST1
C. Ariano (hard copy)
C. Pederson, DRS (hard copy - IRs only)
DRPIII
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PLB1
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Enclosure
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-346
License No:
Report No:
Licensee:
FirstEnergy Nuclear Operating Company (FENOC)
Facility:
Davis-Besse Nuclear Power Station
Location:
5501 North State Route 2
Oak Harbor, OH 43449-9760
Dates:
January 1, 2004, through February 14, 2004
Inspectors:
S. Thomas, Senior Resident Inspector
J. Rutkowski, Resident Inspector
M. Salter-Williams, Resident Inspector
S. Campbell, Fermi Senior Resident Inspector
M. Holmberg, Reactor Inspector
T. Ploski, Senior Emergency Preparedness Inspector
B. Jickling, Emergency Preparedness Inspector
G. Costo, NRR Emergency Preparedness Specialist
Approved by:
Christine A. Lipa, Chief
Branch 4
Division of Reactor Projects
Enclosure
1
SUMMARY OF FINDINGS
IR 05000346/2004002; 1/1/2004 - 2/14/2004; Davis-Besse Nuclear Power Station; Adverse
Weather Protection, Identification and Resolution of Problems, and Other Activities
This report covers a 6 week period of resident inspection. The inspection was conducted by
resident and region based inspectors. Three green findings, two of which were associated with
Non-Cited Violations, were identified. In addition, one Severity Level IV Non-Cited Violation
was identified. The significance of most findings is indicated by their color (Green, White,
Yellow, Red) using Inspection Manual Chapter 0609, Significance Determination Process
(SDP). Findings for which the SDP does not apply may be Green or be assigned a severity
level after NRC management review. The NRCs program for overseeing the safe operation of
commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,
Revision 3, dated July 2000.
A.
Inspector-Identified and Self-Revealed Findings
Cornerstone: Emergency Preparedness
Severity Level IV. The inspectors identified that the licensee had changed its standard
emergency action level (EAL) scheme by revising one EALs criteria for an Unusual
Event declaration due to the initiation of the Steam and Feedwater Rupture Control
System as a result of a rapid depressurization of the secondary side. The inspectors
determined that this EAL change decreased the effectiveness of the emergency plan,
and that the licensee did not obtain prior NRC approval for this change, contrary to the
requirements of 10 CFR 50.54(q).
Because the issue affected the NRCs ability to perform its regulatory function, it was
evaluated with the traditional enforcement process as specified in Section IV.A.3 of the
Enforcement Policy. According to Supplement VIII of the Enforcement Policy, this
finding was determined to be a Severity Level IV because it involved a failure to meet a
requirement not directly related to assessment and notification. Further, this problem
was isolated to one EAL and was not indicative of a functional problem with the EAL
scheme. Additionally, because the licensee entered this issue into its corrective action
program and completed adequate corrective actions, this finding is being treated as a
Severity Level IV Non-Cited Violation of 10 CFR 50.54(q). (Section 1EP4)
Cornerstone: Mitigating Systems
Green. A finding of very low safety significance was self-revealed when, during
performance of a functional test for the Steam Feedwater Rupture Control System
(SFRCS) steam generator 2 differential pressure switch, the licensee did not perform
the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statement of Technical Specification 3.3.2.2. The pressure switch
was isolated for a period of approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 minutes without control room
knowledge. This rendered the pressure switch incapable of sensing differential
pressure and providing a signal, if needed, to the SFRCS actuation channel 2. Plant
procedures require maintaining knowledge of the proper and actual status of Technical
Specification listed equipment.
Enclosure
2
The finding was more than minor because it involved the configuration control and
human performance attributes of the Mitigating Systems Cornerstone and affected the
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences. The issue was a
Non-Cited Violation of Technical Specification 6.8.1 which required the implementation
of written procedures governing plant operations. (Section 4OA5)
Green. The inspectors identified a finding of very low safety significance and associated
NCV for the licensees failure to determine the cause and implement actions to prevent
recurrence for the inadequate design changes (removed air accumulators) made to the
air operated service water system valves at the outlet of the component cooling water
heat exchangers. Although the licensee had implemented corrective measures for the
service water valve design deficiencies, the licensee failed to recognize the need for a
root cause investigation and to take actions to prevent recurrence for the inadequate
modification process until questioned by the NRC inspectors.
This finding was greater than minor because this example was associated with the
Mitigating Systems Cornerstone and if left uncorrected, could potentially result in other
inoperable safety related equipment or systems. The finding was determined to be of
very low safety significance by management review, because the licensee had taken
actions to restore the air operated service water valves to an operable configuration
and, after identification by the inspectors, the licensee entered the failure to identify the
cause(s) and implement action(s) to prevent recurrence for the inadequate modification
into the corrective action program. This issue was a NCV of 10 CFR 50 Appendix B
Criteria XVI, Corrective Action. (Section 4OA2).
Green. A finding of very low safety significance was identified by the inspectors for
inadequate preparations for the onset of frazil ice conditions prior to January 6, 2004.
Lack of coordination between licensee departments resulted in incomplete preparations
prior to the onset of frazil ice conditions.
The inspectors determined that the finding was more than minor because, if left
uncorrected, it could contribute to the likelihood of those events that upset plant stability.
Specifically, the failure to adequately prepare for frazil ice conditions could result in a
plant shutdown. The finding was of very low safety significance because the finding:
(1) was not associated with the likehood of primary or secondary system LOCA
initiation; (2) did not contribute to the likelihood that mitigation systems would be
unavailable; and (3) was not associated with fire or flood. No violation of NRC
requirements occurred. (Section 1R01.1)
B.
Licensee Identified Findings
Violations of very low safety significance, which were identified by the licensee, have
been reviewed by the inspectors. Corrective actions taken or planned by the licensee
have been entered into the licensees corrective action program. These violations are
listed in Section 4OA7.
Enclosure
3
REPORT DETAILS
Summary of Plant Status
The plant was shutdown on February 16, 2002, for a refueling outage. During scheduled
inspections of the control rod drive mechanism nozzles, significant degradation of the reactor
vessel head was discovered. As a direct result of the need to resolve many issues surrounding
the Davis-Besse reactor vessel head degradation, NRC management decided to implement
IMC 0350, Oversight of Operating Reactor Facilities in a Shutdown Condition With
Performance Problems. Significant dates for this extended outage were as follows:
fuel was removed from the reactor on June 26, 2002;
entered operational Mode 6 on February 19, 2003;
fuel reload was completed on February 26, 2003;
entered operational Mode 5 on March 12, 2003;
entered operational Mode 4 on September 13, 2003;
entered operational Mode 3 on September 14, 2003;
completed the normal operating pressure test for the reactor coolant system and started
cooldown to Mode 5 on September 30, 2003;
entered operational Mode 4 on December 28, 2003; and
entered operational Mode 3 on December 30, 2003.
On January 2, 2004, the licensee re-commenced a heatup and achieved normal operating
pressure and temperature on January 5, 2004. On January 8, 2004, the licensee commenced
a reactor plant cooldown from operational Mode 3 to Mode 4. This cooldown was required by
Technical Specification (TS) due to the inoperability of one train of auxiliary feedwater. The
plant entered Mode 4 on January 9, 2004. On January 26, 2004, the plant entered Mode 3 and
attained normal operating pressure and temperature on January 28, 2004. For the entire
inspection period, the Davis-Besse Nuclear Power Station was under the IMC 0350 Process.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and
1R01
Adverse Weather Protection (71111.01)
.1
Insufficient Preparations for Frazil Ice Conditions
a.
Inspection Scope
The inspectors reviewed the licensees cold weather readiness by verifying cold weather
design features and implementation of the licensees procedure DB-OP-06931,
Seasonal Plant Preparation Checklist. The inspectors evaluated the licensees
readiness for seasonal susceptibilities and impending adverse weather conditions.
Enclosure
4
b.
Findings
Introduction: The inspectors identified a finding of very low safety significance for not
being adequately prepared for the onset of frazil ice conditions prior to January 6, 2004,
a point at which the conditions for icing of the intake crib existed.
Description: Davis-Besse Procedure DB-OP-06913, Seasonal Plant Preparation
Checklist, described the conditions when icing of the intake crib could exist.
Specifically these conditions were:
lake temperature near freezing point;
lake level low in the range of 569-570 feet;
windy conditions with low air temperatures; and
no ice cap formed on the lake.
The procedure stated that, by November 1 of each year, arrangements should be made
with the Maintenance Services Department to obtain a high capacity trash pump, suction
and discharge piping necessary to support pump operations and that the equipment be
stored in a suitable location for future use. The purpose of the high capacity pump was
to provide the ability to pump water from the lake to the intake forebay if required.
On January 6, 2004, upon observing decreasing forebay level, the resident inspectors
questioned the operations staff as to whether they were monitoring forebay level and if
they recognized that conditions existed that were conducive to the formation of frazil ice
conditions. As a result of the inspectors questions, the licensee determined that the
conditions for frazil ice formation in the intake crib existed and that as of
January 6, 2004, no preparations for staging of the pump and hoses nor the ability to
contact personnel to provide the pump and hoses on a short notice had been arranged.
The licensee made arrangements to have the high capacity pump and hoses staged on
January 7, 2004.
Analysis: The inspectors determined that not sufficiently coordinating and being
adequately prepared for frazil conditions prior to January 6, 2004, was a performance
deficiency warranting a significance evaluation in accordance with IMC 0612, Power
Reactor Inspection Reports, Appendix B, Issue Screening, issued on June 20, 2003.
The inspectors determined that the finding was more than minor because, if left
uncorrected, it could contribute to the likelihood of those events that upset plant stability.
Specifically, the failure to adequately prepare for frazil ice conditions could result in a
plant shutdown as required by DB-OP-06913. Utilizing the Phase 1 Screening
Worksheet, per Inspection Manual Chapter 0609, Significance Determination Process,
the inspectors determined this performance deficiency impacted the Initiating Event
Cornerstone because it constituted a transient initiator contributor. The inspectors
answered no to Phase 1 Initiating Event questions because the finding: (1) was not
associated with the likehood of primary or secondary system LOCA initiation; (2) did not
contribute to the likelihood that mitigation systems would be unavailable; and (3) was not
associated with fire or flood.
Enforcement: The Seasonal Plant Preparation Checklist was not required by
10 CFR Part 50, Appendix B; therefore, no violation of regulatory requirements
Enclosure
5
occurred. This issue was considered to be a finding of very low safety significance
(FIN 50-346/2004002-01). This licensee entered the event into its corrective action
system as CR 04-00179.
.2
Procedure Specified Heater not Available for EDG 2 Room
a.
Inspection Scope
The inspectors reviewed the licensees response to unexpected annunciator alarm
1-1-K EDG 2 TRBL received on January 25, 2004, which was the result of decreasing
temperatures in EDG 2 Room. The inspectors interviewed licensee personnel and
reviewed control room logs, alarm procedures, operating procedures, and condition
reports.
b.
Findings
Licensee was unable to locate the necessary equipment required per the annunciator
alarm response procedure to provide additional temporary heating for the room.
Section 4OA5 discusses the regulatory aspects of this finding.
1R05
Fire Protection (71111.05Q)
.1
Area Inspections
a.
Inspection Scope
The inspectors conducted fire protection inspections, which were focused on the
availability, accessibility, and condition of fire fighting equipment, the control of transient
combustibles, and the condition and operating status of installed fire barriers. The
inspectors selected fire areas for inspection based on their overall contribution to
internal fire risk, as documented in the Individual Plant Examination of External Events,
their potential to impact equipment which could initiate a plant transient, or their impact
on the plants ability to respond to a security event. Using the documents listed at the
end of this report, the inspectors verified that fire hoses and extinguishers were in their
designated locations and available for immediate use, that fire detectors and sprinklers
were unobstructed, that transient material loading was within the analyzed limits, and
that fire doors, dampers, and penetration seals appeared to be in satisfactory condition.
The following two areas were inspected:
Fire Area R; Auxiliary Shutdown Panel And Transfer Switch Room
Fire Area DG; No. 1 Electrical Penetration Room
b.
Findings
No findings of significance were identified.
Enclosure
6
1R13
Maintenance Risk Assessment and Emergent Work Evaluation (71111.13)
a.
Inspection Scope
The inspectors reviewed the licensees response to risk significant activities. These
activities were chosen based on their potential impact on increasing overall plant risk.
