ML052070351
| ML052070351 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/26/2005 |
| From: | Anderson C NRC/RGN-I/DRP/PB5 |
| To: | Balduzzi M Entergy Nuclear Operations |
| References | |
| IR-05-003 | |
| Download: ML052070351 (51) | |
See also: IR 05000293/2005003
Text
July 26, 2005
Mr. Michael A. Balduzzi
Site Vice President
Entergy Nuclear Operations, Inc.
Pilgrim Nuclear Power Station
600 Rocky Hill Road
Plymouth, Massachusetts 02360-5508
SUBJECT:
PILGRIM NUCLEAR POWER STATION - NRC INTEGRATED INSPECTION
REPORT 05000293/2005003
Dear Mr. Balduzzi:
On June 30, 2005, the US Nuclear Regulatory Commission (NRC) completed an inspection at
your Pilgrim reactor facility. The enclosed integrated inspection report documents the
inspection findings, which were discussed on July 7, 2005, with Mr T. Kirwin and 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.
On the basis of the results of this inspection, no findings of significance were identified. After
this inspection period ended, the NRC issued a letter on July 14, 2005, issuing a Severity Level
III Notice of Violation and Proposed Imposition of Civil Penalty. This inspection report
acknowledges issuance of that letter and assigns a tracking number for the Severity Level III
Notice of Violation. Additionally, licensee-identified violations which were determined to be of
very low safety significance are listed in Section 4OA7 of this report. If you contest any NCV in
this report, you should provide a response with the basis for your denial, within 30 days of the
dated of this inspection report, to the U.S. Nuclear Regulatory Commission, ATTN: Document
Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region
I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001; and the NRC Resident Inspector at Pilgrim.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its
enclosures will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of the NRCs document
Michael Balduzzi
2
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/
Clifford Anderson, Chief
Projects Branch 5
Division of Reactor Projects
Docket No.
50-293
License No.
Enclosure:
Inspection Report 05000293/2005003
w/Attachment: Supplemental Information
cc w/encl:
G. J. Taylor, Chief Executive Officer, Entergy Operations
M. Kansler, President, Entergy Nuclear Operations, Inc.
J. T. Herron, Senior Vice President and Chief Operating Officer
C. Schwarz, Vice-President, Operations Support
S. J. Bethay, Director, Nuclear Safety Assurance
O. Limpias, Vice President, Engineering
J. F. McCann, Director, Licensing
C. D. Faison, Manager, Licensing
M. J. Colomb, Director of Oversight, Entergy Nuclear Operations, Inc.
B. S. Ford, Manager, Licensing, Entergy Nuclear Operations, Inc.
T. C. McCullough, Assistant General Counsel
S. Lousteau, Treasury Department, Entergy Services, Inc.
R. Walker, Department of Public Health, Commonwealth of Massachusetts
The Honorable Therese Murray
The Honorable Vincent deMacedo
Chairman, Plymouth Board of Selectmen
Chairman, Duxbury Board of Selectmen
Chairman, Nuclear Matters Committee
Plymouth Civil Defense Director
D. OConnor, Massachusetts Secretary of Energy Resources
J. Miller, Senior Issues Manager
Office of the Commissioner, Massachusetts Department of Environmental Protection
Office of the Attorney General, Commonwealth of Massachusetts
Electric Power Division, Commonwealth of Massachusetts
R. Shadis, New England Coalition Staff
D. Katz, Citizens Awareness Network
Chairman, Citizens Urging Responsible Energy
J. Sniezek, PWR SRC Consultant
Michael Balduzzi
3
C. McCombs, Acting Director, MEMA and Commonwealth of Massachusettts, SLO Designee
Commonwealth of Massachusetts, Secretary of Public Safety
Michael Balduzzi
4
Distribution w/encl:
S. Collins, RA
M. Dapas, DRA
S. Lee, RI EDO Coordinator
C. Anderson, DRP
D. Florek, DRP
P. Krohn, DRP
B. Norris, DRP
R. Ennis, NRR
W. Raymond, DRP, Senior Resident Inspector
C. Welch, DRP, Resident Inspector
A. Ford, DRP, Resident OA
Region I Docket Room (with concurrences)
DOCUMENT NAME: E:\\Filenet\\ML052070351.wpd
SISP Review Complete: DJF1 (Reviewers Initials)
After declaring this document An Official Agency Record it will be released to the Public.
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
RI/DRP
RI/DRP
/
NAME
DFlorek
CAnderson
DATE
07/25/05
07/26/05
OFFICIAL RECORD COPY
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No:
50-293
License No:
Report No:
Licensee:
Entergy Nuclear Operations, Inc.
Facility:
Pilgrim Nuclear Power Station
Location:
600 Rocky Hill Road
Plymouth, MA 02360
Inspection Period:
April 1, 2005 - June 30, 2005
Inspectors:
W. Raymond, Senior Resident Inspector
C. Welch, Resident Inspector
T. Burns, Reactor Inspector
D. Silk, Senior Emergency Preparedness Inspector
A. Ziedonis, Reactor Engineer Intern
J. Schoppy, DRS, Senior Reactor Inspector
T. OHara, DRS, Reactor Inspector
D. Szwarc, Reactor Engineer Intern
J. McFadden, Health Physicist
Approved By:
Clifford Anderson, Chief
Projects Branch 5
Division of Reactor Projects
Enclosure
ii
TABLE OF CONTENTS
SUMMARY OF FINDINGS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1R02
Evaluations of Changes, Tests, or Experiments . . . . . . . . . . . . . . . . . . . . . . . . 1
1R04
Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
1R05
Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
1R08
Inservice Inspection (ISI), Inspection Procedure . . . . . . . . . . . . . . . . . . . . . . . . 3
1R11
Licensed Operator Requalification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
1R13
Maintenance Risk Assessments and Emergent Work Control . . . . . . . . . . . . . . 5
1R14
Personnel Performance During Non-routine Plant Evolutions . . . . . . . . . . . . . . 6
1R15
Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1R16
Operator Work-Arounds
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
1R17
Permanent Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
1R19
Post-Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R20
Refueling and Other Outage Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9
1R22
Surveillance Testing
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1R23
Temporary Plant Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
1EP2 Alert and Notification System (ANS) Testing . . . . . . . . . . . . . . . . . . . . . . . . . 14
1EP3 Emergency Response Organization (ERO) Augmentation Testing
. . . . . . . . 14
1EP4 Emergency Action Level (EAL) Revision Review . . . . . . . . . . . . . . . . . . . . . . 14
1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies . . . . 15
RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
2OS1 Access Control to Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . 15
2OS2 ALARA Planning and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18
2OS3 Radiation Monitoring Instrumentation and Protective Equipment . . . . . . . . . . 21
OTHER ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4OA1 Performance Indicator (PI) Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
4OA3 Event Follow-up . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
4OA5 Other . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25
4OA6 Meetings, Including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
4OA7 Licensee-Identified Violations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26
SUPPLEMENTAL INFORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . A-2
LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2
LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-14
Enclosure
iii
SUMMARY OF FINDINGS
IR 05000293/2005003; 04/01-06/30/2005; Pilgrim Nuclear Power Station, Other
The report covered a 13 week period of inspection by resident inspectors and announced
inspections by regional inspectors including a senior emergency preparedness Inspector, senior
reactor inspector, health physics inspector and reactor inspectors. One Severity Level III Notice
of Violation and Proposed Imposition of Civil Penalty issued in a letter dated July 14, 2005, is
documented in this report. The significance of most findings is indicated by their color (Green,
White, Yellow, Red) using Inspection Manual Chapter (IMC) 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, July 2000.
A.
Inspector Identified and Self-Revealing Findings
Miscellaneous
SL-III In a letter dated July 14, 2005, the NRC issued a Severity Level III Notice of
Violation and Proposed Imposition of Civil Penalty to Entergy in the base amount of
$60,000 associated with a Severity Level III problem. The Severity Level III problem
involved four violations of NRC requirements related to Technical Specification 5.4.1, 10 CFR Part 50 Appendix B, and 10 CFR Part 26. The specific violations involved: (1) a
Pilgrim control room supervisor sleeping for approximately four minutes in the control
room; (2) a reactor operator observing the sleeping control room supervisor, but
deliberately not taking immediate actions to awaken the control room supervisor, inform
appropriate site personnel and initiate a condition report; (3) a Shift Manager, in
careless disregard of requirements, although taking some actions, not informing
appropriate site personnel and initiating a condition report; and (4) the sleeping control
room supervisor not being relieved of duty and for-cause Fitness-for-Duty tested. There
were no actual safety consequences resulting from this event because there were no
plant conditions that warranted immediate action.
B.
Licensee Identified Violations
Violations of very low safety significance, which were identified by Entergy, have been
reviewed by the inspector. Corrective actions taken or planned by Entergy have been
entered into Entergys corrective action program. The violations are listed in Section
4OA7 of this report.
REPORT DETAILS
Summary of Plant Status
Pilgrim Nuclear Power Station operated at reduced power in end-of-cycle coast down at the
beginning of the period. The plant shutdown to conduct refueling outage RFO#15 on
April 17, 2005. The outage was conducted with no major events. Entergy completed the
scheduled outage tests and maintenance, including the 10 year inservice inspection and reactor
vessel exams; repaired cracked welds on steam dryer tie-bars; and removed a leaky fuel
bundle. Following the outage, the reactor was made critical on May 11 and full power was
reached on May 15. The plant operated during the period at 100 percent (%) core thermal
power, except for short periods of planned operation at reduced power for routine testing and
condenser maintenance.
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R02
Evaluations of Changes, Tests, or Experiments (71111.02)
a.
Inspection Scope (19 samples)
The inspectors reviewed six safety evaluations (SEs) listed in the attachment, all of
which were either issued during the past two years or associated with plant
modifications that were completed during the past two years. The SEs reviewed were in
the Initiating Event, Mitigating Systems, and Barrier Integrity cornerstones. The
selected SEs were reviewed to verify that changes to the facility or procedures as
described in the Updated Final Safety Analysis Reports (UFSAR) were reviewed and
documented in accordance with 10 CFR 50.59, that safety issues pertinent to the
changes were properly resolved or adequately addressed, and that Entergy had
appropriately evaluated whether the changes and tests could be accomplished without
obtaining license amendments.
The inspectors also reviewed 13 screened-out evaluations for changes, tests and
experiments for which Entergy determined that SEs were not required. This review was
performed to verify that Entergys threshold for performing SEs was consistent with 10 CFR 50.59. The listing of the SEs and screened-out evaluations reviewed is provided in
the attachment.
In addition, the inspectors reviewed the administrative procedures that were used to
control the screening, preparation, and issuance of the SEs to ensure that the
procedure adequately covered the requirements of 10 CFR 50.59.
The inspectors reviewed condition reports (CRs) associated with 10 CFR 50.59 issues
to ensure that Entergy was identifying, evaluating, and correcting problems associated
with these areas and that the planned or completed corrective actions for the issues
were appropriate. The inspectors also reviewed self-assessments related to 10 CFR 50.59 SEs at Pilgrim. The listing of the condition reports and self assessments reviewed
is provided in the attachment.
2
Enclosure
b.
Findings
On March 10, 2005, Entergy approved a change to Procedure No. 8.M.3-1, Special Test
for Automatic ECCS Load Sequencing of Diesels and Shutdown Transformer with
Simulated Loss of Offsite Power and Special Shutdown Transformer Load Test, using
its 10 CFR 50.59 process. One aspect of the change involved removing the automatic
sequencing of the reactor building closed cooling water (RBCCW) and salt service water
(SSW) pumps (emergency loads) during the simulated loss of offsite power (LOP)/loss-
of-coolant accident (LOCA) testing to allow Entergy to perform the testing without
impacting the refueling outage critical path. The inspectors noted that Technical
Specifications (TSs) 4.9.A.1.b and 4.9.A.1.c require the energization of the auto-
connected emergency loads through the load sequencer to verify loading onto the
emergency diesel generators (EDGs) and shutdown transformer (SDT), respectively.
Entergy justified the change based, in part, on the TS 4.9.A.1 bases wording (the bases
did not specifically mention these emergency loads) and their existing load sequencing
overlap testing. The inspectors determined that the overlap testing does not actually
verify the load sequencing following an EDG start from ambient conditions as prescribed
by TS 4.9.A.1.b nor SDT loading following an EDG trip as prescribed by TS 4.9.A.1.c.
As a result, the inspectors questioned whether Entergy needed a TS Amendment to
make this change.
The NRC requires more information to determine whether this issue is an acceptable
item. Pending further review and discussion with the Office of Nuclear Reactor
Regulation (NRR), this is an unresolved item (URI). (URI 05000293/2005003-01,
2005012-02 Adequacy of Entergys Change to LOP/LOCA Testing Without Seeking
a TS Amendment)
1R04
Equipment Alignment (71111.04)
a.
Inspection Scope (5 samples)
The inspector completed a partial system review of the below-listed risk significant
systems during periods when its redundant train or system was out-of-service for
maintenance and/or testing or on restoration of the train. The position of key valves,
breakers, and control switches, required for system operability, were determined by field
walkdown and/or review of the main control board indicators. To ascertain the required
system configuration, the inspectors reviewed plant procedures, system drawings, the
Updated Final Safety Analysis Report, and the Technical Specifications. The references
used for this review are described in the attachment to this report. This inspection
activity represented five samples.
A RHR train, during maintenance on the B train per MR 05102963.
Alternate Shutdown Cooling (Mode 1) on 4/25/05.
Alternate Shutdown Cooling (Mode 2) on 4/26/05 while shutdown cooling was
secured for maintenance.
B RBCCW train, post-refuel startup readiness.
3
Enclosure
B Core Spray Train, post-refuel startup readiness.
b.
Findings
No findings of significance were identified.
1R05
Fire Protection (71111.05)
a.
Inspection Scope (12 samples)
The inspector toured selective areas of the plant to observe conditions related to: (1)
transient combustibles and ignition sources; (2) fire detection systems; (3) manual
firefighting equipment and capability; and (4) passive fire protection features. The
inspector evaluated whether the material condition of active and passive fire protection
systems features and their operational lineup and readiness were adequate. The
inspector also reviewed the applicable fire hazard analysis fire zone data sheets and
selective surveillance procedures to ensure that the specified fire suppression systems
surveillance criteria were met. The references used for this review are described in the
attachment to this report. This inspection activity represented twelve samples.
Fire Zone 1.30, Drywell
Fire Zones 2.9A & 2.10A, Condenser Bay
Fire Zone 2.8, Condensate Pump Area
Fire Zones 1.21 & 1.22, RBCCW & TBCCW pumps/heat exchangers rooms
Fire Zones 2.11 & 2.12, Feedwater Pumps Area
Fire Zones 4.1 & 4.3, A and B Emergency Diesel Generator Rooms
Fire Zones 4.2 & 4.4, A and B Emergency Diesel Day Tank Rooms
b.
Findings
No findings of significance were identified.
1R08
Inservice Inspection (ISI), Inspection Procedure (71111.08)
a.
Inspection Scope (1 sample)
The inspector observed selected in-process nondestructive examination (NDE)
activities. Also, the inspector reviewed documentation of NDE and repair/replacement
activities. The activities reviewed were based on the inspection procedure objectives
and risk priority of those components and systems where degradation could result in a
significant increase in risk of core damage. The observations and documentation
reviews were performed to verify activities were accomplished in accordance with the
American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code
requirements. The inspector reviewed reports that documented the performance and
results of ISI examinations completed during this period. Also, the inspector evaluated
Entergys effectiveness in resolving relevant indications identified during ISI activities.
4
Enclosure
The inspector observed manual ultrasonic testing (UT) and reviewed documentation of
the magnetic particle (MT), penetrant (PT), and visual test (VT) activities to verify the
effectiveness of the examiner and process for identifying degradation of risk significant
systems, structures and components. The inspector reviewed documentation to
determine whether test examiners qualifications were current and in accordance with
the ASME Code requirements and, as applicable, performance demonstration initiative
(PDI) qualifications were current. The inspector reviewed the UT test of the reactor
pressure vessel (RPV) head to flange weld RPV-HF-240-360 and the UT test
examination results of RPV top head welds RPV-TH-M4 and M5. Condition reports
(CR) were initiated for indications identified in each of the two top head welds. The
indications identified in these welds were characterized, sized and entered into the
corrective action program for engineering evaluation and disposition.
The inspector examined Entergys evaluation and disposition for continued operation
without repair or rework of non-conforming conditions identified during ISI activities by
review of CR-2003-01095 (reduced wall thickness on torus shell) and CR 2005-01916
(indication in N2F nozzle to safe end weld). The inspector reviewed a portion of the
remote in-vessel visual inspection of the reactor steam dryer base metal, structural
welds, tie bar welds and the tack welds of the leveling screws. The review was
conducted to evaluate examiner skill, test equipment performance, examination
technique, and inspection environment (water clarity) to verify Entergys ability to identify
and characterize observed indications.
The inspector did not review outage welding activities on the pressure boundary of
ASME Code Class 1 or 2 systems since no ASME Section XI welding activities on these
systems had been performed prior to the period the inspector was on site or were
underway during the period the inspector was on site.
The inspector reviewed CR 2005-01914, CR 2005-02322, and Maintenance Request
05108176 associated with acceptance of a socket weld in the standby liquid control
system. The socket weld was UT and PT tested but an indication identified during the
UT test could not be characterized and sized because of procedure limitations. The
inspector conducted the review to verify the activities were in accordance with the
applicable ASME Code requirements.
The inspector reviewed Condition Report (CR) 2005-01608 and Indication Notification
Report (INR) RFO15-05-01, Rev. 1, which identified cracks discovered in the welds of
four tie bars of the steam dryer. The tie bars maintain spacing and provide lateral
support to the steam dryer banks. The inspector reviewed the CR to determine whether
the tie bar weld cracks identified during non-destructive testing were reported,
characterized, evaluated and appropriately dispositioned and entered into the corrective
action program. Also, the inspector reviewed CR 2005-01916 which had been initiated
to report the identification of a circumferential indication in the N2F (recirculation inlet)
nozzle to safe-end weld. The indication had not been identified during previous
ultrasonic tests. The indication was identified during this examination period using the
qualified Performance Demonstration Initiative (PDI) ultrasonic test and was evaluated,
5
Enclosure
characterized, sized and dispositioned as accept for continued use by fracture
mechanics evaluation in accordance with ASME Section XI.
b.
Findings
No findings of significance were identified.
1R11
Licensed Operator Requalification (71111.11)
1.
Licensed Operator Simulator Training
a.
Inspection Scope (1 sample)
The inspector observed the performance of an operator crew during a simulator training
session on June 8, 2005. The training was conducted per module O-RQ-06-02-102 as
part of licensed operator requalification program. The inspector also reviewed training
conducted on June 8 on the procedures related to offsite power and observed a joint
training session between the Pilgrim Station and the Transmission System operator
(TSO). The simulator scenarios involved operational transients and loss of power
events. The training with the TSO personnel reviewed procedures and protocols to
monitor grid reliability and to enhance communications in response to degraded grid
conditions. The inspector evaluated whether the crew met the training scenario
objectives and performed the critical tasks. The inspector evaluated whether the crew
was properly using system operating procedures and emergency operating procedures.
The inspector also evaluated whether the post-training review discussed any relevant
lessons learned and highlighted actions to improve crew performance. This inspection
activity represented one sample.
b.
Findings
No findings of significance were identified.
1R13
Maintenance Risk Assessments and Emergent Work Control (71111.13)
a.
Inspection Scope (6 samples)
The inspector evaluated on-line risk management for planned and emergent work. The
inspector reviewed maintenance risk evaluations, work schedules, recent corrective
actions, and control room logs to verify that other concurrent planned and emergent
maintenance or surveillance activities did not adversely affect the plant risk already
incurred with the out of service components. The inspector evaluated whether Entergy
took the necessary steps to control work activities, took actions to minimize the
probability of initiating events and maintained the functional capability of mitigating
systems. The inspector assessed Pilgrims risk management actions during plant
walkdowns. The inspector also discussed the risk management with maintenance,
engineering and operations personnel as applicable for the activities. Other references
6
Enclosure
used for the inspection are identified in the attachment to this report. The inspection
covered the following six samples:
MR 05102963, 05102965, 05102969, B RHR System Valve Maintenance
MR 0001297, Planned Startup Transformer maintenance on April 4-6
MR 01108097, B RBCCW Heat Exchanger Inspection and Repair
Local Leak rate testing of MO-1001-50 on April 26, 2005.