The inspection was conducted to verify the planning, control, and performance of the
work were done in a manner to reduce overall plant risk and minimize the duration
where practical, and that contingency plans were in place where appropriate. The
licensees daily configuration risk assessments, observations of shift turnover meetings,
observations of daily plant status meetings, and the documents listed at the end of this
report were used by the inspectors to verify that the equipment configurations had been
properly listed, that protected equipment had been identified and was being controlled
where appropriate, and that significant aspects of plant risk were being communicated
to the necessary personnel. The following nine risk significant issues were evaluated by
the inspectors:
On January 4, 2004, the development and implementation of a problem solving
plan to investigate and respond to a Reactor Coolant Pump 2-2 High Seal Return
Flow alarm which, if concurrent with other indications, would indicate degrading
seal performance and potentially seal failure;
On January 6, 2004, the licensee response to problems with Auxiliary Feedwater
Pump 1 governor speed increaser motor and subsequent troubleshooting of a
casing steam leak. The speed increaser motor was replaced and the steam leak
was initially addressed by a temporary modification (see Section 1R23);
On January 19-22, 2004, the modification of 10 Turbine building doors due to a
preliminary analysis of the effects of a Main Steam Line Break using new
computer models which initially determined that these doors would not be able to
withstand the initial pressure wave caused by a guillotine break of the main
steam line and could subject safety related equipment to an adverse
environment;
On January 22, 2004, the replacement of the gasket for Feed Water Discharge
Valve 1009 [Motor Driven Feed Pump to Main Feedwater Discharge Check]
which required the securing of the Motor Driven Feedwater Pump and the use of
the Startup Feedwater Pump to maintain feedwater flow to the steam generators;
On January 26-27, 2004, the replacement of Main Steam Line 1 Isolation Valve,
Solenoid Valve SV101C which required defeating Turbine Bypass Valve and
Main Steam Line 1 Isolation Valve interlocks;
On January 29, 2004, the development and implementation of a problem solving
plan to investigate and respond to the reoccurrence of a casing steam leak on
Auxiliary Feed Water Pump 1 observed during the performance of the AFPT 1
Quarterly Test;
On February 2, the development and implementation of a problem solving plan
to investigate and respond to the Startup Transformer X01 A Phase Bushing oil
leak. The licensee removed the transformer from service and entered
TS 3.8.1.1.(a) due to having one offsite circuit of A.C. Electrical power
inoperable. The licensee performed a 10 CFR 50.59 evaluation to implement
changes to the Davis-Besse TS Bases, Updated Safety Analysis Report (USAR),
plant procedures and associated engineering analysis to allow the onsite
Class 1E AC power system to be powered from the offsite power source by a
Enclosure
7
qualified circuit consisting of the Main Power Transformer backfed from the
345 kV offsite transmission system and powering the Unit Auxiliary Transformer
and 13.8 kV buses from the Main Power Transformer;
On February 7 - 8, 2004, entered an Orange risk condition with both startup
transformer X02 and the 345 kV Bus K isolated to remove the Ohio-Edison Line
from service to replace cotter pins missing from the clevis of eight towers outside
of Sandusky, Ohio; and
On February 13, 2004, the development and implementation of a problem
solving plan to investigate and respond to Component Cooling Water Pump
Room Ventilation Train 2 anomalies.
b.
Findings
No findings of significance were identified.
1R14
Personnel Performance During Nonroutine Plant Evolutions (71111.14)
a.
Inspection Scope
The inspectors reviewed the licensees performance during planned non-routine
evolutions. The inspectors attended Infrequently Performed Evolution briefs, pre-job
briefs, reviewed operations evolution orders, and observed evolutions in the control
room and in the field. The inspectors reviewed selected prior events to determine if they
were adequately addressed to prevent recurring events, verified if the operators
responded in accordance with procedures and training, and determined if the evolution
was conducted in a safe and conservative manner. The following six non-routine
evolutions were reviewed:
On January 8 - 9, 2004, the inspectors observed operations personnel
performance during a reactor plant cooldown from operational Mode 3 to
Mode 4. This cooldown was required by TS 3.7.1.2.(a) due to the inoperability of
one train of auxiliary feedwater;
On January 22, 2004, the inspectors observed operations personnel
performance during the swapping of the motor driven feedwater pump and the
start-up feedwater pump to facilitate replacement of the gasket for Feed Water
Discharge Valve 1009 [Motor Driven Feed Pump to Main Feedwater Discharge
Check];
On January 26, 2004, the inspectors observed control room personnel
performance during the starting of Reactor Coolant Pump 1-1 during plant
heat-up to normal operational temperature and pressure;
On January 26, 2004, the inspectors observed operations personnel
performance during the bypassing of Main Steam Isolation Valve MS 101 to
facilitate replacement of Steam Feed Rupture Control System Solenoid Valve
SV101C;
On January 27, 2004, the inspectors observed control room personnel
performance during the reactor coolant system isolation check valve leak test for
Decay Heat Valves DH 76 and DH 77. The purpose of the test was to
demonstrate the operability of the Reactor Coolant system pressure isolation
valves, by individually leak testing check valves DH 76 and DH 77; and
Enclosure
8
On February 7, 2004, the inspectors reviewed the performance of the control
room operators and their oversight management during the Absolute Position
Indication Functional Test. The purpose of the test was to verify functionality of
the Absolute Position Indication Channels of the Control Rods. The inspectors
observed the Infrequently Performed Evolution briefs, the Pre Job briefs and
observed evolutions in the control room and at the control rod cabinets.
b.
Findings
No findings of significance were identified.
1R19
Post-Maintenance Testing (71111.19)
a.
Inspection Scope
The inspectors reviewed post-maintenance testing activities to ensure that the testing
adequately verified system operability and functional capability with consideration of the
actual maintenance performed. The inspectors used the appropriate sections of the
TSs and the USAR, as well as the documents listed at the end of this report, to evaluate
the scope of the maintenance and verify that the work control documents required
sufficient post-maintenance testing to adequately demonstrate that the maintenance
was successful and that operability was restored. The inspectors observed and
evaluated test activities associated with the following four samples:
On January 3, 2004, testing the functionality of the transfer of Auxiliary
Feedwater Pump Suction, within 10 seconds, to Service Water upon loss of
normal supply after the replacement of a time delay relay in the circuitry;
On January 16, 2004, testing the Auxiliary Feedwater Pump Turbine 1 High
Speed Stop and Overspeed Trip setting after repairing a pump casing steam
leak;
On January 17, 2004, testing the Auxiliary Feedwater Pump Turbine 2 High
Speed Stop and Overspeed Trip setting after replacing the turbine casing seal
material; and
On January 27, 2004, stroke time testing of the Main Steam Line 1 Isolation
Valve after replacement of Main Steam Line 1 Isolation Valve Solenoid Valve
SOV101C.
b.
Findings
No findings of significance were identified.
Enclosure
9
1R20
Refueling and Outage (71111.20)
a.
Inspection Scope
The inspectors evaluated the licensees efforts to remove loose debris from containment
as required by Plant Procedure DB-OP-06900, Attachment 11, Revision 15, and
On January 6, 2004, the inspectors made a containment under vessel entry with a
licensee team that was doing VT-2 and boric acid inspections. The activities of the
accompanied team included removing FLUS temporary instrumentation (installed under
a temporary modification), placing permanent placards on I-beams, looking for any
indication of leakage, and verifying that the under vessel area was free of debris and
any equipment other than that designed to be in that location. The inspectors also
attended the brief for entry under the reactor vessel, which was an entry into a locked
high radiation area, and observed radiation work practices and use of dosimetry.
On January 19, 2004, the inspectors made a tour of the containment including the east
steam generator enclosure, the incore instrument tank area, and the 565' floor elevation
outside of the steam generator enclosures.
b.
Findings
No findings of significance were identified.
1R22
Surveillance Testing (71111.22)
a.
Inspection Scope
The inspectors observed the surveillance test and/or evaluated test data to verify that
the equipment tested met TSs, USAR, and licensee procedural requirements, and also
demonstrated that the equipment was capable of performing its intended safety
functions. The inspectors used the documents listed at the end of this report to verify
that the test met the TS frequency requirements; that the test was conducted in
accordance with the procedures, including establishing the proper plant conditions and
prerequisites; that the test acceptance criteria were met; and that the results of the test
were properly reviewed and recorded.
The following two activities were evaluated:
On January 28, 2004, the functional test of the auxiliary feed pump turbine 1 inlet
isolation on low steam line pressure interlock; and
On February 9, 2004, the control rod assembly insertion time test. The
inspectors attended pre-activity briefs and witnessed the majority of the testing
either from the control room or at the control rod drive mechanism panels.
b.
Findings
No findings of significance were identified. All control rod drive mechanisms met their
insertion time acceptance criteria.
Enclosure
10
1R23
Temporary Plant Modifications (71111.23)
a.
Inspection Scope
The inspectors reviewed Temporary Modification 04-0002, K3-1 Auxiliary Feedwater
Pump Turbine, to verify that the modification did not affect the safety functions of this
risk significant safety system. The temporary modification was a 24 gauge stainless
steel deflector and the installation of a temporary material on the turbine governor
cooling lines above the outboard bearing housing. The purpose of the modification was
to direct moisture from a steam casing leak away from the outboard bearing housing.
The inspectors reviewed the temporary modification package and associated
10 CFR 50.59 screening and compared them to system, USAR, and TS requirements to
determine if there were any effects on system operability or availability and to verify
temporary modification consistency with plant documentation and procedures.
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 provisions for
the operation, maintenance, and periodic testing of the ANS in the Davis-Besse
Stations Emergency Planning Zone to determine whether the ANS equipment was
adequately maintained and tested in accordance with Emergency Plan commitments
and procedures. The inspectors reviewed records of 2002 and 2003 preventive and
non-scheduled maintenance activities and a sample of 2003 ANS operability test results.
The inspectors also discussed the status of the ANS siren upgrade project, which was
roughly 50 percent complete, and determined that work on this project was expected to
resume in 2005.
b.
Findings
No findings of significance were identified.
1EP3
Emergency Response Organization (ERO) Augmentation Testing (71114.03)
a.
Inspection Scope
The inspectors reviewed and discussed with EP staff the procedures that included the
primary and alternate methods of initiating an ERO activation to augment the onshift
ERO, plus the provisions for maintaining the ERO call-out roster and for periodically
updating the ERO Telephone Directory. The inspectors also reviewed critique records
of semi-annual, unannounced, off-hours staff augmentation drills that were conducted in
Enclosure
11
2002 and 2003 to determine the adequacy of the drills critiques and associated
corrective actions.
The inspectors also reviewed training records of a random sample of 30 Davis-Besse
Station personnel, who were assigned to key and support ERO positions, to determine
whether they were currently trained for their assigned ERO positions.
b.
Findings
No findings of significance were identified.
1EP4
Emergency Action Level and Emergency Plan Changes
a.
Inspection Scope
The inspectors, with NRC Headquarters assistance, performed a review of the current
site-specific Emergency Action Levels (EAL), as found in Revision 3 to Emergency Plan
Implementing Procedure (EPIP) RA-EP-01500, and Revision 10 to discontinued EPIP
EI-1300.01 to determine whether refinements or other changes made to any EAL since
1985 may have decreased the effectiveness of the licensees emergency planning.
Applicable portions of the following were used as reference criteria: 10 CFR 50.54(q);
10 CFR 50.47(b); 10 CFR Part 50, Appendix E; Revision 1 of Nuclear Regulatory Guide
(NUREG) 0654, Criteria for Preparation and Evaluation of Radiological Emergency
Response Plans and Preparedness in Support of Nuclear Power Plants; and Revision 2
of the Nuclear Management and Resources Council/National Environmental Studies
Project-007, Methodology for Development of Emergency Action Levels.
The inspectors also reviewed a sample of records that provided the bases of
refinements and other changes that were made to certain EALs since 1985 in response
to concerns identified in 1985 and 1986 NRC Inspection Reports and various licensee
self-assessments.
b.
Findings
Introduction: The licensee changed one indicator of its Unusual Event EAL 5.A.1, which
addressed initiation of the Steam and Feedwater Rupture Control System (SFRCS)
following a rapid depressurization of the secondary side, such that this indicator only
clearly addressed an automatic initiation of the SFRCS and not also a manual initiation
of this system, as was clearly stated in the prior version of this EAL. The NRC
determined that the change to this indicator decreased the effectiveness of the
licensees emergency plan. The licensee did not submit this change to NRC for prior
approval. This is a violation of 10 CFR 50.54(q) and, because it impacted the regulatory
process, traditional enforcement was applied. Since this issue was entered into the
licensees corrective action program, adequate corrective actions were completed, and
because this item involved a failure to meet a regulatory requirement not directly related
to assessment or notification, this issue was determined to be a Severity Level IV
Non-Cited Violation (NCV).
Enclosure
12
Description: The licensees site-specific EALs were based on the guidance of
NUREG 0654, Revision 1. The EAL 5.A.1 was the site-specific interpretation of the
NUREGs Unusual Event Example Initiating Condition 17, rapid depressurization of a
pressurized water reactors secondary side. In 1988, the licensee revised one indicator
of Unusual Event EAL 5.A.1 to delete words which clearly indicated that either a manual
or an automatic initiation of the SFRCS on low main steam line pressure would warrant
an Unusual Event declaration. The EP staff informed the inspectors that this indicator
was changed due to feedback from a Control Room Simulator training session. The
wording of the previous and current EALs were as follows:
Previous EAL Indicators
Increasing containment pressure (if leak is inside containment) OR unusually loud noise
OR visual sighting outside containment
AND
Valid SFRCS initiation automatically or manually on low main steam line pressure
Current EAL Indicators
Increasing containment pressure (if leak is inside containment)
Unusually Loud Noise
Visual sighting outside containment
AND
Valid SFRCS automatic initiation on low main steam line pressure
The licensees analysis of the revised indicator involving SFRCS initiation was that a
manual initiation of the SFRCS to preclude its automatic actuation was considered to be
an automatic actuation. When the EAL was revised, the licensee concluded that the
above change to the indicator involving SFRCS initiation did not decrease the
effectiveness of the emergency plan.
In contrast, the inspectors determined that the change to this indicator represented a
decrease in effectiveness of the emergency plan because the re-worded indicator
narrowed the scope of the indicator by not clearly addressing manual initiation of the
SFRCS. The wording of an EALs indicator needed to be straightforward such that any
Shift Manager could make a timely and accurate decision on whether or not to declare
an emergency without having to recall details from a training session or research other
procedures.
The inspectors concluded that the aforementioned change to the SFRCS indicator and
its technical bases should have been submitted for NRC review and approval prior to
implementation of revised EAL 5.A.1. However, since the licensee had concluded in
1988 that the change to this indicator did not decrease the effectiveness of the
emergency plan, this change was not submitted to the NRC for review prior to
implementation of the revised indicator.
Enclosure
13
Analysis: The inspectors determined that the licensee failed to meet the requirements
of 10 CFR 50.54(q) when it failed to identify a decrease in effectiveness of its standard
EAL classification scheme following the 1988 revision. A standard classification and
action level scheme is required by 10 CFR 50.47(b)(4). Additionally, no actual safety
consequence was identified; however, the inspectors determined that the issue had a
potential for impacting the NRCs ability to perform its regulatory function. Therefore, in
accordance with NRCs Enforcement Policy and Appendix B of Manual Chapter 0609,
traditional enforcement was applied instead of the Significance Determination Process
(SDP).