April 21, 2005 during bus outages on A6 and A8.
April 28, 2005, during special testing of the ECCS and EDG load sequencing.
b.
Findings
No Findings of significance were identified.
1R14
Personnel Performance During Non-routine Plant Evolutions (71111.14)
a.
Inspection Scope (2 samples)
The inspectors observed the following non-routine planned evolutions or portions thereof
to assess the performance of the control room operators. The inspections focused on
command and control, communications, procedure adherence, and response to
abnormal conditions and/or alarms.
Procedure 2.2.20, Core Spray; for reactor cavity fill with core spray from the
CST.
Procedure 8.M.3-1, Special Test for Automatic ECCS Load Sequencing of
Diesels and Shutdown Transformer with Simulated Loss of Off-Site Power and
Special Shutdown Transformer Load Test.
b.
Findings
No findings of significance were identified.
1R15
Operability Evaluations (71111.15)
a.
Inspection Scope (7 samples)
The inspector reviewed selected operability determinations to assess the adequacy of
the evaluations, the use and control of compensatory measures, compliance with the
technical specifications, and the risk significance of the issues. The inspector used the
technical specifications, Final Safety Analysis Report, associated Design Basis
Documents, Procedure ENN-OP-104 Operability Determinations, and the additional
references listed in the attachment to this report for Section 1R15. This review covered
seven inspection samples.
7
Enclosure
OE and REO CR 200501136, Potential Safety Limit Violation for Analyzed
Operational Occurrences, (CR 200501136, GE Part 21 Report SC05-03)
Condition Reports 200502037, 200501711, 200501851 and 200501748 identified
safety-related snubbers SS-23-20-36, SS-2-20-25, SS-2-20-02, and SS-10-20-
08 did not have a visible hydraulic fluid level. The inspector reviewed test results
acquired per 3.M.4-37, Hydraulic and Mechanical Snubbers Functional Test, to
verify Entergys determination that the snubbers were operable.
OE and REO CR200502618, 8.M.3-1 loss of off-site power test discrepancies
MO-1001-29A torque switch setting low out of specification past operability
determination (CR 200501820).
OE and REO CR 200501028, MO-1001-29A control power fuse installed in
neutral vs power feed.
CR 200503168, B EDG POT Fuse Failure During Test Run (MR 05109337)
CR 200503140 and 200503151, Both EDGs Inoperable due to High Ambient
Temperatures
b.
Findings
No findings of significance were identified.
1R16
Operator Work-Arounds (71111.16)
a.
Inspection Scope (2 samples)
The inspector reviewed the operator work around, burden, and tour lists to evaluate the
potential cumulative impact of the equipment deficiencies on the operators ability to
implement abnormal or emergency operating procedures. The inspector walked down
the control room panels and selected plant areas to review the impact of the deficiencies
and to ensure that applicable deficiencies were captured in Entergys deficiency list. The
inspector discussed the operator workarounds with station personnel to assess the
aggregate impact on plant operations. During the review, the inspector used the criteria
contained in Entergys procedure 1.3.34.4. This inspection covered one inspection
sample of the cumulative effects of operator workarounds.
This review covered one inspection sample of specific operator workarounds. The
inspector reviewed Entergys actions to address item #349, inoperable emergency
lights, in the list of operator compensatory measures. The inspector reviewed the
deficiencies to determine if the functional capability of the system or human reliability in
responding to an initiating event was affected. The inspector evaluated the effect of the
deficiency on the operators ability to implement abnormal and emergency operating
procedures.
The inspector determined whether Entergy evaluated deficiencies for potential impact as
operator workarounds, entered them into the corrective action process, and had planned
maintenance activities to correct the identified operational deficiencies. References
used during this inspection are identified in the attachment to this report.
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Enclosure
b.
Findings
No findings of significance were identified.
1R17
Permanent Plant Modifications (71111.17)
a.
Inspection Scope (6 samples)
The inspectors reviewed six risk-significant plant modification packages selected from
the design changes that were completed within the past two years. The review was
performed to verify that: (1) the design bases, licensing bases, and performance
capability of risk significant structures, systems, and components (SSCs) had not been
degraded through the modifications; and, (2) the modifications performed during
increased risk configurations did not place the plant in an unsafe condition. The listing
of the modifications reviewed is provided in The attachment.
The selected plant modifications were distributed among the Initiating Event, Mitigating
Systems, and Barrier Integrity cornerstones. For these selected modifications, the
inspectors reviewed the design inputs, assumptions, and design calculations to
determine the design adequacy. The inspectors also reviewed field change notices that
were issued during the installation to confirm that the problems associated with the
installation were adequately resolved. In addition, the inspectors reviewed the post-
modification testing, functional testing, and instrument and relay calibration records to
determine readiness for operations. Finally, the inspectors reviewed the affected
procedures, drawings, design basis documents, and UFSAR sections to verify that the
affected documents were appropriately updated.
For the accessible components associated with the modifications, the inspectors also
walked down the systems to detect possible abnormal installation conditions.
The inspectors reviewed condition reports (CRs) associated with plant modification
issues to ensure that Entergy was identifying, evaluating, and correcting problems
associated with these areas and that the planned or completed corrective actions for the
issues were appropriate. The inspectors also reviewed self-assessments related to
plant modification activities at Pilgrim. The listing of the condition reports and self
assessments reviewed is provided in The attachment.
b.
Findings
No findings of significance were identified.
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Enclosure
1R19
Post-Maintenance Testing (71111.19)
a.
Inspection Scope (6 samples)
The inspector reviewed post-maintenance test activities on risk significant systems to
verify that the effect of the test on the plant had been evaluated adequately, the test was
properly performed in accordance with procedures, the test data met the required
acceptance criteria, and the test activity was adequate to verify system operability and
functional capability following maintenance. The inspector assessed whether systems
were properly restored following testing and that discrepancies were appropriately
documented in the corrective action process. The inspection activity represented six
samples:
Post Work Test for MR05102967 and 05102970 for B RHR System, 4/1/05
PWT per 8.5.3.14 for MR 01108097 following B RBCCW HX Overhaul
PWT for MRs P0000060, 03111321, P0000052 MSIV AO-203-1D actuator,
springs, and packing replacement per 8.7.1.6 for local leak rate testing and
8.I.11.21 for valve stroke timing and fail safe operation.
PWT for MR 031119011 MSIV AO-203-2A packing replacement per 8.7.1.6 for
local leak rate testing and 8.I.11.21 for valve stroke timing and fail safe
operation.
PWT for MRs 03109316, 03109315, 03109315 Rev 1(CR 200502219),
03109314, 03109387, for local leak rate testing per 8.7.1.5 of the feedwater
check valves following soft seat replacement and refurbishment.
Procedure 3.M.3-24.16, Quick look Operations Procedure; for static and
dynamic testing of motor-operated valve (MO) MO-1001-29A (CR 200501820).
b.
Findings
No findings of significance were identified.
1R20
Refueling and Other Outage Activities (71111.20)
a.
Inspection Scope (1 sample)
1.
Review of Outage Plan
The inspector reviewed the RFO-15 outage Shutdown Risk Assessment and procedure
TP05-002, RFO15 Compensatory Measures, to verify that Entergy addressed the
outage impact on defense-in-depth for the five shutdown critical safety functions:
electrical power availability, inventory control, decay heat removal, reactivity control, and
containment. The inspector reviewed how Entergy provided adequate defense-in-depth
for each safety function, and the planned contingencies to minimize the overall risk
where redundancy was limited or not available. Consideration of operational experience
was also assessed. The inspector periodically reviewed the daily risk up-date,
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Enclosure
accounting for schedule changes and unplanned activities. The references used during
this review are listed in the attachment to this report.
2.
Monitoring of Plant Shutdown and Cooldown Activities
The inspector reviewed Entergys action to shut the plant down in accordance with
procedures 2.1.14, Station Power Changes, and 2.1.5, Controlled Shutdown from
Power, 2.2.19.1, Residual Heat Removal-Shutdown Cooling Mode of Operation, and
2.2.85, Augmented Fuel Pool Cooling. Portions of various activities to place the plant
in a cold shutdown condition and on shutdown cooling were observed by the inspector to
assess operator performance, communications, command and control and procedure
adherence. The inspector reviewed the reactor vessel cool down rate, recorded per
OPER-07 and 2.1.7, Vessel Heat up and cool down, to determine whether it was within
technical specification requirements. Other references used are listed in the attachment
to this report.
The inspector also conducted inspection walkdowns of plant areas not normally
accessible during plant power operations (drywell, condenser bay, and main steam
tunnel) to verify the integrity of structures, piping and supports, and to confirm systems
appeared functional.
3.
Fuel Shuffle Activities and Reactivity Control
The inspector reviewed refueling activities to determine whether they were conducted in
accordance with the technical specifications and procedure 4.3, Fuel Handling. The
inspector independently reviewed, on a sampling basis, core alteration activities. The
inspector observed core alterations to assess whether core reactivity was properly
controlled. The inspector observed activities from the control room and the refueling
floor at various times. The inspector determined whether the location of fuel and core
components was tracked in accordance with the fuel movement schedule. The inspector
reviewed Entergys actions to meet the requirements of Technical Specification 3.10 for
core alterations, including the requirements for core monitoring using the source range
monitors and the functional checks of the refueling interlocks. The inspector reviewed
Entergys use of and technical bases for alternate core quadrant definitions as described
in procedure 4.3. The inspector observed communications and the coordination of
activities between the control room and the refueling floor while fuel handling activities
were in progress. The inspector independently reviewed Entergys action to verify the
proper core loading per procedure 4.5. Other references used during this review are
described in the attachment to this report.
4.
Control of Outage Activities
Outage Risk
The inspectors performed routine daily checks of the outage risk profile and determined
whether Entergy maintained defense-in-depth for the key safety functions
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Enclosure
commensurate with the outage risk control plan and that control room operators were
kept cognizant of the plant configuration.
Clearance Activities
The inspectors reviewed a sample of risk significant clearance activities to verify
whether tags were properly hung and/or removed, equipment was appropriately
configured per the clearance requirement, and that the clearance did not impact
equipment credited to meet the shutdown critical safety functions.
RBCCW system clearances: 30-0002, 29-0001-G, 30-0009-C
Inventory Control
The inspector reviewed Entergys actions to establish, monitor and maintain the proper
water inventory in the reactor during the outage, and in the reactor and spent fuel pool
after flooding the reactor cavity for refueling activities. The inspector reviewed the plant
system flow paths and configurations established for reactor makeup and determined
whether the configurations were consistent with the outage plan. The inspector
reviewed Entergys evaluations and corrective actions related to Condition Report
200501535.
The inspector reviewed the implementation of Entergys procedures for foreign material
exclusion control (FME) for the open reactor vessel, reactor cavity and spent fuel pool.
The inspector reviewed Entergys actions to verify that FME issues were documented
and resolved. References used for this review are described in the attachment to this
report.
Electrical Power
The inspector evaluated the status of electrical systems to determine whether they met
the technical specifications requirements and Entergys outage risk control plan. The
inspector reviewed the work plans for the switchyard during the A6 bus outage while the
shutdown transformer and the station blackout diesel generator were unavailable.
Decay Heat Removal (DHR) System Monitoring
The inspector observed spent fuel pool (SFP) and reactor decay heat removal system
status and operating parameters to determine whether the cooling systems operated
properly. The review included periodic review of SFP & reactor cavity level,
temperature, and RHR flow . The inspector conducted partial system walkdowns to
determine whether the proper system configuration was established for alternate spent
fuel pool cooling following an RHR train swap. The inspector also determined whether
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Enclosure
procedures were in place to establish alternate decay heat removal systems (i.e.
augmented fuel pool cooling) to recover from a loss of SFP cooling.
Containment Control
The inspector reviewed Entergys activities during the outage to control primary and
secondary containment and to clean and prepare the containment for closure prior to
plant restart. The inspector performed a walkdown of the drywell prior to reactor startup
to review cleanup and demobilization controls in areas where work was completed to
assure that tools, materials and debris were removed. This review focused on the
removal of debris which might impact the performance of the safety systems.
5.
Monitoring Plant Startup, Heatup and Approach to Critical
The inspector observed operator performance during the plant startup activities during
the period of May 8 through May 12, 2005. The inspection consisted of control room
observations, plant walkdowns and a review of the operator logs, plant computer
information, station procedures 2.1.1, Startup from Shutdown, and 2.1.14, Station
Power Changes. The inspector observed the approach to critical on May 11, 2005.
The inspector assessed whether Entergy met the Technical Specification requirements
during heatup and startup activities. The inspector assessed whether Entergy met the
Technical Specification requirements for compliance with the banked position withdrawal
sequence (BPWS) and the rod worth minimizer.
The inspector reviewed plant restart activities in accordance with procedure 2.1.1 to
determine whether, on a sampling basis, technical specifications, license conditions, and
other requirements for mode changes were met. The inspector evaluated whether
reactor coolant system (RCS) integrity was maintained throughout the restart process by
periodically reviewing RCS leakage calculations and by review of systems that monitor
conditions inside the containment.
The inspector reviewed the test results for the in-sequence shutdown margin
determination to determine whether the calculated test results per 9.16.1, In-Sequence
Critical For Shutdown Margin Demonstration, met the technical specification
requirements.
6.
Problem Identification and Resolution
The inspectors reviewed condition reports to determined whether Entergy was
identifying outage related issues and had entered them into the corrective action
program. The inspectors reviewed a sample of the corrective actions to verify they were
appropriate to resolve the issues. The references used in this review are listed in the
attachment to this report.
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Enclosure
b.
Findings
No findings of significance were identified.
1R22
Surveillance Testing (71111.22)
a.
Inspection Scope (6 samples)
The inspector observed and/or reviewed surveillance testing results to determine
whether the test acceptance criteria was consistent with Technical Specifications and
related Performance Indicators, that the test was performed in accordance with the
written procedure, the test data was complete and met procedural requirements, and
the components were capable of performing their intended safety functions. The
inspection activity represented six samples:
Procedure 8.10.1, Refueling Interlocks Functional Test, 4/20/05
Procedure 3.M.3-24.16, Quick look Operations Procedure, for static and dynamic
testing of motor operated valve (MO) MO-1400-25A.
Procedure 8.7.1.6, Local Leak Rate Testing of the Main Steam Isolation Valves.
Procedure 8.7.1.5, Local Leak Rate Testing of Primary Containment Penetrations
and Isolation Valves, for MO-1001-28A.
Procedure 8.M.3-1, Special Test for Automatic ECCS Load Sequencing of Diesels
and Shutdown Transformer with Simulated Loss of Off-Site Power and Special
Shutdown Transformer Load Test.
2.1.8.5 & 2.1.8.3, Class 1 Reactor Pressure Test, 5/9/05.
b.
Findings
No findings of significance were identified.
1R23
Temporary Plant Modifications (71111.23)
a.
Inspection Scope (1 sample)
The inspector reviewed temporary alteration 05-1-026, installed per procedure 3.M.2-
40, Refuel Outage Temporary Alteration Reactor Shutdown/Floodup Level Indicator.
A walkdown was performed to determine whether temporary equipment was installed in
accordance with the work instructions. The inspector reviewed applicable category A
drawings to determine whether they were up-to-date with the temporary alteration.
Alignment data for the temporary indicator and detector was reviewed to determine
whether it was within the established acceptance criteria. The inspector observed the
detectors response to actual reactor level changes.
b.
Findings
No findings of significance were identified.
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Enclosure
Cornerstone: Emergency Preparedness (EP)
1EP2 Alert and Notification System (ANS) Testing (71114.02)
a.
Inspection Scope (1 sample)
An onsite review of Entergys ANS was conducted to ensure prompt notification of the
public for taking protective actions. During the inspection at Pilgrim, the inspector
reviewed the test and maintenance documentation for the siren system. Condition
reports (CRs) generated as a result of siren testing were reviewed for causes, trends
and corrective actions. The inspection was conducted in accordance with NRC
Inspection Procedure 71114, Attachment 02, and the applicable planning standard,
10 CFR 50.47(b)(5) and its related 10 CFR 50, Appendix E requirements were used as
reference criteria.
b.
Findings
No findings of significance were identified.
1EP3 Emergency Response Organization (ERO) Augmentation Testing (71114.03)
1.
Inspection Scope (1 sample)
A review of Pilgrims ERO augmentation staffing requirements and the process for
notifying the ERO was conducted to ensure the readiness of key staff for responding to
an event and timely facility activation. The inspector reviewed procedures and CRs
associated with the ERO notification system and process. The inspection was
conducted in accordance with NRC Inspection Procedure 71114, Attachment 03, and
the applicable planning standard, 10 CFR 50.47(b)(2) and its related 10 CFR 50,
Appendix E requirements were used as reference criteria.
2.
Findings
No findings of significance were identified.
1EP4 Emergency Action Level (EAL) Revision Review (71114.04)
a.
Inspection Scope (1 sample)
Prior to this inspection, the NRC had received and acknowledged the changes made to
the Pilgrim Emergency Plan and implementing procedures. These changes were made
by Entergy in accordance with 10 CFR 50.54(q), after Entergy had determined the
change did not result in a decrease in effectiveness of the Plan and concluded that the
changes continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR 50. The inspector conducted a sampling review of the changes which could
potentially result in a decrease in effectiveness. This review does not constitute an
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Enclosure
approval of the changes and, as such, the changes are subject to future NRC
inspection. The inspection was conducted in accordance with NRC Inspection
Procedure 71114, Attachment 4, and the applicable requirements in 10 CFR 50.54(q)
were used as reference criteria.
b.
Findings
No findings of significance were identified.
1EP5 Correction of Emergency Preparedness Weaknesses and Deficiencies (71114.05)
a. Inspection Scope (1 sample)
The inspector reviewed CRs initiated by Pilgrim from drills, tests, and audits and the
associated corrective actions to determine the significance of the issues and to
determine if repeat problems were occurring. A list of the CRs reviewed are contained
in the attachment to this report. Also, the 2003 and 2004 audit reports were reviewed
to assess Pilgrims ability to identify issues, assess repetitive issues and the
effectiveness of corrective actions through their independent audit process. This
inspection was conducted according to NRC Inspection Procedure 71114, Attachment
05, and the applicable planning standard, 10 CFR 50.47(b)(14) and its related 10 CFR 50, Appendix E requirements were used as reference criteria.
b.
Findings
No findings of significance were identified.
2.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety (OS)
2OS1 Access Control to Radiologically Significant Areas (71121.01)
3.
On-site Inspection dates April 25 - 29, 2005
a.
Inspection Scope (7 samples)
The inspector reviewed radiological work activities and practices and procedural
implementation during observations and tours of the facilities and inspected
procedures, records, and other program documents to evaluate the effectiveness of
Pilgrims access controls to radiologically significant areas. This inspection activity
represents the completion of seven samples relative to this inspection area (i.e.,
inspection procedure sections 02.02.a thru d and 02.04.a thru c) in partial fulfillment of
the annual inspection requirements.
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Enclosure
Plant Walkdowns and RWP Reviews (02.02.a thru d)
During the inspection week of April 25, 2005, the Pilgrim plant was in the middle of a
scheduled refueling outage. The inspector reviewed the work activities taking place in
the radiologically controlled area to identify any exposure significant work areas within
radiation areas, high radiation areas (<1 Roentgen/hour), or airborne radioactivity
areas in the plant. For selected exposure significant work areas, the inspector
reviewed associated controls and surveys of those areas to determine if controls (e.g.,
surveys, postings, barricades) were acceptable. When possible, the inspector, with a
survey instrument, walked down those areas or their perimeters to determine whether
prescribed radiation work permits (RWPs), procedures, and engineering controls were
in place, whether surveys and postings were complete and accurate, and whether air
samplers were properly located. The inspector reviewed the radiation work permits
(RWPs)(as listed in the List of Documents Reviewed section) used to access these
and other high radiation areas and identified what work control instructions or control
barriers had been specified. The inspector reviewed electronic-personal-dosimeter
(EPD) alarm set points (both for the integrated dose and for the dose rate) for
conformity with survey indications and plant policy. The inspector contacted workers
to determine whether they knew what actions were required when their EPD
noticeably malfunctions or alarms. Also, the inspector reviewed radiation work permits
(RWPs) to identify any airborne radioactivity areas with the potential for individual
worker internal exposures of greater than 50 millirems (committed effective dose
equivalent). The inspector focused on work areas with a history of, or the potential for,
airborne transuranic radioactivity.