Enforcement: 10 CFR 50.54(q) states, in part, that the licensee may make changes to
these plans without Commission approval only if the changes do not decrease the
effectiveness of the plans. Proposed changes that decrease the effectiveness of the
approved emergency plans may not be implemented without application to and approval
by the Commission. In 1988, the licensee made a change to its standard EAL scheme
in the EPIPs that reduced the effectiveness of the emergency plan. This change
was not submitted to the NRC for approval prior to implementation in October 1988.
The licensee entered this issue into their corrective action program as Condition
Report (CR) 04-01475 and completed adequate corrective actions. Corrective
actions included revising the affected emergency plan implementing procedure and
issuing required reading packages to relevant personnel on the bases for this
procedure revision.
Changing an emergency plan commitment without prior NRC approval impacts the
NRCs ability to perform its regulatory function and is therefore processed through
traditional enforcement, as specified in Section IV.A.3 of the Enforcement Policy, issued
May 1, 2000 (65 FR 25388). According to Supplement VIII of the Enforcement Policy,
this finding was determined to be a Severity Level IV because it involved a failure to
meet a requirement not directly related to assessment and notification. Further, this
problem was isolated to one EAL and was not indicative of a functional problem with the
licensees EAL scheme. Additionally, because the licensee entered this issue into its
corrective action program (CR04-01475) and has completed adequate corrective
actions, this finding is being treated as Non-Cited Violation (Severity Level IV) consistent
with Section VI.A of the Enforcement Policy. (NCV 50-346/2004002-02).
1EP5
Correction of Emergency Preparedness Weaknesses and Deficiencies (71114.05)
a.
Inspection Scope
The inspectors reviewed those quarterly audits, which were performed during 2002 and
2003 by Nuclear Quality Assessment staff, that addressed various aspects of the EP
program and a sample of resulting corrective action documents to verify that these
independent assessments met the requirements of 10 CFR 50.54(t) and that adequate
corrective actions were taken on identified concerns. The inspectors also reviewed a
sample of critique reports and corrective action documents that were associated with the
2003 biennial exercise, as well as EP drills conducted in 2002 and 2003, to verify that
the licensee fulfilled its drill commitments and to evaluate the licensees efforts to
identify, track, and resolve concerns identified during these activities. The inspectors
also reviewed the EP staffs self-assessment report and a sample of related corrective
Enclosure
14
action program records associated with an actual Unusual Event declaration in
August 2003 due to a regional power blackout that affected offsite power supplies to the
Davis-Besse Station.
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES
4OA1 Performance Indicator (PI) Verification (71151)
a.
Inspection Scope
The inspectors reviewed the licensees records associated with the performance
indicators (PI) listed below. The inspectors verified that the licensee accurately
reported the indicators in accordance with relevant procedures and Nuclear Energy
Institute guidance endorsed by NRC. Specifically, the inspectors reviewed licensee
records associated with PI data reported to the NRC for the period April 2003 through
December 2003. Reviewed records included: procedural guidance on assessing
opportunities for the three PIs; assessments of PI opportunities during pre-designated
Control Room Simulator training sessions, the 2003 biennial exercise, and drills;
revisions of the roster of personnel assigned to key ERO positions; and results of
periodic ANS operability tests. The following PIs were reviewed:
Alert and Notification System;
Emergency Response Organization Drill Participation; and
Drill and Exercise Performance.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
.1
Routine Review of Identification and Resolution of Problems
a.
Inspection Scope
The inspectors routinely reviewed issues during baseline inspection activities and plant
status reviews to verify that they were being entered into the licensees corrective action
system at the appropriate threshold, that adequate attention was being given to timely
corrective actions, and that adverse trends were identified and addressed. Minor issues
entered into the licensees corrective action system as a result of inspectors
observations are included in the list of documents reviewed which is attached to this
report.
Enclosure
15
b.
Findings
No findings of significance were identified.
.2
Auxiliary Feedwater Pump 1 Governor Issues
a.
Inspection Scope
On September 23, 2003, and again on January 31, 2004, Auxiliary Feedwater Pump 1
failed to meet the acceptance criteria for its response time test. This issue was selected
as one of the Identification and Resolution of Problems samples for further in-depth
evaluation due to the potential risk significance aspect of Auxiliary Feedwater Pump 1
not being able to meet its response time. The inspectors reviewed condition reports,
root cause evaluations, maintenance history, engineering analyses, previous corrective
actions and other licensee documentation to ascertain the previous deficiencies
identified for the response time test failures. The inspectors verified that the licensees
corrective actions for this issue included the following performance attributes:
complete and accurate identification of the problem in a timely manner
commensurate with its significance;
evaluation and disposition of operability/reportability issues;
consideration of extent of condition, generic implications, common cause, and
previous occurrences;
classification and prioritization of the resolution of the problem commensurate
with its safety significance;
identification of corrective actions which are appropriately focused to correct the
problem; and
completion of corrective actions in a timely manner commensurate with the
safety significance of the issue.
b.
Findings and Observations
There were no findings identified associated with the reviewed deficiencies; however,
the licensee investigation of the most recent response time test failure was ongoing.
In response to the failure on January 31, 2004, the licensee replaced the governor
on AFPT 1, and the overall response time test results for the new governor were within
specifications.
.3
Operations Root Cause Corrective Actions
a.
Inspection Scope
On January 6, 2004, during the performance of DB-MI-03204, Channel Functional
Test and Calibration of SFRCS Actuation Channel 2 Steam Generator Differential
Pressure Inputs, the licensee did not perform Action Statement 16 of TS 3.3.2.2 [see
Section 4OA5]. As a result of the missed TS action statement, the licensee conducted a
root cause investigation (CR 04-00181) to determine why the organizations
administrative controls did not exhibit the appropriate level of rigor and formality towards
the adherence to licensed conditions. The licensee attributed the root cause of the
Enclosure
16
event to: (1) less than adequate implementation of work practices, and (2) less
than adequate implementation of managerial methods. The inspectors evaluated
CR 04-00181 and its associated corrective actions as one of the Identification and
Resolution of Problems samples for further in-depth evaluation.
The inspectors reviewed the condition report to ensure that:
the full extent of the condition, generic implications, common cause and
previous occurrences were considered;
the issues were properly classified and prioritized;
the root causes and contributing causes were identified;
the corrective actions were appropriately focused to correct the problem;
the corrective actions were completed within a timely manner; and
the effectiveness of previous actions designed to correct similar events were
considered.
The licensee concluded that the causes and contributing factors for inconsistent crew
performance could only be addressed through continual involvement of station
managements monitoring, coaching, feedback and correction. As a result, Operations
management developed a series of corrective actions that were to be completed prior to
entry to Mode 3 or prior to reactor startup. Any remaining items were considered long
term actions. The inspectors verified the licensees Mode 3 corrective actions had been
implemented during periods of continuous control room observation from January 1 to
January 5, 2004, January 8 to January 9, 2004, and January 26 to January 28, 2004.
Specifically, the inspectors verified that:
formal peer checks of TS entries were being performed by a second licensed
individual;
formal peer checks were being logged in the narrative log;
reactor operators were co-authorizing the start of maintenance instructions that
affected TS equipment;
reactor operators were tracking short duration TS actions using an electronic
timer;
that Crew Updates to keep the crew informed of significant changes in plant
status were being performed during transient conditions and normal plant
operations; and
the Operations Superintendent and Operations Manager were spending a
portion of each day monitoring and mentoring shift personnel.
b.
Findings and Observations
There were no findings identified. The inspectors verified that the licensee had an
appropriate schedule for the implementation of the remaining corrective actions.
.4
(Closed) URI 05000346/2002014-06: Question Regarding Licensee Compliance with
Code Relief Valve Requirements.
During November of 2002, the NRC identified a concern for potentially inadequate
over-pressure protection for the containment air coolers (CACs), decay heat removal
Enclosure
17
(DHR) coolers, emergency diesel generator jacket water (EDGJW) heat exchangers
and associated system piping. For example, the NRC had questioned the use of locked
open valves between the relief valve and the Code components requiring relief
protection with respect to meeting the American Society of Mechanical Engineers
(ASME) Code requirements for positive controls and interlocks on stop valves.
On January 23, 2004, the inspectors completed an on-site inspection of this concern
focused on the location of the system relief valves to ensure over-pressure protection
was provided for the CACs, EDGJW heat exchangers and DHR coolers under
operating/design basis conditions. The inspectors discussed the specific requirements
and system configurations associated with over-pressure protection with NRC staff in
the Office of Nuclear Reactor Regulation and no concerns for Code compliance were
identified. Specifically, the inspectors confirmed that:
The EDGJW coolers and CACs were not Code stamped vessels and thus did
not have component level design requirements governing over-pressure
protection. The over-pressure protection for the CACs was provided by pressure
relief devices for the service water system in which the CACs were installed.
The DHR coolers were Code stamped vessels, which had component level
over-pressure protection requirements from the original design Code (ASME
Code,Section III and Section VIII, 1968 Edition). The inspectors confirmed that
the configuration and location of the system over-pressure protection devices
was consistent with these requirements.
For the component cooling water, service water and decay heat removal piping
systems which contained these components, the applicable design Code was the
ASME Code,Section III, 1971 Edition. This design Code contained specific
requirements associated with the location, capacity and types of relief protection
required. The inspectors confirmed that the configuration and location of the
system over-pressure protection devices were consistent with these
requirements for the piping sections containing these components.
For these systems and components, the licensee had not produced a written document
that explicitly identified how the applicable over-pressure protection requirements from
the design Codes were implemented. The inspectors were concerned that without an
explicit written over-pressure protection record, changes in plant operating lineups or
system components could render the Code over-pressure protection strategy ineffective
and result in damaged equipment. Based upon this observation, the licensee
implemented corrective actions (CR 04-00582) to document the over-pressure
protection strategy for these systems and components in controlled safety-related
calculations.
The inspectors did not identify any normal or emergency operating system
configurations or lineups that would result in isolating the CACs, EDG JW coolers and
DHR coolers from over-pressure protection devices, without considering these
components and associated piping systems inoperable. Further, no deviations from
applicable Code requirements were identified with respect to location of relief protection
devices for these components. This URI is closed.
Enclosure
18
.5
(Closed) AV 05000346/2003021-01: Potential Inability for HPI Pumps to Perform Safety
Related Function
This issue is discussed in this report under closure of LER 05000346/2003-002-01. The
NRCs review of the information provided by the licensee and final determination of
significance (White), is documented in NRC Inspection Report 05000346/2004005,
which was issued on March 5, 2004.
.6
Routine Review of Identification and Resolution of Problems
a.
Inspection Scope
From September 8, 2003, through October 10, 2003, and December 16, 2003, in the
Region III Office, the inspectors performed a review of condition reports associated with
LER 50-346/03-001-00/01/02. The inspectors reviewed these reports to confirm that the
licensee had appropriately described the scope of the problems. Additionally, the
inspectors review included confirmation that the licensee had an appropriate threshold
for identifying issues and had implemented effective corrective actions. The inspectors
performed these reviews to ensure compliance with 10 CFR Part 50 Appendix B,
Criterion XVI Corrective Action, requirements. The specific corrective action
documents that were reviewed by the inspectors are listed in the attachment to this
report.
b.
Findings
Introduction: The inspectors identified a Non-Cited Violation (NCV) of 10 CFR Part 50,
Appendix B, Criterion XVI, "Corrective Action," having very low safety significance
(Green) for the licensees failure to determine the cause and implement actions to
prevent recurrence for the inadequate design changes made to service water (SW)
valves SW1424, SW1429 and SW1434.
Description: In LER 50-346/03-001, the licensee identified that under a loss of
instrument air (LOIA), the SW system valves temperature control valves (SW1424,
SW1429 and SW1434) used to throttle SW flow through the component cooling water
(CCW) heat exchangers would not reach their design (full open) position. The licensee
implemented corrective actions to restore the design function of these valves as
discussed in Section 4OA3 of this report. The inspectors identified that the licensee had
incorrectly determined that these inoperable SW valves did not represent a significant
condition adverse to quality in CR 03-04158.
The licensee initially screened CR 03-04158, which documented the inoperable air
operated valves (AOVs) SW1424, SW1429 and SW1434 as a significant condition
adverse to quality. Subsequently, the licensee downgraded this CR to a fix only
(category CF), which did not require determining the causes of this condition. The
licensee based this decision on the assumption that the causes for these SW AOV
deficiencies was the same as that found for inoperable AOVs MU66A-D and MU38
documented in CR 02-02494 and CR 02-02254. For valves MU66A-D and MU38, the
licensee had identified root causes which included, a lack of design basis information
and calculations to support orientation, set-up, and sizing of air operated valves during
Enclosure
19
original plant construction. However, for valves SW1424, SW1429 and SW1434, the
licensee had removed air accumulators for the valve actuators in 1991 under
modification 87-1351, which rendered these valves incapable of reaching their design
open condition with a LOIA. The inspectors noted that the licensees modification
process included additional barriers such as detailed design reviews and post
modification testing to confirm the design functions of these valves. The inspectors
concluded that the inadequate SW modification, represented a separate and distinct
significant condition adverse to quality from that found for other AOVs, because the
CCW and SW systems were rendered inoperable by the inadequate implementation of
the modification process. Because the licensee had not correctly classified
CR 03-04158 as a significant condition adverse to quality, a root cause investigation
was not performed and consequently the cause(s) for the inadequate modification were
not known. Therefore, the inspectors were concerned that the licensee had not
implemented corrective actions to fix the modification process errors that had resulted in
the inadequately designed SW valve modifications. Although the licensee had
implemented corrective measures for the SW valve design deficiencies, the licensee
failed to recognize the need for a root cause investigation and to take preventative
actions for the inadequate modification process until questioned by the NRC inspectors.