Job-In-Progress Reviews (02.04.a thru c)
During the inspection week of April 25, 2005, the inspector reviewed and observed
work activities on several radiation work permits (RWPs) including Numbers 05-0054,
05-0071, 05-0080, 05-0082, and 05-0101 (as listed in the List of Documents Reviewed
section). The inspector reviewed all radiological job requirements (RWP requirements
and work procedure requirements) and attended the RWP pre-job briefing on April 26,
2005 for the dryer weld repair work. During these reviews, the inspector determined
whether the radiological conditions in the work area were being adequately
communicated to workers through briefings and postings. The inspector reviewed
radiological controls including surveys, radiation protection job coverage,
contamination controls, and consideration of dosimetry in high radiation work areas
with significant dose rate gradients to determine whether they were adequate.
Related Activities
During the inspection week of April 25, 2005, the inspector observed Radiologically-
Controlled Area (RCA) entries and exits being made by radiation workers at the
primary RCA access control point to determine whether they complied with
requirements for RCA entry and exit, wearing of record dosimetry, and issuance and
use of alarming electronic radiation dosimeters. The inspector toured various
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Enclosure
elevations in the drywell and reactor building during this refueling outage to determine
whether the radiological controls being implemented were adequate. The inspector
reviewed observed work activities for compliance with the radiation work permit (RWP)
requirements. During these observations and tours the inspector reviewed, for
regulatory compliance, the posting, labeling, barricading, and level of radiological
access control for locked high radiation areas (LHRAs), high radiation areas (HRAs),
radiation and contamination areas, and radioactive material areas.
The inspector performed a selective examination of documents (as listed in the List of
Documents Reviewed section) to evaluate the adequacy of radiological controls. The
review in this area was against criteria contained in 10 CFR 19.12, 10 CFR 20
(Subparts D, F, G, H, I, and J), Technical Specifications, and procedures.
Identification and Resolution of Problems
During the inspection week of April 25, 2005, the inspector selected fourteen
issues/condition reports (CRs) identified in the Corrective Action Program (CAP) for
detailed review (i.e., CR-PNP-2004-02331, -03585, and -03625 and CR-PNP-2005-
00086, -00810, -00853, -00857, -00930, -01318, -01351, -01367, -01370, -01627, and
-01778). The documented reports for the issues were reviewed to determine whether
the full extent of the issues were identified, appropriate evaluations were performed,
and appropriate corrective actions were specified and prioritized.
b.
Findings
No findings of significance were identified.
4.
On-site Inspection dates June 13 - 16, 2005
a.
Inspection Scope (7 samples)
The inspector reviewed radiological work activities and practices and procedural
implementation during observations and tours of the facilities and inspected
procedures, records, and other program documents to evaluate the effectiveness of
Pilgrims access controls to radiologically significant areas. This inspection activity
represents the completion of seven samples relative to this inspection area (i.e.,
inspection procedure sections 02.02.f, 02.03.a thru d, and 02.05.a and b) in partial
fulfillment of the annual inspection requirements.
Plant Walk Downs and RWP Reviews (02.02.f)
During the inspection week of June 13, 2005, the inspector examined Entergys
physical and programmatic controls for highly activated or contaminated materials
(non-fuel) stored within spent fuel and other storage pools.
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Enclosure
Problem Identification and Resolution (02.03.a, b, c, and d)
During the inspection week of June 13, 2005, the inspector reviewed Entergys self-
assessment activities for any results related to the access control program since the
last inspection. The intent of this review was to determine if identified problems are
entered into the corrective action program for resolution. The inspector also reviewed
corrective action reports related to access controls and included in this review any
high radiation area radiological events that have occurred since the last inspection in
this area. The inspector selected eight CRs identified in the CAP for detailed review
(i.e., CR-PNP-2005-02021, -02151, -02175, -02205, -02221, -02793, -02796, -
02903). The inspector discussed the corrective action reports with several members
of the radiological protection staff to determine whether the follow-up activities were
being conducted in an effective and timely manner commensurate with their
importance to safety and risk. There were no self-assessments, conducted since the
last inspection, which covered health physics access controls directly. Also, there
were no Entergy Performance Indicator events or documentation packages for the
Occupational Exposure Cornerstone which required review.
High Risk Significant, High Dose Rate HRA and VHRA Controls (02.05.a and b)
During the inspection on the week of June 13, 2005, the inspector met at various
times with several radiation protection supervisors and discussed the controls and
procedures for high-dose-rate high radiation areas (HRAs) and for very high radiation
areas (VHRAs). The inspector reviewed the subject procedures (as listed in the List of
Documents Reviewed section) to determine whether the level of worker protection was
adequate.
Related Activities
The inspector performed a selective examination of documents (as listed in the List of
Documents Reviewed section) to evaluate the adequacy of radiological controls. The
review in this area was against criteria contained in 10 CFR 19.12, 10 CFR 20
(Subparts D, F, G, H, I, and J), Technical Specifications, and Entergys procedures.
b.
Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (71121.02)
5.
On-site Inspection dates April 25 - 29, 2005
a.
Inspection Scope (3 samples)
The inspector reviewed the effectiveness of Entergys program to maintain
occupational radiation exposure as low as is reasonably achievable (ALARA). This
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Enclosure
inspection activity represents the completion of three samples relative to this
inspection area (i.e., inspection procedure sections 02.01.b, 02.03.b, and 02.04.a.1) in
partial fulfillment of the biennial inspection requirements.
Inspection Planning (02.01.b)
During the inspection week of April 25, 2005, the inspector reviewed the work being
conducted during the current refueling outage (RFO 15). The inspector examined the
outage work scheduled during this inspection period and the associated work activity
exposure estimates and the historical work activity data. The inspector selected work
activities which were likely to result in the highest personnel collective exposures. The
selected work activities/RWPs are described in the List of Documents Reviewed
section.
Verification of Dose Estimates and Exposure Tracking Systems (02.03.b)
The inspector reviewed, during the inspection week of April 25, 2005, Entergys
method for adjusting exposure estimates, or replanning work, when unexpected
changes in scope or emergent work were encountered. The dryer repair evolution
involving diving in the separator/dryer pool was an example of significant emergent
work during this outage. The inspector reviewed in-progress ALARA reviews for
RWPs 05-0054, -0080, -0082, -0099, and -0101 (as listed in the List of Documents
Reviewed section) to determine whether adjustments to estimated exposure (intended
dose) were based on sound radiation protection and ALARA principles and not just
adjusted to account for failures to control the work.
Job Site Inspections and ALARA Control (02.04.a.1)
Based on scheduled work activities during the inspection week of April 25, 2005 and
the associated exposure estimates, the inspector selected work activities in radiation
areas, airborne radioactivity areas, or high radiation areas for observation. The
inspector concentrated on work activities that presented the greatest radiological risk
to workers, for example, the work that was estimated to result in the highest collective
doses, the diving activities to repair the dryer tie-rods, and the under vessel work. The
inspector evaluated Entergys use of ALARA controls for these work activities. The
inspector accomplished this by evaluating Entergys use of engineering controls to
achieve dose reductions and whether the procedures and controls were consistent
with Entergys ALARA reviews.
Related Activities
On April 26, 2005, the inspector observed a pre-job radiological briefing for the diving
evolution to repair the dryer in the separator/dryer pool on the refueling floor of the
reactor building. This briefing included a detailed discussion of the ALARA
recommendations. On April 27, the inspector attended a site ALARA committee
meeting which addressed the dryer repair work activity for which the dose estimate
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Enclosure
was greater than one person-rem. The inspector also noted that the actual outage
dose to date was tracking slightly below the estimated outage dose to date.
The inspector performed a selective examination of documents (as listed in the List of
Documents Reviewed section) for regulatory compliance and for adequacy of control
of radiation exposure. The review was against criteria contained in 10 CFR 20.1101
(Radiation protection programs), 10 CFR 20.1701 (Use of process or other
engineering controls), and procedures.
b.
Findings
No findings of significance were identified.
6.
On-site Inspection dated June 16 - 19, 2005
a.
Inspection Scope (3 samples)
The inspector reviewed the effectiveness of Entergys program to maintain
occupational radiation exposure as low as is reasonably achievable (ALARA). This
inspection activity represents the completion of three samples relative to this
inspection area (i.e., inspection procedure sections 02.02.a, b, and c) in partial
fulfillment of the biennial inspection requirements.
Radiological Work Planning (02.02.a, b, and c)
During the inspection on the week of June 13, 2005, the inspector obtained from
Entergy a list of work activities ranked by actual/estimated exposure that had been
completed during the outage of earlier this year (refueling outage 15) and selected the
three work activities/radiation work permits (RWPs) of highest exposure significance.
These RWPs involved work activities connected with in-service inspection of reactor
vessel nozzles in the drywell, with electrical work on motor-operated valves in the
drywell, and with work activities on the refuel floor connected with reactor disassembly
and reassembly. The inspector reviewed the ALARA work activity evaluations,
exposure estimates, and exposure mitigation requirements to determine whether
Entergy had established procedures, engineering and work controls, based on sound
radiation protection principles, to achieve occupational exposures that were ALARA.
The inspector also reviewed to determine whether Entergy had reasonably grouped
the radiological work into work activities, based on historical precedence, industry
norms, and/or special circumstances. The inspector compared the results achieved
(dose rate reductions, person-rem used) with the intended dose established in
Entergs ALARA planning for these work activities. The inspector reviewed the
reasons for any inconsistencies between intended and actual work activity doses.
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Enclosure
Related Activities
The inspector performed a selective examination of documents (as listed in the List of
Documents Reviewed section) for regulatory compliance and for adequacy of control
of radiation exposure. The review was against criteria contained in 10 CFR 20.1101
(Radiation protection programs), 10 CFR 20.1701 (Use of process or other
engineering controls), and Entergys procedures.
b.
Findings
No findings of significance were identified.
2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03)
1.
On-site Inspection dates April 25 - 29, 2005
a.
Inspection Scope (2 samples)
The inspector reviewed the program for health physics instrumentation to determine
the accuracy and operability of the instrumentation. This inspection activity represents
the completion of two samples relative to this inspection area (i.e., inspection
procedure sections 02.04.b and c) in partial fulfillment of the biennial inspection
requirements.
Problem Identification and Resolution (02.04.b and c)
During the inspection week of April 25, 2005, the inspector reviewed corrective action
program reports related to exposure significant radiological incidents that involved
radiation monitoring instrument deficiencies since the last inspection in this area.
During this review, the inspector looked at activities such as problem identification,
characterization, tracking, and the identification and implementation of corrective
actions which would achieve lasting results. The inspector performed this examination
to determine if those activities were being conducted in an effective and timely manner
commensurate with their importance to safety and risk. There were no recent self-
assessment activities which addressed radiation monitoring instrument deficiencies.
Related Activities
During the tours of the drywell and reactor conducted during this inspection week of
April 25, 2005, the inspector examined the calibration status and operability of
selected radiation protection equipment in use in the plant. Also, the inspector
performed a selective examination of documents (as listed in the List of Documents
Reviewed section) for regulatory compliance and adequacy in this area. The review
was against criteria contained in 10 CFR 20.1501, 10 CFR 20 Subpart H, Technical
Specifications, and procedures.
22
Enclosure
b.
Findings
No findings of significance were identified.
2.
On-site Inspection June 13 - 16, 2005
a.
Inspection Scope (2 samples)
The inspector reviewed the program for health physics instrumentation and protective
equipment to determine the accuracy and operability of the instrumentation and of the
equipment. This inspection activity represents the completion of two samples relative
to this inspection area (i.e., inspection procedure sections 02.06.a and b) in complete
fulfillment of the biennial inspection requirements.
Self-Contained Breathing Apparatus (SCBA) Maintenance and User Training (02.06.a
and b)
During the inspection on the week of June 13, 2005, the inspector reviewed the status
and surveillance records of SCBA staged and ready for use in the plant against
procedural requirements. The inspector also examined Entergys capability for refilling
and transporting SCBA air bottles to and from the control room and operations support
center during emergency conditions. Additionally, the inspector evaluated whether
control room operators and other emergency response and radiation protection
personnel were trained and qualified in the use of SCBA (including personal bottle
change-out). Additionally, the inspector reviewed whether only personnel who
possess manufacturer-certified training/qualifications were allowed to perform
maintenance and repairs on SCBA components vital to the units function. Entergy
stated that all maintenance and repairs on SCBA components vital to the units
function were performed by a vendor. Entergy provided documentation from the
SCBA manufacturer stating that the subject vendor was an authorized distributor and
service center for their SCBAs. The inspector reviewed available vital component
maintenance records for three SCBA units currently designated as ready for service.
Also, the inspector reviewed the records used to ensure that the required, periodic air
cylinder hydrostatic testing was documented and up to date and that the DOT-
required-retest air cylinder markings were in place.
Related Activities
During the tours of the reactor and radioactive waste buildings conducted during the
inspection week of June 13, 2005, the inspector examined the calibration status and
operability of selected radiation protection equipment in use in the plant. Also, the
inspector performed a selective examination of documents (as listed in the List of
Documents Reviewed section) for regulatory compliance and adequacy in this area.
The review was against criteria contained in 10 CFR 20.1501, 10 CFR 20 Subpart H,
Technical Specifications, and Entergys procedures.
23
Enclosure
b.
Findings
No findings of significance were identified.
4.
OTHER ACTIVITIES [OA]
4OA1 Performance Indicator (PI) Verification (71151)
a.
Inspection Scope (3 samples)
The inspector reviewed Entergys procedure for developing the data for the EP PIs
which are: (1) Drill and Exercise Performance (DEP); (2) ERO Drill Participation; and
(3) ANS Reliability. The inspector also reviewed Entergys drill/exercise reports,
training records and ANS testing data to verify the accuracy of the reported data.
Data generated since the April 2004 EP PI verification was reviewed during this
inspection. Therefore, data from the second, third and fourth quarters of 2004 and the
first quarter of 2005 were reviewed. The review was conducted in accordance with
NRC Inspection Procedure 71151. The acceptance criteria used for the review were
10 CFR 50.9 and NEI 99-02, Revision 1, Regulation Assessment Performance
Indicator Guideline.
b.
Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
1.
Routine Review of Corrective Action Program Issues
a.
Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of
Problems, the inspector performed a screening of each item entered into Entergys
corrective action program. This review was accomplished by reviewing printouts of
each condition report, attending daily screening meetings and/or accessing Entergys
database. The purpose of this review was to identify conditions such as repetitive
equipment failures or human performance issues that might warrant additional follow-
up.
b.
Findings
No findings of significance were identified.
24
Enclosure
2.
Corrective Action Program Semi-annual Trend Review
a.
Inspection Scope
As required by Inspection Procedure 71152, Identification and Resolution of
Problems, the inspector performed the semi-annual trend review to identify trends,
either Entergy or NRC identified, that might indicate the existence of a more significant
safety issue. Included within the scope of this review were condition reports from June
2004 - June 2005, the 3rd and 4th quarter 2004 corrective action trend reports, and the
daily plant status report listings of operations equipment problems, operability
evaluations, and temporary alterations.
b.
Findings
No findings of significance were identified. No trends were noted which suggests the
presence of a more significant safety issue. The majority of the repetitive issues /
trends identified by the inspector had also been recognized by Entergy and were
captured in adverse trend CRs, including an emerging adverse trend in emergency
diesel generator performance (CR 2005-1058) that is currently being evaluated by
Entergy. Two trends noted by the inspector not captured in an adverse trend CR were
related to tracking and maintaining current required personnel qualifications and
observing scaffolding requirements.
4OA3 Event Follow-up (71153)
1.
Licensee Event Report Review and Closeout (2 samples)
a.
(Closed) LER 05000293/2005-01, High Pressure Coolant Injection System Inoperable
due to Fuse Failure in Valve Control Circuit. The inspector reviewed Entergys actions
associated with Licensee Event Report (LER) 50-293/2005-01. Entergys actions were
addressed in the corrective action program as Condition Report 20050517. This event
was similar to the events reported in LERs 2004-02 and 2002-01. The event was also
described in NRC Report 2005-006 for which a Green NCV was identified. The
inspector reviewed Entergys actions to inspect and replace potentially susceptible
fuses of the type caused the event. The LER provided an accurate description of the
event and followup actions, taken or planned, were appropriate to address the event
cause. This LER is closed.
b.
(Closed) LER 05000293/2005-02, One Less Than the Technical Specification
Required Minimum Number of Operable Drywell Pressure Channels due to Licensed
Operator Error. The inspector reviewed Entergys actions associated with Licensee
Event Report (LER) 50-293/2005-02. Entergys actions were addressed in the
corrective action program as Condition Reports 2005-1439 and 200502800. The
inspector reviewed actions to restore the channel to an operable status, assure other
channels remained operable, and to address human performance errors. The LER
provided an accurate description of the event and followup actions, taken or planned,
25
Enclosure
were appropriate to address the event cause. This licensee identified finding involved
a violation of TS 3.2.B Instrumentation that Initiates or Control the Core and
Containment Cooling Systems. The enforcement aspects of the violation are
discussed in Section 4OA7. This LER is closed.
4OA5 Other
1.
TI 2515/163, Operational Readiness of Offsite Power
The inspector performed Temporary Instruction 2515/163, Operational Readiness of
Offsite Power. The inspector collected and reviewed Entergys procedures and
supporting information pertaining to the offsite power system specifically relating to the
areas of offsite power operability, the maintenance rule (10 CFR 50.65), and the
station blackout rule (10 CFR 50.63). The inspector reviewed Entergys training in the
procedures related to offsite power and observed a joint training session between the
Pilgrim Station and the Transmission System operators on June 8, 2005. The
inspector reviewed this data against the requirements of 10 CFR 50.63; 10 CFR 50.65; 10 CFR 50 Appendix A General Design Criterion 17, Electric Power Systems;
and Plant Technical Specifications. This information was forwarded to NRR for further
review.
2.
TI 2515/161 - Transportation of reactor control rod drives in Type A packages
During the inspection week of June 13, 2005, this area was inspected to verify that
Entergys radioactive material transportation program complied with specific
requirements of 10 CFR Parts 20 and 71, and Department of Transportation (DOT)
regulations contained in 49 CFR Part 173. The inspector interviewed Entergy
personnel and determined that entergy had undergone refueling/defueling activities
twice since January 1, 2002. Entergy made five shipments of reactor control rod
drives in Department of Transportation Specification 7A Type A packages during this
time period. The inspector reviewed the documentation for each of these shipments.
No findings of significance were identified.
3.
Closed URI 05000293/2004005-02: Operator Inattentiveness in the Control Room
In a letter (ADAMS accession number: ml051960068) dated July 14, 2005, the NRC
issued a Severity Level III Notice of Violation and Proposed Imposition of Civil Penalty
to Entergy in the base amount of $60,000 associated with a Severity Level III problem.
The Severity Level III problem involved four violations of NRC requirements related to
Technical Specification 5.4.1, 10 CFR Part 50 Appendix B, and 10 CFR Part 26. The
specific violations involved: (1) a Pilgrim control room supervisor sleeping for
approximately four minutes in the control room and therefore being neither alert or
attentive to his duties; (2) a reactor operator observing the sleeping control room
supervisor, but deliberately not taking immediate actions to awaken the control room
supervisor, inform appropriate site personnel and initiate a condition report; (3) a Shift
Manager, in careless disregard of requirements, although taking some actions, not
26
Enclosure
informing appropriate site personnel and initiating a condition report; and (4) the
sleeping control room supervisor not being relieved of duty and for-cause Fitness-for-
Duty tested. VIO 05000293/2005003-002, Inattentive Control Room Supervisor
with Wilfull Inappropriate Response by Other Control Room Licensed Staff .