Analysis: The performance deficiency associated with this event, is the failure of the
licensee to identify the cause(s) and implement action(s) to prevent recurrence for the
inadequate SW modification 87-1351. The inspectors concluded that the finding was
greater than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power
Reactor Inspections Reports, Appendix B, Issue Disposition Screening. The
inspectors concluded that the issue was more than minor; because if left uncorrected,
the inadequate modification process controls could potentially result in other inoperable
safety related equipment or systems. The finding was assigned to the Mitigating
Systems Cornerstone because the specific example of inadequate modification process
controls was associated with the SW and CCW mitigating systems. The finding also
affected the cross-cutting area of Problem Identification and Resolution because
although the deficient modification 87-1351 was discovered by the licensees staff, it
was not adequately resolved until questioned by the NRC inspectors. The inspectors
determined that the finding could not be evaluated using the Significance Determination
Process (SDP) in accordance with IMC 0609, Significance Determination Process,
because the SDP for the Mitigating Systems Cornerstone only applied to a degraded
systems/components, not to the process failures that could result in degraded
systems/components. Therefore, this finding was reviewed by the Regional Branch
Chief in accordance with IMC 0612, Section 05.04c, who agreed with the inspectors that
this finding was of very low safety significance (Green), because the licensee had taken
actions to restore the air operated SW valves to an operable configuration and, after
identification by the inspectors, the licensee entered the failure to identify the cause(s)
and implement action(s) to prevent recurrence for the inadequate SW modification
87-1351 into the corrective action program (CR 03-07859).
Enforcement: Title 10 CFR Part 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
Enclosure
20
condition is determined and corrective action taken to preclude repetition. The
inspectors considered the inadequate SW modification 87-1351 a significant condition
adverse to quality, because it rendered the component cooling water and service water
systems incapable of meeting their design functions under accident conditions.
Contrary to the above, as of October 9, 2003, the licensee had failed to determine the
causes and implement corrective action to preclude repetition for the inadequate SW
modification 87-1351 installed in 1991. This issue was not an immediate concern,
because the licensee implemented corrective actions to correct the specific design
deficiencies in these valves as discussed in Section 4OA3. Because of the very low
safety significance, and because this issue was entered into the licensees corrective
action program, this violation is being treated as a Non-Cited Violation consistent with
Section VI.A of the NRC Enforcement Policy (NCV 50-346/04-02-03). The licensee
entered this issue into the corrective action program in CR 03-07859.
4OA3 Event Followup (71153)
.1
(Closed) LER 05000346/2003-002, Revision 00 and 01: Potential Degradation of High
Pressure Injection Pumps Due to Debris in Emergency Sump Fluid Post Accident
On October 22, 2002, with the reactor defueled, the licensee identified a deficiency
regarding the internal clearances of the HPI pumps ability to pass debris or particles
that may be entrained in the process fluid during some post accident scenarios.
Specifically, it was determined that the pumps internal openings that supplied
lubricating water flow to the hydrostatic bearing were smaller than the ECCS sump
screen openings. Certain reactor accident scenarios required the HPI pump (via the low
pressure injection pump) to pump water that had collected in the containment ECCS
sump and inject it back into the reactor coolant system (recirculation mode). It was
during this mode of operation that the potential existed for debris from the sump to be
transported to the HPI pump and cause blockage of lubricating water to the hydrostatic
bearing.
The licensee performed extensive analysis, pump modifications, qualification testing,
in-plant testing, and reduction of fibrous insulation in containment to ensure adequate
HPI pump performance during the recirculation mode. The NRR staff from the Division
of Engineering, Mechanical and Civil Engineering Branch and the Division of Systems
Safety and Analysis, Plant Systems Branch performed a review of the analysis, testing
and modifications performed by the licensee and concluded that the licensees overall
approach to the modification of the high pressure injection pumps is acceptable and
provided reasonable assurance that they will perform their required functions when
called upon. This review is detailed in Task Interface Agreement (TIA) 2003-04
response dated February 11, 2004, and is included as an attachment to this report.
With respect to prior plant operation with the potential to degrade the HPI pumps during
the recirculation mode of operation post accident, the licensee provided a response to
NRC Inspection Report 05000346/2003021 on December 5, 2003. This inspection
report had determined that the preliminary significance of this issue was greater than
Green. The NRCs review of the information provided by the licensee and final
determination of significance (White) is documented in NRC Inspection Report 05000346/2004005. This LER is closed.
Enclosure
21
.2
(Closed) Licensee Event Report (LER) 50-346/03-001, Revisions 00, 01, and 02:
Potential Inability of Air-Operated Valves to Function During Design Basis Conditions
a.
Inspection Scope
From September 8, 2003, through October 10, 2003, and from February 23, 2004,
through February 26, 2004, the inspectors reviewed LER 2003-001, which documented
that several AOVs were not capable of performing their designated safety functions for
all required conditions. The inspectors also reviewed corrective actions documented in
condition reports, and design information (including calculations) associated with two
AOVs (CC1467 and CC1469 - decay heat removal heat exchanger outlet valves) which
were removed from this LER in subsequent LER revisions.
b.
Findings
Description: On January 30, 2002, with the reactor defueled, the licensee identified that
several AOVs had negative operating margins and subsequently determined that eight
valves were not capable of performing their safety functions for all required conditions.
On March 31, 2003, the licensee submitted a 10 CFR 50.73 report documenting this
condition as an operation or condition prohibited by the Technical Specifications 3.7.3.1,
3.7.4.1 and 3.6.3.1.
On August 18, 2003, and November 26, 2003, the licensee issued Revision 1 and
Revision 2, respectively to this report to update the report and to document the risk
evaluation of this condition. The licensee described the causes for the inoperable AOVs
as discussed below.
Valve MU3 is an air operated isolation valve which is normally open to allow letdown
flow to pass from the letdown coolers to the purification demineralizers. With LOIA, this
valve is designed to fail closed. However, the licensee identified that with LOIA, the
spring force alone was not sufficient to close valve MU3 against maximum reactor
coolant system differential pressure. The licensee implemented engineering change
request (ECR) 03-0111-00 to replace the valve actuator with a new larger piston
actuator and nitrogen bottles which would ensure this valve shuts with design differential
pressure (CR 03-01040).
For the degraded capability of valve MU3 to close, the licensee determined that spring
relaxation was not recognized by plant personnel for preventative maintenance
purposes.
Valve CC 1495 is an AOV which is normally open to provide cooling water to
non-essential components such as the spent fuel pool heat exchangers or reactor
coolant pump seal return coolers. This valve is designed to close on a safety features
actuation system (SFAS) Level 3 signal or a low level in the CCW surge tank. However,
the licensee identified that upon LOIA, the air accumulator was undersized and would
not ensure that the valve would fully close. The licensee implemented ECR 03-0136-00
to install a larger air accumulator associated with the actuator for valve CC 1495
(CR 03-01253).
Enclosure
22
The licensee attributed the apparent cause of the valve CC 1495 deficiencies to original
procurement practices that resulted in obtaining an undersized air accumulator. The
licensee identified that the air operated valve vendors had been being supplied with
inaccurate system information and had used less than conservative sizing
methodologies.
SW system isolation valves SW1356, SW1357, and SW1358, are normally open to
provide a flow path for SW to the containment air coolers (CACs). During normal and
emergency operation, two of the three CACs are in service and the remaining CAC will
have its SW isolation valve closed to support containment isolation. However, with
LOIA, the air accumulators for these valves did not have sufficient capacity to hold the
valves shut for up to 30 days to support containment isolation. The licensee
implemented ECR 02-0836 to install larger air accumulators (CR 02-07781).
The licensee attributed the valve SW1356, SW1357, and SW1358 deficiencies to a lack
of understanding (during original plant construction) of the plants design and licensing
basis and a failure to correlate this information into the design of the accumulators for
these valves.
SW system valves SW1424, SW1429 and SW1434 are temperature control valves used
to throttle SW flow through the CCW heat exchangers. During emergency operation,
these valves go to their full open position upon receipt of an SFAS Level 2 signal to
maximize SW flow through the CCW heat exchangers. These valves are required to fail
open upon LOIA. These valves have spring air cylinder actuators which require the
presence of air to position the valve. Upon LOIA, the licensee identified that these
valves would not fully open. The licensee performed a dynamic differential pressure test
(without instrument air) for valve SW1434, and this valve opened to 28 degrees from
fully shut and stalled in this position. The licensee initiated ECR 03-0299-00 to install an
air accumulator for each valve to ensure the motive air force exists for valve operation
under accident conditions (CR 03-04158). The licensee identified in this LER that they
had made an incorrect engineering assumption in the design of the actuators for these
valves when the original air accumulators were removed and actuators replaced in 1991
(modification 87-1315). The inspectors identified that the licensee had not correctly
classified this condition as a significant condition adverse to quality and identified root
causes as discussed in Section 4OA2.
The licensee implemented the corrective actions for each of the inoperable valves as
stated above to restore these systems to an operable condition. In the original version
of LER-03-001, the licensee had identified that the decay heat removal heat exchanger
outlet valves CC1467 and CC1469 were not operable because of undersized operators.
The licensee subsequently determined that these valves were operable based upon a
revised calculation C-ME-016.04-035, Component Level Review Calc for AOV
CC1467/1469," which demonstrated that these valves had adequate operating margins
with the existing accumulators. The inspectors reviewed calculation C-ME-016.04-035
and C-ME-016.04-031, Maximum Expected Differential Pressure For Valves CC-1467
and CC-1469," to confirm that the licensee had used industry accepted methodologies
to demonstrate that sufficient operating margins (e.g. to account for uncertainties)
existed for these valves. Because the licensee intended to apply the same approach to
Enclosure
23
demonstrate operating margins for each of the modified AOVs, the inspectors did not
identify any operability concerns for the modified AOVs.
Analysis: The licensee documented the risk for the degraded AOVs as discussed
below.
C
For valve MU3, the licensee determined that the valve would have functioned to
isolate letdown flow under the reduced differential pressure which would exist
under accident conditions. Additionally, a motor operated valve MU2A existed
which was fully functional and would have isolated letdown flow under accident
conditions.
C
For valve CC1495, the licensee determined that the valve would not have fully
closed to isolate nonessential component cooling water loads. However, motor
operated valves CC5096 and CC5097 were fully functional and would have
automatically isolated the nonessential cooling loads on a low CCW surge tank
level.
C
The licensee determined that valves SW1356, SW1357, and SW1358, would not
have been capable of maintaining containment isolation capability. Because this
SW piping is a closed fluid filled system inside containment the licensee
determined that this condition would not create a release pathway for
post-accident radioactive material.
C
The licensee determined that SW1424, SW1429 and SW1434, would not have
been capable of stroking to their full open position upon receipt of an SFAS
Level 2 signal with loss of instrument air. This degraded SW flow condition was
evaluated as part of the licensees risk evaluation and determined to have
minimal safety significance.
The licensee also performed a calculation to determine the increase in core damage
frequency, core damage probability, large early release frequency and large early
release probability due to the condition described in LER 2003-001. Based upon the
results of this calculation, the licensee determined that these valve conditions were
considered to have minimal safety significance. The licensees calculation and risk
evaluation were reviewed by the inspectors and a Region III Senior Reactor Analyst
(SRA).
Although a number of AOVs were affected, the inspectors determined that the degraded
CCW heat exchanger service water outlet valves SW1424, SW1429, and SW1434
posed the most significant challenge to risk. For these valves, the licensee had
removed the air accumulators that assist in valve stroking, and then determined through
a dynamic differential pressure test performed on SW1434, that the spring operator
force alone was not adequate to fully open the valve. The failure position of these
valves during a LOIA initiating event was potentially insufficient to assure adequate SW
cooling through the decay heat coolers to support cooling of safety related equipment
needed to mitigate a loss of coolant accident (LOCA). Specifically, the heat loads on
the CCW system, could exceed the reduced cooling capacity of the decay heat coolers
Enclosure
24
when sump recirculation is initiated after depletion of the borated water storage tank
(BWST) inventory following a LOCA.
The inspectors determined that this issue was more than minor because it affected the
ability of mitigating systems to perform their function during certain accident scenarios
(e.g. mitigating systems cornerstone objective was challenged). Because this issue
represented an actual loss of a safety function of a system, the inspectors performed a
Phase 2 risk screening using the Davis-Besse site-specific SDP worksheets. The
inspectors concluded that the LOIA was the only initiating event where a potential for a
significant contribution to core damage frequency (CDF) occurs, because a reactor
coolant pump (RCP) seal LOCA can be caused by a LOIA. For a RCP seal LOCA (very
small LOCA), the licensees calculations determined that the time required to empty the
BWST was approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. For this sequence, the licensee determined that
prior to the depletion of the BWST, the heat load from the decay heat coolers would not
be imposed on the CCW system. Therefore, the licensee concluded that the CCW
system cooling was adequate for the injection phase on a RCP seal LOCA, and ample
time was available for operator action to fully restore CCW cooling by fully opening the
SW outlet valves (per existing procedures).
For the Phase 2 evaluation of this issue, the inspectors only considered cutsets
involving high pressure recirculation because of the long period of time before BWST
depletion. Using the LOIA worksheet and considering the two high pressure
recirculation sequences with 1 point credit (10-1) for operator recovery, the inspectors
determined that the resulting risk (change in CDF) was 10-5. The inspectors considered
that this risk estimate was very conservative due to the ample time (15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />) that
operators would have to recover the full capability of the CCW heat exchangers by fully
opening the SW outlet valves.
A Region III SRA completed a Phase 3 risk evaluation focused on the recovery credit for
restoration of CCW under the conditions discussed above. The Region III SRA
reviewed the licensees human reliability analysis (HRA) for this evolution and also
performed an independent HRA calculation with the methodology utilized in the SPAR
model. When an SFAS Level 2 signal is received following the initiation of a LOCA, the
operators are required to verify the status of the SW valves to the CCW heat exchanger.
Indications available to the operators include the indicating light for this valve which for
this scenario, would have been off, instead of lit. With insufficient SW flow, the CCW
heat exchanger outlet temperature alarm would have come in at 115/F, which would
require an operator to implement procedure DB-OP-02523, Component Cooling Water
System Malfunctions, and take actions to restore the system. Even though the BWST
would not be depleted until 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, the licensee assumed that the SW valves would
need to be opened within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The SRA agreed with this conservatism. The SRA
used the SPAR model human error worksheet for recovery of the CCW system with the
following assumptions: 1) the operators would have had more than enough time to
restore the CCW system; and 2) stress, complexity, experience, procedures, fitness for
duty, and work processes were set to the nominal value. This resulted in a CCW
system non-recovery probability of 10-3, or 3 points in the SDP LOIA worksheet. With
the more realistic credit given for operator action, the SDP worksheets indicated that the
risk characterization was 7 (10-7) or Green. Because the CCW system does not provide
cooling to important containment cooling systems, including the CACs and containment
Enclosure
25
spray systems, the SRA determined that this issue did not play an important role in the
large early release frequency and would not cause the risk to rise above the Green
threshold. Therefore, the SRA determined that this issue is characterized as having
very low risk significance.
Enforcement: This licensee-identified finding involved a violation of TS 3.7.3.1 (CCW
system was required to be operable), TS 3.7.4.1 (SW system was required to be
operable), and TS 3.6.3.1(containment isolation valves were required to be operable).
However, based upon calculations performed by the licensee for the degraded CCW
system performance and the availability of redundant safety equipment (e.g. other motor
operated valves), the violations of TS 3.7.3.1, TS 3.7.4.1 and TS 3.6.3.1 are not more
than very low safety significance. Therefore, the inspectors determined that this
licensee-identified finding met the criteria of Section VI of the NRC Enforcement Policy,
NUREG-1600, for being dispositioned as NCVs. This finding was determined to be of
very low safety significance and is dispositioned in Section 4OA7 of this report. This
LER is closed.
.3
(Closed) Licensee Event Report (LER) 50-346/2003-05; Revisions 00 and 01:
Containment Gas Analyzer Heat Exchanger Valves Found Closed Rendering the
Containment Gas Analyzer Inoperable.
The licensees initial submittal of this LER discussed a condition where the
component cooling water isolation valves on the inlet and outlet to the heat
exchangers located in each of the two Containment Gas Analyzers Systems
(CGAS) were found stuck shut. This resulted in the CGAS being inoperable. This issue
was evaluated by the inspectors and documented in Inspection Report 50-346/03-017.
Supplement 01 to this LER, dated January 23, 2004, described the following two
additional issues that were identified by the licensee during the extent of condition
evaluation, that directly impacted the proper operation of the hydrogen gas analyzers:
The instrument air supplied to the moisture trap drain check valve associated
with the hydrogen analyzers heat exchanger was non-safety grade. Post
accident, no credit can be taken for this air supply. Therefore, the drain valve is
assumed to not have functioned. Additionally, the regulator which supplied the
air to the drain valve was set too low for the drain valve to operate as designed.
The moisture traps potentially contaminated condensate, via the drain valve,
would flow to a floor drain in a room not served by the emergency ventilation
system. This constituted a potential containment bypass pathway.
The licensee addressed the first issue by eliminating the reliance of the drain check
valve on instrument air by replacing the air operated drain valves with solenoid operated
valves, powered by independent essential power sources. The second issue was
corrected by routing the potentially contaminated condensate to an existing ECCS floor
drain.
The inspectors determined that the improper application of non-safety related instrument
air to the containment gas analyzers was a licensee identified violation of 10 CFR 50,
Appendix B, Criterion III, Design Control. This finding is unrelated to SSCs that are
Enclosure
26
needed to prevent accidents from leading to core damage. To determine if this finding
had an effect on large early release frequency (LERF), the inspectors used MC 0609,
"Significance Determination Process," Appendix H, Containment SDP. The finding is
characterized as a Type B finding (having no impact on core damage frequency (CDF))
and compared to Table 3 in Appendix H. The inspectors determined that the hydrogen
analyzer had no impact on the containment-related SSCs listed in Table 3
(i.e. containment penetration seals, containment isolation valves or purge and vent lines)
and would not influence LERF. Based on this, the finding has very low safety
significance.
The inspectors determined that the potential containment bypass pathway caused by
the improper trap condensate drain path was a licensee identified minor violation of
10 CFR 50, Appendix B, Criterion III, Design Control. This issue was determined to be
of minor significance because there was no evidence that the trap had ever functioned,
and since the air supplied to the trap was of insufficient pressure for operation of the
trap, it was highly unlikely that the trap would have ever functioned.
Based on the inspectors review of the LER and the licensees corrective actions to
address the design deficiencies, this LER is closed.
.4
(Closed) Licensee Event Report (LER) 50-346/2003-006: Potential Errors in Analysis of
Block Walls Regarding HELB Differential Pressure and Seismic Events
This event report documented that on May 21, 2003, the licensee, while reviewing their
existing structural analyses for walls within the Auxiliary Building, identified via
calculation, that masonry Wall 2257 would not remain operable when subjected to
compartment pressurization loads from a high energy line break (HELB) concurrent with
loads from a design seismic event. Wall 2257 forms the boundary between Room 241
(a passageway on the 565 foot elevation of the Auxiliary Building) and Room 240 (Boric
Acid Addition Tanks). The licensee has determined that failure of the block wall could
adversely affect Component Cooling Water Auxiliaries Return Isolation Valves, a
Service Water supply line to Containment Air Cooler 1, and the functioning of Boric Acid
Addition equipment within Room 240.
The licensee has determined that the structural analyses done in the 1980's in response
to Generic Letter 80-11 [Masonry Wall Design] incorrectly modeled that a high energy
line break in a room adjacent to Room 241 would not impact Room 241, although the
rooms are open to each other. The licensee determined the misapplication of the HELB
model was a result of a lack of communication between design engineering groups
performing different sets of calculations and inadequate development and maintenance
of calculations in the 1980s.
In August 2003, the licensee modified a door to Room 240 so that, in the event of
differential pressure across the wall and door, as would be seen with a postulated
HELB, the door would open to reduce differential pressure across Wall 2257. This
venting of pressure by the door reduced the calculated loads on Wall 2257 during a
HELB and concurrent seismic event and the wall would remain operable.
Enclosure
27
The inspectors determined that failure to provide measures for controlling design
interfaces and for coordination among participating design organizations was a licensee
identified violation of 10 CFR 50, Appendix B, Criterion III, Design Control. This issue
was determined to be of more than minor significance because the calculation errors
were significant enough that a modification was required to resolve the wall loading
issues. The finding affects the Mitigation Systems Cornerstone and was considered to
have a very low safety significance (Green) using Appendix A, Attachment 1 of the
Significance Determination Process because the event does not involve the total loss of
any safety function as identified by the licensee. This issue was discussed further in
Section 4OA7 of this report.
Based on the inspectors review of the LER, a physical review of the modification and
location of equipment within Room 280, and a review of the licensees corrective actions
including the Root Cause for the event, this LER is closed.
.5
(Closed) Licensee Event Report (LER) 50-346/2003-013: Unplanned Reactor Trip
Condition with the Reactor Shutdown
On September 30, 2003, with the plant in Mode 3, following the completion of a normal
operating pressure test using non-nuclear heat, an unplanned reactor trip occurred due
to a valid Shutdown Bypass High Pressure signal. This signal inserted control rod
Group 1 while the operators were performing a plant cooldown. The Reactor Protection
System and all components responded as intended. There were no post-trip response
equipment issues identified. The licensee attributed the reactor trip to less than
adequate operator performance, procedural guidance, and pre-job brief. Corrective
actions included: providing more prescriptive guidance in the plant shutdown and
cooldown procedure, the incorporation of a requirement to conduct a pre-job brief which
include specific operating experience gained from this event, discussion of the event
with on-shift operating personnel, developing a case study of the event for initial and
continuing training, and developing an Operations Improvement Action Plan. The
licensee documented the unplanned reactor trip in CR 03-08374. A Non-Cited Violation
for this issue was discussed in Inspection Report 05000346/2003022
(NCV 05000346/2003022-01). The LER was reviewed by the inspectors and no
additional findings of significance were identified. This LER is closed.
.6
Turbine Building Door High Energy Line Break Non-Emergency Report
a.
Inspection Scope
On January 19, 2004, the licensee reported, in accordance with
10 CFR 50.72(b)(3)(ii)(B), a design issue associated with the capabilities of certain
turbine building doors. The licensee discovered that certain doors may not be able to
withstand the initial pressure wave caused by a guillotine break of a main steam line in
the turbine building. The licensees preliminary analysis indicated that the initial
pressure wave could cause the failure of the doors leading to both trains of low voltage
switchgear, and the resultant steam environment could potentially render all low voltage
AC equipment and station batteries inoperable.
Enclosure
28
The inspectors evaluated the impact of this design issue on the current plant conditions,
and the licensee plans to re-enforce 10 turbine building doors prior to exceeding steam
pressures that could potentially challenge the existing door structures. This issue was
entered into the licensees corrective action program (CR 04-00402, 04-00442,
04-00478, and 04-00512).
b.
Findings
No findings of significance were identified. Adequacy of corrective actions, impact of
past system inoperability, and potential enforcement actions for the design deficiency
were planned to be assessed subsequent to the licensees LER submittal.
4OA5 Other Activities
A.
Evaluation of Restart Issues
One of the key building blocks in the licensees Return to Service Plan was the
Management and Human Performance Excellence Plan. The purpose of this plan was
to address the fact that management ineffectively implemented processes, and thus
failed to detect and address plant problems as opportunities arose. The primary
management contributors to this failure were grouped into the following areas:
Nuclear Safety Culture;
Management/Personnel Development;
Standards and Decision-Making;
Oversight and Assessments; and
Program/Corrective Action/Procedure Compliance.
The inspectors had the opportunity to observe the day-to-day implementation that the
licensee made toward completing Return to Service Plan activities. Almost every
inspection activity performed by the resident inspectors touched upon one of those five
areas. Observations made by the resident inspectors were routinely discussed with the
Davis-Besse Oversight Panel members and were used, in part, to gauge licensees
efforts to improve their performance in these areas on a day-to-day basis.
To better facilitate the inspection and documentation of issues not specifically covered
by existing inspection procedures, but important to the evaluation of the licensees
readiness for restart, the Special Inspection for Residents inspection plan was
developed and implemented. Inspection Procedure 93812, Special Inspection, was
used as a guideline to document these issues and remains in effect for future resident
inspection reports until a time to be determined by the Davis-Besse Oversight Panel.
The inspectors performed inspections, as required, to adequately assess licensee
performance and readiness for restart in the following areas:
performance of plant activities, including maintenance activities;
follow-up of specific Oversight Panel Technical issues;
licensee performance during restart readiness meetings;
licensee performance in categorizing, classifying, and correcting deficient plant
conditions during the restart process;
Enclosure
29
licensee performance at meetings associated with work backlogs, including the
deferral of work orders, operator workarounds, temporary modifications, and
permanent modifications; and
activities associated with safety conscious work environment and safety culture.
The following issues were evaluated during this inspection period.
.1
Observation of Licensee Performance During Plant Heatup and Cooldown
a.
Scope
The inspectors continuously observed operation department personnel performance
significant operations evolutions. The inspectors focused on control room observations,
but also included evaluation of shift turnovers, scheduling meetings, pre-job briefs, and
plant lineups. Observations, while primarily in the control room, included tours of the
auxiliary building and containment and attendance at scheduled planning and trouble
shooting meetings. Time periods of continuous observations included:
January 2 through January 5, 2004, [completion of plant heatup to normal
operating pressure];
January 8, 2004, [plant cooldown to Mode 4 to repair an auxiliary feedwater
pump]; and
January 26 through January 28, 2004, [plant heatup to normal operating
pressure and temperature].
b.
Observations
The inspectors had the following observations:
alarm response procedure usage had improved and was more consistent;
mode constraints were properly addressed prior to applicable Mode ascension;
pre-job briefings were appropriately detailed and the licensee effectively utilized
reverse briefing techniques during pre-evolution briefs;
licensee
response to
developing
equipment
issues was
observed to
be
appropriate;
operations department personnel received appropriate technical support to follow
and investigate emergent equipment issues; and
emergent equipment issues were properly investigated using formal licensee
processes.
c.
Conclusions
The inspectors identified no findings of significance. These examples illustrated
improved performance by Operations in the areas observed.
Enclosure
30
Enclosure
31
.2
Non-Compliance with TS Action Statement
a.
Inspection Scope
The inspectors reviewed licensee response to an error in maintaining control of the
status of a Steam Feedwater Rupture Control System (SFRCS) steam generator
differential pressure switch which resulted in an inadvertent non-compliance with a
TS action statement.
b.
Findings
Introduction: A Non-Cited Violation of TSs, having very low safety significance was
self-revealed when, during performance of a functional test on the Steam Feedwater
Rupture Control System (SFRCS) steam generator differential pressure switch, the
licensee isolated the pressure switch and maintained that isolation for a period of
approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 minutes without the control room personnel knowing that
the pressure switch was isolated for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This rendered the pressure
switch incapable of sensing differential pressure and providing a signal, if needed, to the
SFRCS actuation channel 2. Failure to maintain the proper status of TS equipment was
a violation of plant procedures required by TS 6.8.1., Procedures and Programs.
Description: On January 6, 2004, while in Mode 3, the licensee was performing a
Channel Functional Test of Steam Feedwater Rupture Control System Channel 4
Steam Generator Differential Pressure Switch for Steam Generator 2 per DB-MI-0324.