There were no actual safety consequences resulting from this event because there
were no plant conditions that warranted immediate action.
This Severity Level III Notice of violation closes out the unresolved item (URI
05000293/2004005-02) associated with operator Inattentiveness in the control room.
4.
Review of Third Party Assessment Reports
The inspector reviewed the results of the Pilgrim Plant Evaluation conducted by the
World Association of Nuclear Operators (WANO) in February 2005. The inspector
noted that the WANO assessment results were consistent with the NRC's assessment
of Pilgrim activities.
4OA6 Meetings, Including Exit
On April 29, 2005, the inspector presented the inspection results to Mr. P. Dietrich,
General Manager-Plant Operations, and other members of his staff who
acknowledged the inspection results.
On June 13, 2005, the inspector presented the inspection results to Mr. S. Bethay,
Safety Assessment Director, and other members of the site staff who acknowledged
the inspection results.
On June 23, 2005, the inspectors presented the inspection results to Mr. Robert
Smith, Engineering Director, and other members of Entergy management who
acknowledged the inspection results and confirmed the information reviewed by the
inspectors was not considered proprietary.
On June 24, 2005, the inspector presented the inspection results to Mr. Brian Ford,
Licensing Manager, and other members of his staff. Entergy had no objections to the
NRCs observations. The inspector confirmed that proprietary information was not
provided or examined during the inspection.
On July 7, 2005, the inspector presented the inspection results to Mr. T. Kirwin, Plant
Production Manager, and other members of the site staff who acknowledged the
inspection results. The inspector confirmed that proprietary information was not
provided or examined during the inspection.
4OA7 Licensee-Identified Violations
The following violations of very low safety significance (Green) were identified by
Entergy and are a violation of NRC requirements which met the criteria of Section VI
27
Enclosure
of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited
Violation.
1.
Technical Specification (TS) 3.2.B requires two operable instrument channels per trip
system of drywell pressure instrumentation that initiate the core and containment
cooling system. If one channel of a trip system is inoperable, then the TS requires
Entergy to repair the trip system or place the reactor in a Cold Shutdown Condition
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after this trip system is made or found to be inoperable. Contrary to
the above during plant operations at 87% full power on April 14, 2005, a licensed
operator made drywell pressure transmitter PT-1001-89B inoperable when he
inadvertently closed its isolation valve during a valving operation on the CRD backfill
system. This caused one of the two division B drywell pressure instrument channels
to be inoperable. On April 16, Entergy restored PT-1001-89B to an operable
condition. PT-1001-89B was out of service for a total of 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> and the reactor was
not placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the pressure instrument
channel was made inoperable. This finding is of very low safety significance because
during the time PT-1001-89B was isolated, the other drywell pressure instruments
were operable and would have actuated the emergency core cooling, containment
isolation and diesel generator initiation circuits in response to a high drywell pressure
condition. Entergy documented this issue in the corrective action program in CRs
200501439 and 200502800.
2.
Technical Specification 5.4.1 requires written procedures be established and
implemented. Contrary to the above on June 1, 2005, Entergys procedure for
calibrating emergency diesel generator (EDG) B control relays were not adequately
established and implemented and resulted in a wiring error on a capacitor in the
monitoring circuit for Breaker 52-609, that impacted the availability of the EDG. The
calibration procedure, 3.M.3-1, did not refer to the capacitor even though it had to be
removed to bench calibrate the relay. Work practices were inadequate because
procedure 3.M.3-51 was not used to document lifting the capacitor leads. The faulty
wiring was not identified due to an inadequate post-maintenance testing. Entergy
discovered the error during EDG surveillance testing on June 27, 2005, and required
additional EDG unavailability time to repair the wiring error. This finding is of very low
safety significance because the error affected the alarm circuit only and there was no
impact on EDG function. Entergy documented this issue in the corrective action
program in CR 200503183.
3.
Technical Specification 5.4.1 requires in part that written procedures be established
and implemented covering the activities recommended in Regulatory Guide 1.33,
Revision 2. Contrary to the above, on October 22, 2004, Entergy found valve 30-HO-
43 one-quarter turn open vs closed, as required by procedure 2.2.30, Reactor Building
Closed Cooling Water (RBCCW) System. The incorrect RBCCW valve position
resulted in a 3 gallon per minute leak thru vent valve 30-HO-43 and the A train of
RBCCW to be inoperable for approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> until the vent valve was closed.
The train was inoperable because of the inability to fulfill its thirty-day mission time
without the need to refill the expansion tank. The finding is of very low safety
significance because the condition did not exceed half of the allowable Limiting
28
Enclosure
Condition for Operation (LCO) time. Entergy documented this issue in the corrective
action program in CR 200503265.
4.
Technical Specification 5.4.1 requires in part that written procedures be established
and implemented covering the activities recommended in Regulatory Guide 1.33,
Revision 2. Contrary to the above, on March 28, 2005, Entergy personnel did not
follow procedure 8.7.1.19, Pressure Drop Test of Air Supply for Standby Gas
Treatment (SBGT) System Dampers. Specifically, Entergy personnel did not make
periodic projections to predict whether the final leak rate results would be acceptable
nor did they calculate the final results and compare them to the acceptance criteria
prior to securing from Attachment 1 of the test. As a result, Entergy did not recognize
on March 28 that the B train of SBGT was inoperable due to air system leakage.
Entergy recognized on March 29 that the B train was inoperable. Entry into a Limiting
Condition for Operation (LCO) and mitigative actions to place the system in a fail-safe
configuration were therefore unnecessarily delayed. The finding is of very low safety
significance because the condition did not exceed half of the allowable LCO time.
Entergy documented this issue in the corrective action program in CR 200501130.
ATTACHMENT: SUPPLEMENTAL INFORMATION
A-1
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Entergy personnel:
M. Balduzzi
Site Vice President
S. Bethay
Nuclear Assessment Director
P. Dietrich
Plant Manager
J. Detwiler
Radiation Protection Technician
R. Emmitt
Radiation Protection Specialist (Support)
B. Ford
Licensing Manager
L. Foreaker
Radiological Instruments Supervisor
W. Grieves
Quality Assurance Superintendent
T. Kirwin
Plant Production Manager
K. Larson-Sullivan
Sr. Emergency Planner
B. McDonald
Radiation Protection Specialist (Support)
P. McNulty
Radiation Protection Manager
B. Reynolds
Administrative Specialist
E. Salomon
Sr. Emergency Planner
L. Seehaus
Radiation Protection Technician
T. Sowdon
Emergency Planning Manager
B. Sullivan
Operations Superintendent
P. Sullivan
Sr. Emergency Planner
T. Tetzlaff
Radiation Protection Supervisor
J. Veglia
Programs & Components Manager
T. White
Design Control Manager
S. Wilson
Facilities and Equipment Specialist
G. Zavaski
Radiation Protection Specialist (Projects)
R. Blagborough
Senior Engineer
T. Collis
System Engineer
S. Das
Senior Lead Engineer
J. Falconieri
Senior Engineer
H. Goettsch
Senior Component Engineer
J. Keen
System Engineer
D. Mahesh
Senior Component Engineer
D. Sitkowski
Senior Engineer
R. Smith
Engineering Director
D. Titus
Senior Engineer
J. Yingling
Senior Engineer
D. Young
Senior Engineer
S. Woods
Structural Engineer
A-2
Attachment
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
Adequacy of Entergys Change to LOP/LOCA Testing
Without Seeking a TS Amendment. (Section 1R02)05000293/2005003-002
Inattentive Control Room Supervisor with Willful
Inappropriate Response by Other Control Room Licensed
Staff .
Closed
Operator Inattentiveness in the Control Room
05000293/2005-01 LER High Pressure Coolant Injection System Inoperable due
to Fuse Failure in Valve Control Circuit.