Channel 4 provides an input to SFRCS actuation channel 2. As required by the
procedure the switch was isolated and an entry noting the removal from service at
1621 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.167905e-4 months <br /> was made in the Unit Log. The Unit Log entry stated that entered
T.S. 3.3.2.2 action 16 . . . . Action statement 16 of that specification requires that
action be taken within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the inoperable channel in a tripped condition or
return the switch to service. However, the technicians performing the test determined
that leaking isolation valves precluded them from performing the functional test. The
technicians left the pressure switch isolated and started discussions on appropriate
further action. Those discussions included work group supervisors and operations
personnel. Those discussions did include the need to restore the pressure switch, but
time frame for action was not made clear. At approximately 1845 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.020225e-4 months <br />, the pressure
switch was returned to service. For approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 minutes, the pressure
switch was isolated and would not perform its function and the associated instrument
channel was not placed into a tripped condition.
Analysis: The finding was determined to be more than minor in accordance with
IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition
Screening, issued on April 29, 2002. The finding: (1) involved the configuration control
and human performance attributes of the Mitigating Systems Cornerstone; and
(2) affected the cornerstone objective to ensure the availability, reliability, and capability
of systems that respond to initiating events to prevent undesirable consequences.
In accordance with IMC 0609, Significance Determination Process, Appendix A,
Attachment 1, dated March 18, 2002, the inspectors performed an SDP Phase 1
screening and determined that the issue affected the Mitigation Systems Cornerstone in
Enclosure
32
that SFRCS is used to ensure sufficient removal of core decay heat in the event of
various accident conditions including a steam generator rupture. This finding was of
very low safety significance because at the time of the occurrence, the reactor was in
Mode 3 with no substantial decay heat and one complete actuation train of SFRCS
remained operable.
Enforcement: The inspectors concluded that this is a performance issue because
maintaining knowledge of system configuration and ensuring control of system
configuration was reasonably within the licensees ability to control and the event could
have been prevented. The performance deficiency associated with this event is the
control room staff did not adequately monitor and control system status which resulted
in an unanticipated entry into a TS action statement requirement. Technical Specification 6.8.1.a requires implementation of procedures recommended by
Regulatory Guide 1.33. Regulatory Guide 1.33 lists Administrative Procedures which
address authorities and responsibilities for safe operation and shutdown. The licensee
developed DB-OP-00000, Conduct of Operations, Revision 07, a safety-related
procedure, to, in part, provide guidance on how Operations personnel carry out their
duties and responsibilities as delineated in Station Procedures, Policies, Directives, and
Manuals. Step 6.2.1 of DB-OP-00000 states Operations Personnel . . . shall be
responsible for monitoring the equipment, instrumentation and controls within their area
and taking timely and proper action to ensure safe, conservative operation of the unit.
Contrary to those requirements, the channel 4 differential pressure switch was isolated
and could not perform its function for a period of approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 24 minutes,
which was in excess of the time period that was specified in the TSs and in excess of
the time period that the licensee had planned, without the knowledge of the operating
control room crew and without the crew having the associated channel placed into a
tripped condition. Because of the very low safety significance and because the issue
has been entered into the licensees corrective action program (CR 04-00181), it is
being treated as a Non-Cited Violation, consistent with Section VI.A of the NRC
Enforcement Policy (NCV 05000346/2004002-04).
.3
Procedure Specified Heater not Available for EDG 2 Room
On January 25, 2004, the control room received unexpected annunciator alarm 1-1-K
EDG 2 TRBL. The alarm was due to a low room temperature of 62°F in EDG 2 room.
Supplementary actions in DB-OP-02037, Emergency Diesel Generator Alarm
Panel 37 Annunciators, Revision 02 step 3.3 states in part If the Shift Supervisor
requests the use of a portable temporary heater, then....obtain a quartz-type heater.
The shift engineer was tasked with determining the availability of a quartz type heater
and contacting the appropriate personnel to have them installed. The licensee could not
locate any quartz type heaters onsite.
The inspectors determined this to be a minor violation of TS 6.8.1.a which requires that
written procedures be established and implemented for alarm conditions. Contrary to
the procedural requirements of DB-OP-02037, the equipment required to respond to the
alarm condition was not available. This was not an immediate safety issue and was
determined to be of minor safety significance because the operators were able to install
a different type of temporary heater which was able to raise the room temperature and
clear the alarm. This issue was not subject to enforcement action in accordance with
Enclosure
33
Section IV of the NRCs Enforcement Policy. The licensee documented the issue in
their corrective action program (CR 04-00652).
.4
Classification, Categorization, and Resolution of Restart Related Issues
The resident inspectors continued to monitor the licensees activity related to properly
classifying, categorizing and resolving their backlog of work orders, corrective actions,
and modifications required to be completed prior to transitioning to Mode 4. To
accomplish this, the inspectors:
attended and assessed licensee management meetings;
monitored the management of open Mode 2 and 1 restraints;
evaluated the licensee classification of emergent deficient conditions; and
evaluated closed Mode restraints.
As part of this inspection, the inspectors attended selected Mode Change Readiness
Review meetings, and Senior Leadership Team meetings where classification of
condition reports, prioritization of work activities, and setting of work completion dates
took place.
The inspectors attended several Plant Support Center Meetings. The purpose of these
meetings was to status significant restart equipment issues and focus licensee
resources to efficiently and effectively work activities to provide more realistic work
completion schedules.
The inspectors attended various work planning meetings. During the meetings there
were discussions among the planners, workers, and management on the approaches
needed to correct equipment issues.
No findings of significance were identified.
.5
Closure of Restart Checklist Items
The Davis-Besse Oversight Panel met to review the following two Restart Checklist
Items and approved their closure:
a.
Restart Checklist Item 2.a: Reactor Pressure Vessel Head Replacement
Inspection Procedure 71007, Reactor Vessel Head Replacement Inspection, provided
guidance on the recommended inspection activities to be conducted when a reactor
vessel head is replaced. The scope of the inspection activities usually included: design
and planning, reactor vessel head fabrication, reactor vessel head removal and
replacement, and post-installation testing. The inspection activities also included
evaluation of the temporary containment access opening and subsequent restoration
following head replacement.
Inspection activities were performed to evaluate the licensees performance in the
following areas.
Enclosure
34
Design and Planning/Reactor Vessel Head Inspection
Inspection Report 05000-46/2002007 documented review of the non-destructive
examinations performed on the replacement head welds that occurred at the Midland
Michigan site and the American Society of Mechanical Engineers (ASME) Code data
packages for the replacement head. Based on our inspection, we confirmed that
adequate records were assembled to ensure that the replacement head was designed
and fabricated in conformance with ASME Code requirements and that the original
ASME Code Section III N-stamp remained valid.
Reactor Vessel Removal and Replacement
The physical removal of the old reactor vessel head from containment and the
movement of the new reactor vessel head into containment were observed as routine
resident plant status activities and were not specifically documented in an inspection
report.
Inspection Reports 05000346/2002010 and 05000346/2003017 documented
radiological inspections associated with head replacement activities. Specific inspection
activities included:
walkdowns of selected portions of the radiologically restricted area, including
areas within the Auxiliary and Containment Buildings where significant
radiological work involving the reactor head and containment breach was
occurring;
observed work occurring both inside and outside of the Containment Building
including preparation for the reactor head moves and Containment Building
breach;
walkdowns of areas outside of the Containment Building where equipment for
making the Containment breach was operating to verify that controls for
containing radioactive materials generated in the breach process were adequate;
reviewed the reactor head encapsulation process to verify that contamination
control and radiological shielding were adequate to minimize dose to workers
and to meet 10 CFR and 49 CFR requirements for the eventual transportation of
the reactor head to a burial site; and
observed aspects of the preparation of a shipment of the reactor head including
the shipping documentation.
Containment Vessel Restoration
Inspection Report 05000346/2002007 documented that:
the engineering evaluation associated with construction of the temporary
containment access opening considered appropriate loads and demonstrated
that stress in the containment shell materials would not exceed design limits;
the temporary containment vessel opening was restored such that the original
ASME Code construction requirements were maintained;
the work activities to construct and restore the temporary containment opening
and closure occurred in a controlled manner and in accordance with procedure
requirements; and
that the licensee managers demonstrated an active oversight role for the control
of the contractors on the containment building temporary construction opening.
35
Enclosure
Inspection Report 05000346/2003005 documented that:
based on the results of the containment integrated leak rate check, containment
integrity had been restored where the containment had been opened for
replacement of the reactor head.
Based on the results of these two inspection activities, the licensees efforts to construct
a temporary containment access, restoration of the temporary access following reactor
head movement into containment, and subsequent leak testing were adequate.
Post Installation Testing
Inspection Report 05000346/2003023 documented inspection during reactor coolant
system leak testing activities. The inspection included walkdowns of the reactor coolant
system while at normal operating pressure as well as detailed evaluation of your
inspections of the reactor vessel bottom head and closure head penetrations, and
control rod drive mechanism flange connections following the 7 day pressure holding
period. As a result of these pressure test activities, we have reasonable assurance that
there are no pressure boundary leaks in the reactor coolant system.
Inspection Report 05000346/2004002 documented inspection of DB-SC-03270, Control
Rod Assembly Insertion Time Test. This activity was observed to evaluate proper
control rod movement and reactor vessel head alignment. This test was successfully
completed on February 10, 2004.
Conclusion
Based on the completion of the inspection activities described in this section, the
inspectors concluded that sufficient basis existed for the closure of this checklist item.
This was checklist item approved for closure by the Oversight Panel on
February 10, 2004.
b.
Restart Checklist Item 2.e (High Pressure Injection Pump Internal Clearance / Debris
Resolution
This issue is discussed in this report under closure of LER 05000346/2003-002-01. The
NRC staff evaluated the HPI pump modifications performed to address concerns
identified by the licensee associated with the potential for debris to damage the pump
during recirculation phase operation. The NRC review included evaluating the validity of
the licensees mock-up tests approach, determining whether the testing demonstrated
acceptable pump performance under design-basis conditions. The NRC staff concluded
in the TIA 2003-04 response dated February 11, 2004, that the licensees overall
approach to the modification of its HPI pumps and its testing, is acceptable and provided
reasonable assurance that the HPI pumps will perform their required functions when
called upon. See attached Task Interface Agreement 2003-04, Evaluation of
Davis-Besse Modifications to the High Pressure Injection Pump and Associated
Mock-up Testing for details of the staffs evaluation. In addition, the NRC staff
evaluated a concern with the adequacy of the minimum flow capability provided for the
pumps to prevent pump failure if the pumps were operated when no injection was
occurring (NRC Inspection Report 05000346/2003010). NRC inspectors agreed with
Enclosure
36
the licensees determination that the pumps would be able to perform their safety
function at the minimum flow of 53 gallons per minute based on the results of a
minimum flow test performed with one of the pumps.
On February 19, 2004, the Davis-Besse Oversight Panel met to discuss this issue and
concluded that Restart Checklist Item 2.e is closed.
.6
Performance of Technical Instruction 2515/154
Spent Fuel Material Control and Accounting at Nuclear Power Plants (TI 2515/154)
a.
Inspection Scope
The inspectors, using the guidance contained in TI 2515/153, interviewed licensee
personnel and reviewed spent fuel pool records to determine whether or not the licensee
had ever removed irradiated fuel rods (pins) from a fuel assembly or reconstituted fuel
assemblies. Based on the results of the interviews and reviews, the inspectors
conducted additional interviews and reviews to gather general information concerning
the licensees Material Control and Accounting program.
b.
Findings
The licensee first began removing irradiated fuel rods from assemblies and
reconstituting assemblies in 1991. These activities were controlled according to vendor
procedures prior to the implementation of the licensees procedure, DB-NE-00100, Fuel
Handling Administration on September 30, 1992. The licensees program tracks
individual fuel rods from the point of removal from a fuel assembly to where they are
stored in the spent fuel pool. All of these removed fuel rods are contained in a failed
fuel basket assembly which is stored in the spent fuel racks. The failed fuel basket
resembles a fuel assembly; however, it does not contain a top nozzle assembly and
therefore its weight is much lighter than a normal fuel assembly. Spent fuel rods and
the failed fuel assembly basket are physically separated from non-fuel components as to
provide reasonable assurance that fuel and non-fuel items are not mistaken for each
other.
The licensees Material Control and Accounting procedures are approved by the plant
manager and controlled in accordance with their quality assurance program. The roles
and responsibilities for all Material Control and Accounting activities are defined in
procedure DB-NE-00100, Fuel Handling Administration. The licensees written
procedures for the movement of individual spent fuel rods within the spent fuel pool
incorporates by reference the fuel vendors procedures. The organization responsible
for documenting and maintaining records of discrete activities within the spent fuel pool
is the Nuclear Fuels Department which has oversight of all spent fuel pool operations.
The Nuclear Fuels Department maintains records documenting all spent fuel pool
operations conducted by contractors and/or fuel vendors and performs an annual
physical inventory of the spent fuel pool that includes resolution of all discrepancies.
Enclosure
37
4OA6 Meetings
.1
Resident Inspector Exit Meetings
The inspectors presented the inspection results to Mr. L. Myers, and other members of
licensee management on February 25, 2004. The licensee acknowledged the findings
presented. No proprietary information was identified.
.2
Interim Exit Meetings
Interim exit meetings conducted:
The closure of URI 50-346/02-014-06, related to relief valves, was discussed
with Mr. B. Allen on January 23, 2004
The inspection to review corrective actions for Licensee Event Report
50-346/03-001, related to potential inability of air-operated valves to function
properly, was discussed with Mr. R. Schrauder on February 26, 2004, by
telephone.
The emergency preparedness program and performance indicators inspection
results were discussed with Mr. L. Myers on February 13, 2004. A second exit
meeting was conducted on the Emergency Action Levels with Mr. J. Vetter on
February 27, 2004, by telephone.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by the
licensee and are violations of NRC requirements which meet the criteria of Section VI of
the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs.
Cornerstone: Mitigation Systems
10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures
shall be established for the identification and control of design interfaces and for
coordination among participating organizations. Contrary to this requirement, on
May 21, 2003, the licensee identified a condition in which calculation and
modeling errors occurred in the 1980s because of inadequate control of and
coordination among various design organizations. Those errors resulted in a masonry
wall within the auxiliary building whose design was insufficient to withstand concurrent
HELB and seismic loading and which was not corrected until August 2003. The failure
of this wall would potentially adversely affect components of TS required systems.
Technical Specification 3.7.3.1 required the CCW system to be operable. Contrary to
the above, the licensee operated since at least 1991 with this system not in an
operable condition due to errors associated with installation of air operated valves
(Section 4.0.A.3). At the time of discovery there were no applicable TS operability
requirements for the affected systems with the reactor defueled. The licensee entered
these issues into its corrective action program as PCAQR 97-1082, CR 02-07781,
Enclosure
38
03-01253, 03-05628, 99-2111, 03-00830, 03-04158, 02-07750, 03-01040, and
CR 03-04878. The licensee performed additional calculations and relied on redundant
safety equipment to demonstrate that this violation is not of more than very low safety
significance.
Technical Specification 3.7.4.1 required the SW system to be operable. Contrary to the
above, the licensee operated since original plant construction with this system not in an
operable condition due to errors associated with installation of air operated valves
(Section 4.0.A.3). The licensee entered these issues into its corrective action program
as PCAQR 97-1082, CR 02-07781, 03-01253, 03-05628, 99-2111, 03-00830, 03-04158,
02-07750, 03-01040, and CR 03-04878. The licensee performed calculations and relied
on redundant safety equipment to demonstrate that this violation is not of more than
very low safety significance.
Technical Specification 3.6.3.1 required the containment isolation valves to be operable.
Contrary to the above, the licensee operated since original plant construction with
nonoperable containment isolation valves condition due to errors associated with
installation of air operated valves (Section 4.0.A.3). At the time of discovery there were
no applicable TS operability requirements for the affected systems with the reactor
defueled. The licensee entered these issues into its corrective action program as
PCAQR 97-1082, CR 02-07781, 03-01253, 03-05628, 99-2111, 03-00830, 03-04158,
02-07750, 03-01040, and CR 03-04878. The licensee relied on redundant safety
equipment to demonstrate that this violation is not of more than very low safety
significance.
Cornerstone: Barriers
10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design
control measures shall provide for verifying or checking the adequacy of design, such as
by the performance of design reviews, by the use of alternate or simplified calculational
methods, or by the performance of a suitable test program. Contrary to this
requirement, LER 2003-05-01 identified a condition in which non-safety related
instrument air was supplied at an insufficient pressure to the Containment Gas
Analyzers. This inadequate design would have prevented the operation of the moisture
trap drain check valves, during system operation post-accident, resulting in the eventual
flooding of the hydrogen analyzer.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
1
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
M. Bezilla, Site Vice President
K. Byrd, Engineering
G. Dunn, Manager, Regulatory Affairs
J. Grabnar, Manager, Design Engineering
B. Henessy, Performance Engineering
L. Myers, Chief Operating Officer, FENOC
K. Ostrowski, Manager, Plant Operations
J. Powers, Director, Nuclear Engineering
R. Schrauder, Director, Support Services
M. Stevens, Director, Maintenance
S. Cope, Senior Emergency Planning Specialist
P. Smith, Emergency Planning Specialist
J. Vetter, Emergency Planning Supervisor
Ohio Emergency Management Agency
E. Edwards, Emergency Planning Specialist
Attachment
2
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
50-346/2004002-01
Licensee not adequately prepared for the onset of frazil ice
conditions (Section 1R01.1)
50-346/2004002-02
Change to Emergency Plan without prior NRC approval
(Section 1EP4)
50-346/2004002-03
Failure to determine the cause and implement actions to
prevent recurrence for the inadequate design changes
made to the service water system valves at the outlet of
the component cooling water heat exchangers
(Section 4AO2)
50-346/2004002-04
Control room staff did not adequately monitor and control
system status which resulted in a noncompliance with a TS
Action Statement (Section 4OA5.2)
Closed
50-346/2003-002;
Revisions 00 and 01
LER
Potential Degradation of High Pressure Injection Pumps
Due to Debris in Emergency Sump Fluid Post Accident
50-346/2003-001;
Revisions 00, 01, 02
LER
Potential Inability of Air-Operated Valves to Function
During Design Basis Conditions
50-346/2003-005;
Revisions 00 and 01
LER
Containment Gas Analyzer Heat Exchanger Valves Found
Closed Rendering the Containment Gas Analyzer
50-346/2003-006
LER
Potential Errors in Analysis of Block Walls Regarding
HELB Differential Pressure and Seismic Events
50-346/2003-013
LER
Unplanned Reactor Trip Condition with the Reactor
Shutdown
50-346/02-014-06
Question Regarding Licensee Compliance with Code
Relief Valve Requirements.
50-346/03-021-01
Potential Inability for HPI Pumps to Perform Safety Related
Function
Attachment
3
LIST OF DOCUMENTS REVIEWED
1R01
Adverse Weather Protection
DB-OP-06931; Seasonal Plant Preparation Checklist; Revision 07
CR 04-00179; Lack of Preparation for Frazil Ice Conditions
CR 04-00242; Intake Crib Mod to Prevent Ice Blockage May Ineffective
CR 04-00652; Procedure Specified Heater is not on Site
1R05
Fire Protection
Fire Hazards Analysis Report
Drawing A-224F; Fire Protection General Floor Plan Elevation 603' 0"
DB-FP-00009; Fire Protection Impairment and Fire Watch; Revision 05
1R13
Maintenance Risk and Emergent Work
Problem Solving Plan for RCP 2-2 High Seal Return Flow; January 4, 2004
CR 04-00057; Seal Return Flow Transmitter
WO 200077085; Troubleshoot/repair/replace FTMU60B
DB-OP-02515; Reactor Coolant Pump and Motor Abnormal Operation; Revision 05
DB-OP-02006; Reactor Coolant Pump Alarm Panel 6 Annunciators; Revision 07
CR 01-2019; Initial Results of Investigation into NRC Information Notice 2000-20
CR 04-00402; Door 515 May Not Have Sufficient Capacity for a HELB
CR 04-00442; MS Line Break in the Turbine Building Adversely Affects Doors in the
Aux. Building
CR 04-00478; MS Line Break Potential Significant Effects
CR 03-11108; Bonnet Gasket Leakage on FW1009
CR 03-11199; FW1009 Valve Gasket Joint Leaking
CR 04-00463; FW1009, MDFP Discharge Check Valve Leaks
WO 200056845; Replace Bonnet Gasket on FW 1009
Attachment
4
WO 200079617; DB-FW1009: Replace Bonnet Gasket
WO 200011379; DB-FW1009 MDFP to Main Feedwater Discharge Check
CR 04-00670; SV101C Leaks By When Energized
WO 200080939; DB-SV101C: Replace
CR 04-00737; AFPT 1 Casing Leak
DB-OP-06201; Main Steam System Operating Procedure; Revision 3
CR 04-00871; X01 A Phase Bushing Oil Leak
Problem Solving Plan for Startup Transformer X01 HV Bushing Leak
DB-SC-03020; 13.8 KV System Bus A & B Transfer Test; Revision 4
DB-SC-03020; 13.8 KV System Bus A & B Transfer Test; Revision 5
10 CFR 50.59 Evaluation for Changing TS Bases for Plant TS 3/4.8, Electric Power
Systems
Switching Order to Service and Repair the Ohio Edison Beaver Line
Problem Solving Plan for Component Cooling Water Ventilation Train 2 Anomalies
DB-SC-04002; Component Cooling Water Pump Room Ventilation System - Train 2;
Revision 01/Total Rewrite
CR 04-01231; CCW Pump Room Ventilation - Train 2 Control Logic Failure
WO 20007804; AFPT 1 Governor Fails to Control
CR 04-00105; AFW Turbine Governor
1R14
Personnel Performance During Nonroutine Plant Evolutions
DB-OP-06903; Plant Shutdown and Cooldown; Revision 13
DB-OP-06226; Startup Feed Pump Operating Procedure; Revision 05
DB-OP-06225; MDFP Operating Procedure; Revision 07
Operations Evolution Order Transfer to the SUFW Pump and Back to MDFP
DB-OP-06900; Plant Heatup; Revision 25
DB-OP-06005; RC Pump Operation; Revision 12
Attachment
5
Operations Evolution Order to Allow Closure of MS101 for Maintenance on SV101C
DB-OP-06201; Main Steam System Operating Procedure; Revision 03
DB-SP-03300; RCS Isolation Check Valve Leak Test; Revision 01
DB-SC-04115; Absolute Position Indication Functional Test; Revision 02
1R19
Post-Maintenance Testing
Drawing E-44B Sheet 6A; Feedwater System AFP Suction Valves; Revision 13
Drawing E-44B Sheet 18; Feedwater System AFP Suction Valve Transfer Relay;
Revision 05
DB-SP-03152; AFW Train 1 Level Control, Interlock and Flow Transmitter Test,
Revision 09
WO 200076786; SP3152-001 04.000 Train 1 TAC 5.9
System Description 015; Auxiliary Feedwater System; Revision 02
WO 20032093; AFP 1 Suction Valve Transfer
DB-SP-04152; AFPT 1 TSS and Overspeed Trip; Revision 08
DB-SP-04153; AFPT 2 TSS and Overspeed Trip; Revision 07
DB-MM-09150; AFPT Maintenance; Revision 06
DP-SP-03445; SFRCS Channel 2 Trip of MS100 and MS101; Revision 03
WO 200080939; DB-SV101C: Replace
DB-SP-03444; SFRCS Channel 1 Trip of MS100 and MS101; Revision 03
1R20
Refueling and Outage
DB-OP-06900; Plant Startup; Revision 15
1R22
Surveillance Testing
DB-SP-03152; AFW Train 1 Level Control, Interlock and Flow Transmitter Test;
Revision 09
DB-SC-03270; Control Rod Assembly Insertion Time Test; Revision 03
Attachment
6
1R23
Temporary Plant Modifications
Temporary Modification 04-0002; K3-1 Auxiliary Feedwater Pump Turbine
10 CFR 50.59 Screen for the Temporary Spray Shield for Auxiliary Feed Pump Turbine
Casing Leak
Problem Solving Plan for AFW Turbine #1 Casing Leak
CR 02-06767; LIR-AFW-JCL Inputs not Bounding
CR 04-00194; AFW Turbine K3-1 Casing Leak
4OA2 Problem Identification and Resolution
OPS-SAYS-I213.05; Auxiliary Feedwater System Description
DB-SP-03157; AFP 1 Response Time Test; Revision 07
DB-MM-090098; AFPT Governor Maintenance; Revision 04
CR 03-01964; OE-15688 AFW Pump Turbine Governor Linkage Binding
CR 03-08210; Misadjustment of Aux Feed Pump #1 Governor
CR 03-08108; AFW Train 1 Response Time Test Failure
CR 03-07975; Auxiliary Feedwater Train 1 Inoperability Due to Response Time
CR 03-07976; Auxiliary Feed Pump #1 Time Response
CR 03-08370; AFP Turbine K3-1 RPM Above Low Speed Stop Setting During Response
Test
CR 04-00830; AFP 1 Response Time Exceeds Acceptance Criteria During
DB-SP-03157
CR 04-00161; CAC Outlet Valves; dated January 6, 2004
CR 03-10371; Various CAC Related Equipment; dated December 1, 2003
CR 03-06837; Various; dated August 22, 2003
CR 02-07640; No ASME Code Review Documented; dated October 8, 2002
CR 02-06860; Various Heat Exchangers; dated September 27, 2002
PCAQR 88-0737; ASME Code Relief Protection; dated September 20, 1988
Attachment
7
SE-95-0056; Removal of Containment Air Cooler Relief Valves; dated July 3, 1995
OS-020 So 1; Operational Schematic Service Water System; Revision 64
OS-020 So 2; Operational Schematic Service Water System; Revision 33
OS-004 So 1; Operational Schematic Decay Heat Removal Low Pressure Injection
System; Revision 36
OS-004 So 2; Operational Schematic Decay Heat Removal Low Pressure Injection
System; Revision 04
OS-021 So 1; Operational Schematic Component Cooling Water System; Revision 31
OS-021 So 2; Operational Schematic Component Cooling Water System; Revision 21
OS-021 So 3; Operational Schematic Component Cooling Water System; Revision 9
OS-041 So 1; Operational Schematic Emergency Diesel Generator Systems; Revision 19
DB-OP-06316; Diesel Generator Operating Procedure; Revision 12
DB-OP-06012; Decay Heat and Low Pressure Injection System Operating Procedure;
Revision 16
DB-OP-06016; Containment Air Cooling System Procedure; Revision 13
DB-OP-06262; Component Cooling Water System Procedure; Revision 7
DB-OP-02000; RPS, SFAS, SFRCS Trip, or SG Tube Rupture; Revision 12
ECR 02-0343-00; Containment Air Cooler Upgrades; dated June 7, 2003
USAR Change Notice 97-073; Code Discrepancies; dated July 31, 1992
Specification No. 7749-M-400; TS for Containment Air Cooler Units; dated
February 22, 1977
Specification No. 1024/1069; Heat Exchangers for Auxiliary System Service; dated
November 14, 1968
Specification No. M-200; Piping Classes; Revision 6
Drawing M-033; Decay Heat Removal System and Emergency Core Cooling Systems;
Revision 21
Drawing M-033B; Decay Heat Removal Train 1; Revision 43
Drawing M-033C; Decay Heat Removal Train 2; Revision 19
Attachment
8
Drawing M-0368; Component Cooling Water System; Revision 31
Drawing M-036; Component Cooling Water System; Revision 15
Drawing M-041; Service Water System; Revision 18
Drawing M-041C; Service Water System for Containment Air Coolers; Revision 24
4OA3 Event Followup
CR 02-03859; Degraded Material Condition of the Containment Emergency Sump
Screen
CR 02-05461; Past Operability of Containment Emergency Sump
CR 03-03398; Containment Gas Analyzer CCW Deficiencies
CR 03-04882; Containment Gas Analyzer Moisture Trap Design Issue - Potential
Containment Bypass
CR 03-05529; Regulator in Air Supply Line to Hydrogen Analyzers Not On Drawings
CR 02-07169; LIR CCW - Lack of CCW Flow Verification to Essential Components
CR 03-04871; Containment Gas Analyzer Moisture Trap Design Issue
CR 03-05204; Asset Database Errors
CR 03-05605; Adequacy of Testing for TS Components
CR 02-08008; LIR CCW - Hydrogen Analyzer Function in SD-23 Not Consistent With
USAR/Procedure
CR 03-05943; Overly Conservative Licensee Event Report (LER) Reporting Criterion
LER 2003-006; Potential Errors in Analysis of Block Walls Regarding HELB Differential
Pressure and Seismic Events
Operability Evaluation 03-0015; Analysis for CR 03-05399
CR 03-02910; Seismic Analysis of Masonry Walls
CR 03-03937; Masonry Wall Failure
Root Cause Analysis Report for CR 03-02910 and CR 03-03937; July 11, 2003
CR 03-05399; Tornado Differential Pressure Analysis and Seismic Analysis of Masonry
Walls
Attachment
9
M-236B; Piping Isometric, Component Cooling System, Auxiliary Building El 565'-0";
Revision 14
LER 50-346/03-001; Potential Inability of Air-Operated Valves to Function During Design
Basis Conditions; Revision 0 and Revision 1.
CR 03-05628; CCW Heat Exchanger 1-2 Outlet Control Valve
CR 03-04878; SW 1356, SW 1357, SW 1358 Air Accumulators
CR 03-04158; Component Cooling Water Service Water Outlet Isolation Valves
CR 03-01253; CCW To Nonessential Loads Isolation Valve
CR 03-01040; RCP Seal Return Isolation Valve
CR 03-00830; DH Heat Exchanger CCW Discharge Isolation
CR 02-07781; Service Water Outlet Valves From CAC Coolers
CR 02-07750; SW 1356, SW 1357, SW 1358
CR 99-2111; CC1495 Not Fully Closed
PCAQR 97-1082; Valve CC1495 Failed to Stroke Full Closed.
MPR Specification 200-037; Specification for Ball Valve Internal Design Information;
Revision 0
MPR Specification 092-013-A5; Specification for Internal Design Information for Pivoting
Cylinder Air Actuators With Spring Return; Revision 1
TV-1424; Valve/Damper Actuator and/or Accessories Information Sheet Davis-Besse;
dated December 26, 2001
TV-1429; Valve/Damper Actuator and/or Accessories Information Sheet Davis-Besse;
dated March 19, 1999
TV-1434; Valve/Damper Actuator and/or Accessories Information Sheet Davis-Besse;
dated July 11, 2002
Kalsi Engineering Report 1666C; Engineering Evaluation of 12-Inch DACE-12 Neves
Ball Valve Design Modification; dated August 3, 1990
C-ME-011.01-139; Component Level Review Calculation for AOV SW1424/1429/1434;
Revision 0
C-ME-016.04-031; Maximum Expected Differential Pressure For Valves CC-1467 and
CC-1469; Revision 0
Attachment
15
C-ME-016.04-035; Component Level Review Calc For AOV CC1467/1469; Revision 0
C-NSA-099.16-80; Risk Assessment of Air Dependency for Component Cooling Water
Heat Exchanger Valves SW 1424, SW 1434, SW 1429 and CC 1495;
November 4, 2003
Component Cooling Water Heat Exchanger Service Water Outlet Isolation Valves
Operability Analysis for Loss of Instrument Air During a Design Basis Accident; dated
July 23, 2003
DB-PF-04167; Test to Evaluate the Dynamic Data During Stroking of SW 1434; dated
July 14, 2003
Drawing ND-281547-01; D2CE-12 Valve With BJVAR20 Pneumatic Actuator;
Revision T4
CR 04-00181; Missed TS Action Statement
CR 03-11314; Corrective Actions Taken to Improve Operational Deficiencies Appear
Ineffective
CR 03-11033; RRATI - Operations Failure to Meet Standards and Expectations
CR 03-11414; Missed TS Entry
4OA5 Other Activities
DB-OP-00000; Conduct of Operations; Revision 07
DB-OP-03204; Channel Functional Test and Calibration of SFRCS Actuation Channel 2,
Steam Generator Differential Pressure Inputs PDS-2685A, PDS-2685B, PDS-2686C,
and PDS-2686D; Revision 05
CR-00181; Missed TS Action Statement
Root Cause Analysis Report for CR 2004-00181 dated January 18, 2004
DP-OP-06900; Plant Heatup; Revision 25
DB-SP-03302; CF 28 and CF 29 Leak Test; Revision 02
DB-OP-6911; Pre-Startup Checklist; Revision 07
1EP2
Alert and Notification System (ANS) Testing
Procedure RA-EP-00400; Prompt Notification System Maintenance; Revision 2
Procedure RA-EP-00420; Response to Prompt Notification System Malfunction;
Revision 1
Attachment
16
Internal Memorandum; RA-EP-00400 Annual Prompt Notification System Inspection;
dated November 13, 2002
Internal Memorandum; RA-EP-00400 Biennial Prompt Notification System Siren
Acoustic Testing; dated December 2, 2002
Internal Memorandum; 2003 Prompt Notification System Annual Inspection; dated
August 2, 2003
1EP3
Emergency Response Organization (ERO) Augmentation Testing
Procedure RA-EP-00100; Emergency Plan Training Program; Revision 6
Procedure RA-EP-00550; Computerized Automated Notification System; Revision 2
Procedure RA-EP-02110; Emergency Notification; Revision 5
Records of Semi-Annual, Off-Hours, Unannounced Augmentation Drills - March 2002
Through September 2003
Training Records of a Random Sample of 30 Station Personnel Assigned to Key or
Support ERO Positions
CR 03-05185; Review of Beaver Valley Plants Second Unannounced Emergency
Facilities Activation Drill
CR 03-08098; Several Persons Did Not Correctly Make Notification Data Entries During
September 2003 Drill
CR 03-08197; Determine What is Optimum Identification Code to Use for Required
Notification Data Entries
1EP4
Emergency Action Level (EAL) and Emergency Plan Changes
Davis-Besse Nuclear Power Station Emergency Plan; Table 4-1, Summary of
Emergency Action Levels; Revision 22
EPIP RA-EP-01500; Emergency Classification; Revisions 3 and 4
Former EPIP EI-1300.01; Emergency Plan Activation; Revision 10
Internal Memorandum; Low Forebay Water Level Emergency Action Levels; dated
December 22, 1988
Response to Request for Assistance on Low Forebay Water Level Emergency Action
Levels; dated January 31, 1989
Internal Memorandum; Unusual Event of February 8, 1989; dated February 28, 1989
Attachment
17
Internal Memorandum; Review of the Response to the Unusual Event of October 8,
1990; dated October 15, 1990
10 CFR 50.54(q) Review; Change 2 to Revision 3 of Emergency Classification
Procedures Toxic or Flammable Gas Emergency Action Levels; dated August 26, 1991
CR 04-00715; Evaluate EAL Issue at Point Beach Plant for Potential Impact on Davis-Besse
CR 04-01475; Revise an Indicator of EAL 5.A.1
CR-04-01500; Change Required to Emergency Classification Procedure RA-EP-01500
Review of EAL 1.D.1; dated February 11, 2004
Required Reading Packages on Changes to Emergency Classification Procedure; dated
February 27, 2004
NRC Inspection Report 50-346/85011(DRSS); dated May 1, 1985
NRC Inspection Report 50-346/86007(DRSS); dated May 8, 1986
1EP5
Correction of Emergency Preparedness Weaknesses and Deficiencies
Davis-Besse Nuclear Power Station August 14, 2003 Blackout Unusual Event; undated
Internal Memorandum; Three 2002 Integrated Drills Report; dated November 11, 2002
Internal Memorandum; April 10, 2003 Integrated Drill Report; dated June 9, 2003
Internal Memorandum; May 13, 2003 Dry Run Drill Report; dated June 30, 2003
Internal Memorandum; June 10, 2003 Evaluated Exercise Report; dated
August 14, 2003
Internal Memorandum; July 31, 2003 Integrated Drill Report; dated September 29, 2003
Internal Memorandum; October 16, 2003 Integrated Drill Report; dated
December 22, 2003
Internal Memorandum; 2002 Medical Drill Evaluation; dated November 11, 2002
Internal Memorandum; 2003 Medical Drills Evaluation; dated December 22, 2003
Internal Memorandum; December 26, 2002 Post Accident Sampling System Drill Report;
dated February 24, 2003
Internal Memorandum; December 10, 2003 Post Accident Sampling System Drill Report;
dated December 29, 2003
Attachment
18
Davis-Besse Nuclear Quality Assessment Quarterly Report DB-C-02-02; dated
August 9, 2002
Davis-Besse Nuclear Quality Assessment Quarterly Report DB-C-02-03; dated
November 15, 2002
Davis-Besse Nuclear Quality Assessment Quarterly Report DB-C-02-04; dated
February 19, 2003
Davis-Besse Nuclear Quality Assessment Quarterly Report DB-C-03-02; dated
September 1, 2003
Davis-Besse Nuclear Quality Assessment Quarterly Report DB-C-03-03; dated
November 17, 2003
Draft Davis-Besse Nuclear Quality Assessment Quarterly Report DB-C-03-04; dated
January 13, 2004
Procedure RA-EP-00200; Emergency Plan Drill and Exercise Program; Revision 4
Procedure RA-EP-02010; Emergency Management; Revision 4
Training Handout to Radiation Protection Personnel on Expected Responses to a Steam
Generator Tube Leak
CR 03-02926; Three Objectives Not Successfully Demonstrated During April 2003 Drill
CR 03-03831; Revise Notification Guidance for a Radiological Release in Progress That
is Attributable to an Emergency Event
CR 03-03832; One Objective Not Adequately Demonstrated During May 2003 Drill
CR 03-04300; Provide More Training on Steam Generator Tube Leak Release
Pathways
CR 03-04603; Assess Adequacy of Respirator Qualification Records and Spectacle Kits
for Operations Support Center Staff
CR 03-04622; Reassess Roles of Liaisons Sent to State and County Response
Facilities
CR 03-04739; Four Forms Used by ERO Responders or Drill Controllers Not
Proceduralized
CR 03-06673; Emergency Preparedness Staff Training Provisions
CR 03-06737; Weaknesses in an ERO Training Session and Associated Lesson Plan
Attachment
19
CR 03-07662; Evaluate High Temperature Alarm Set Point for the Davis-Besse
Administration Buildings Emergency Diesel Generator Room
CR 03-06831; Opportunities for Improvement From August 14, 2003, Unusual Event
Response
CR 03-06835; Improve Procedures on Optional ERO Activation Following an Unusual
Event Declaration
CR 03-07070; More Opportunities for Improvement From August 14, 2003, Unusual
Event Response
CR 03-07165; Reassess Locations of ERO Telephones in the Control Room and Shift
Managers Office
CR 03-09018; Revise Joint Public Information Center Procedure to Address How to
Handle an Event Reclassification During a Media Briefing
4OA1 Performance Indicator (PI) Verification
PI Desktop Guide for ERO Drill Participation; dated June 2003
PI Desktop Guide for ANS Reliability; dated January 2003
PI Desktop Guide for Drill and Exercise Performance Indicator; dated February 2002
Monthly Reports and Supporting Records for the ERO, ANS, and Drill and Exercise
Performance Indicators - April 2003 Through December 2003
CR 02-00371; One Siren Failed to Sound During February 2002 Test
CR 02-02284; One Siren Failed to Sound During May 2002 Test
CR 02-02997; One Siren Had Motor Failure During July 2002 Test
CR 03-01854; One Siren Failed to Sound During March 2003 Test
CR 03-02410; One Siren Failed to Sound During Extra March 2003 Test
CR 03-04931; Notification Form Error During June 2003 Control Room Simulator
Session
CR 03-06210; Notification Form Error During July 2003 Control Room Simulator Session
CR 03-06265; Wrong Notification Form Used by Emergency Control Center Staff During
July 2003 Drill
CR 03-06769; Declining Trend in Drill and Exercise Performance Indicator
Attachment
20
LIST OF ACRONYMS USED
Agency-wide Document Access and Management System
AFP
Auxiliary Feedwater Pump
AFPT
Auxiliary Feedwater Pump Turbine
Alert and Notification System
Air Operated Valve
American Society of Mechanical Engineers
Apparent Violation
BWST
Borated Water Storage Tank
Containment Air Cooler
Component Cooling Water
Core Damage Frequency
CFR
Code of Federal Regulations
CGAS
Containment Gas Analyzers Systems
CR
Condition Report
Division of Reactor Safety
Emergency Action Level
Engineering Change Request
EDGJW
Emergency Diesel Generator Jacket Water
Emergency Plan Implementing Procedure
Emergency Response Organization
FirstEnergy Nuclear Operating Company
High Pressure Injection
Human Reliability Analysis
IMC
Inspection Manual Chapter
IR
Inspection Report
LER
Licensee Event Report
Loss of Coolant Accident
LOIA
Loss of Instrument Air
Non-Cited Violation
Net Positive Suction Head
NRC
United States Nuclear Regulatory Commission
Nuclear Regulatory Guide
OP
Over-pressure
Publicly Available Records
Performance Indicator
Pressure Operated Relief Valve
Reactor Coolant Pump
Significance Determination Process
SFAS
Safety Features Actuation System
Attachment
21
SFRCS
Steam Feedwater Rupture Control System
Senior Reactor Analyst
Structures, Systems, Components
Task Interface Agreement
TRBL
Trouble
TS
Technical Specifications
Unresolved Item
Updated Safety Analysis Report
Work Order