One Less Than the Technical Specification Required
Minimum Number of Operable Drywell Pressure
Channels due to Licensed Operator Error
LIST OF DOCUMENTS REVIEWED
`
References for Sections 1R02 and 1R17
10 CFR 50.59 Applicability Determinations
SEE 1083, Safety Relief Valve O-Ring Seat Equivalency Evaluation, dated 7/17/03
PDC 03-51, 480 VAC Power from B35 Made Permanent, dated 7/23/03
10 CFR 50.59 Screened-out Evaluations
PDC 03-101, Replacement motor for Control Room Recirc Fan, VRF-101B
PDC 03-58, Replace Unit Auxiliary Transformer
PDC 03-090, ATWS Power Supply PS1A Replacement
ER 03105182, EDG, Install Load Shed Switches C6, B17, B18
PDC 03-76, Reactor Recirc Pump Leak Sealant Injection
SEE 1130, Equivalency Evaluation for UPS B44 AC Voltmeter (UPS Output) Selector Switch
Electro Switch Model PR10-910C8-6, Rev. 0
SEE 1030, Equivalency Evaluation for ASCO Solenoid Valve NP 831664E, Rev. 0
TA 04-1-032, Mechanical Gag to Open Reactor Building Exhaust Backdraft Dampers
SEE 1124, Lower Diaphragm Plate (Button), HCU Scram Actuators, P/N 213A8821P35 or
80025A (Scram Isolation Valves), Rev. 0
8.M.3-1, Special Test for Automatic ECCS Load Sequencing of Diesels and Shutdown
Transformer with Simulated Loss of Offsite Power and Special Shutdown Transformer
Load Test, dated 3/10/05
ER 03114086, MOV Modifications, Rev. 0
A-3
Attachment
PDC 03-059, Modification of Supports on the A EDG Air Start 2" Piping at Valves SV-4586
A/B, dated 7/18/03
New Procedure 7.4.64, Process Radiation Monitor Alarm Response, Rev. 0
Audits and Self-Assessments
Assessment Report, Assess the Quality of the Plant Design Changes, Temporary Modification
and Alteration Process at Pilgrim Station, dated 8/15/03
Calculations
PS79, Emergency Diesel Generator Loading, Rev. 5
PS234, Calculations - Scenarios & Load Categories, dated 8/25/99
PS95, Change No. 5, Rev. 2, dated 8/7/03
PS65, Change No. PS65-1-14, Rev. 1, dated 8/25/03
PS65A, Change No. PS65A-0-19, Rev. 0, dated 8/25/03
Calculation No. M1265, Effect of the Removal of the Biological Shield Blocks on the
Containment Temperatures, Rev. 0
PNPS-1-ERHS-XXII.A-2, Radiological of a Design Basis Fuel Handling Accident Using the
Alternate Source Term, Rev. 1
Completed Surveillances
Procedure No. 8.E.29.1, Salt Service Water (SSW) Instrumentation Calibration and Functional
Test, Attachment 2 & 4, dated 9/10/03
Procedure No. 8.E.29.1, Salt Service Water (SSW) Instrumentation Calibration and Functional
Test, Attachment 4, dated 10/12/04
Procedure No. 8.E.30.1, Closed Cooling Water System (CCWS) Instrumentation Calibration
and Functional Test, Attachment 1, dated 7/16/03
Procedure No. 3.M.3-47, Load Shed Relay Operational/Functional Test, Attachment 2, dated
10/18/03
Procedure No. 8.M.3-1, Special Test for Automatic ECCS Load Sequencing of Diesels and
Shutdown Transformer with Simulated Loss of Offsite Power and Special Shutdown
Transformer Load Test, dated 5/9/03 & 4/28/05
Procedure No. 3.M.3-47.2, B Train Functional Test of Individual Load Shed Components,
Attachment 8-9-10, dated 6/11/03
Procedure No. 3.M.3-47.2, B Train Functional Test of Individual Load Shed Components,
Attachment 11, dated 10/7/03
Procedure No. 7.1.30, HEPA Filter and Charcoal Cell Performance Test Program, dated
8/11/04 & 9/24/04
Procedure No. 7.1.44, Sampling of Charcoal Cells in Standby Gas Treatment and Control
Room Environmental Filters Systems for Methyl Iodide Testing, dated 8/10/04 &
9/29/04
Procedure No. 8.7.2.7, Measure Flow and Pressure Drop Across Control Room High Efficiency
Air Filtration System (CRHEAFS), dated 8/11/04 & 9/29/04
Procedure No. 8.5.3.2.1, Salt Service Water Pump Quarterly and Biennial (Comprehensive)
Operability and Valve Operability Tests, dated 9/24/04
A-4
Attachment
Corrective Action Reports
2003-00045
2003-02617
2003-02679
2003-03402
2003-03403
2003-03825
2004-00269
2004-00329
2004-00352
2004-00390
2004-00552
2004-00639
2004-00834
2004-00971
2004-01107
2004-01849
2004-02455
2004-02484
2004-02753
2004-02754
2004-02889
2004-03657
2004-03664
2004-03822
2005-00232
2005-03093
2005-03130
2005-03112
2005-03114
2005-03115
2005-03116
2005-03117
2005-03118
Drawings
E-18, Schematic Diagram Diesel Generator Load Shedding, Rev. E-18
E-173, Schematic Diagram Cooling Water System Turbine Building, Rev. E-4
E52A1, Outline and Dimension Salt Service Pump Motor, Rev. E2
41100-0428, Schematic Drawing Diesel Generator Load Shedding, Rev. E18
E176, Schematic Diagram - Reactor BLDG Closed Cooling Water System, Sh. 1, Rev. E8,
A-14, Turbine Auxiliary & Control BLDG Service Area EL +37'-0", Rev. E7
M8-4, Assembly Drawing Service Water Pump P208A, B, C, D, E; Rev. E19
M8-27, Service Water Pump (P-208A-E) Motor Base, Rev. A
Evaluations
SEE 1089, Controlled Document Change Notice 03-1583
SEE 718, Air Operated Manifold On Main Steam Isolation Valves, AO-203-1A, B, C & D
SEE 1087, Rosemount Master Trip Units
SEE-1092, Robertshaw Pressure Indicators PI-5001 A, B
ER 03120866, Permanent Removal of Drywell Shield Blocks, Rev. 0
SE 2619, Perform Patching and Painting on Panels and Consoles in Main Control
Room During Power Operations, dated 7/26/91
Miscellaneous
Risk-Informed Inspection Notebook For Pilgrim Nuclear Power Station Unit 1, Rev. 1
Regulatory Guide 1.186, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases
NEI 96-07, Guidelines For 10 CFR 50.59 Implementation, Rev. 1
NEI 97-04, Design Basis Program Guidelines, Rev. 1
SDBD-61, System Design Basis Document for the Emergency Diesel Generator (EDG) and
Auxiliary Systems, Rev. E0, dated 10/18/00
SBDB-30B, Design Basis Document for the Turbine Building Closed Cooling Water System,
Rev. E0
ER 03105182; Addition of Test Switches to Load Shed Relays 105XA1, 105XA2 & 105XB2
ER 04114484; Reverse PDC 03-70 and Provide New Cable Types for Signal and HV Cables of
RM-1705-3A
A500024, Simulator Change Evaluation for ER 02116887, dated 5/9/05
LCS Test Report 0100417; Automatic Valve, Vibration Test, Model B5140-202, dated 9/17/01
Transformer Test Report No. R-3185, Replacement Unit Auxiliary Transformer, dated 8/11/03
TR62729-05N, Dedication/Qualification Test Report Meter Device TS P/N 233-3265, Rev. 0
A-5
Attachment
Supplier Design Document Review Form 9324-02, SUDDSRF #03-81 - Test of ATWS Power
Supply, Rev. 1, dated 10/2/03
Modification Closeout Report, PDC 03-76, Reactor Recirc Pump Leak Sealant Injection
Modification Closeout Report, PDC 03-021, Weld Repair and/or Epoxy Repair to TBCCW Heat
Exchanger E-122A
Control Room Logs, dated 6/16/03 - 8/13/03
LI-101 50.59 Program Information update and Lessons Learned, dated 7/7/04
Entergy Letter No. 2.04.115, Proposed License Amendment for a Limited Scope Application of
the Alternate Source Term (NUREG-1465) for Re-evaluation of the Fuel Handling
Accident Dose Consequences, Rev.1, dated 12/15/04
RTYPE E2.15, Maintenance Rule SSC Basis Document Manual, Rev. 5
Modifications
PDC 03-058, Replace Unit Auxiliary Transformer
ER 02116887, Reduce EDG Loading During Load Shed
PDC 03-021,Weld Repair and/or Epoxy Repair to TBCCW Heat Exchanger E-122A
FRN 03-13-01, Allow Threaded Connections Associated with X-141 Seal Oil Unit (PDC 03-13,
Replacement of Level Switch 63-L10)
PDC 02-44, Improving Performance of Control Room Envelope
FRN No. 99-01-10, Salt Service Water Pump/Motor Bases
Operating Experience
NRC Information Notice 98-22: Deficiencies Identified During NRC Design Inspections
GE Operating Experience Report (OER Ref. No. 1032), Biological Shield Plugs, dated 5/31/73
Entergy Response to NRC Information Notice 98-22, dated 6/25/98
Procedures
ENN-LI-100, Process Applicability Determination, Rev. 5
ENN-LI-101,10CFR50.59 Review Process, Rev. 7
ENN-DC-102, Operating Plant Changes and Modification, Rev. 1
ENN-DC-103, Design Process, Rev. 1
ENN-DC-105, Configuration Management, Rev. 1
ENN-DC-115, ER Response Development, Rev. 4
ENN-DC-116, ER Response Installation, Rev. 4
ENN-DC-117, Post Modification Testing and Special Testing Instructions, Rev. 4
ENN-DC-118, ER Response Closure, Rev. 4
ENN-DC-121, Maintenance Rule, Rev. 2
ENN-DC-171, Maintenance Rule Monitoring, Rev. 2
Procedure No. 1.4.75, Infrequently Performed Tests and Evolutions, Rev. 0
Temporary Test Procedure No. TP04-011, Functional Test of Load Shed Modifications to
TBCCW Pumps P-110A and B, completed 10/26/04.
Procedure No. 8.M.3-1, Special Test For Automatic Load Sequencing of Diesels And
Shutdown Transformer With Simulated Loss of Off Site Power and Shutdown
Transformer Load Test, completed 5/8/05
Procedure No. 8.E.30.1, Closed Cooling Water System (CCWS) Instrumentation Calibration
And Functional Test, completed 7/22/03
A-6
Attachment
Procedure No. 3.M.4-14.2, Salt Service Water Pumps Routine Maintenance, completed
9/22/04
Procedure No. 3.M.1-15, Vibration Monitoring for Preventive Maintenance and Balancing,
completed 9/22/04
Procedure No. NE6.02; Control of Drawings, Sketches, and Data Sheets, Rev. 35
Procedure No. 7.4.64, Process Radiation Monitor Alarm Response, Rev. 0
Safety Evaluations
SE 3400, New APRM FCTR Setpoints For Stability Option 1-D
SE 3398, Incorporate the term Operation with the Potential To Drain the Reactor Vessel
(OPDRV) in the UFSAR Section 1.2.
SE 3397, Cycle 15 Reload Core Design, Rev. 0
SE 3401, Evaluation to Identify The Design Basis for Salt Service Water
Pump Intake Water Level Requirements to Ensure Adequate Protection of System
Design Basis Requirements is Provided, Rev. 0
SE 3399, Permanent Removal of Drywell Shield Blocks, Rev. 0
SE 3397, Cycle 15 Reload Core Design with Revised Reference Loading Pattern, Rev. 1
Safety Review Committee (SRC) and On Site Review Committee (OSRC)
SRC OSRC Subcommittee Meeting Minutes dated 11/21/03, 2/12/04, 10/28/04, 4/12/05
OSRC Meeting Minutes dated 2/11/05, 4/4/05, 5/7/05
System Health Reports and Trending Data
Pilgrim Nuclear Power Station System Health Report, 4th Quarter 2004
EDG & Fuel Storage System Health Report, 4th Quarter 2004
Work Orders (MRs)
P9900631
02118469
03102361
03110734
03111257
References for Section 1R04
M241, Residual Heat Removal System, Sheet 1&2
2.2.19.1, Residual Heat Removal System - Shutdown Cooling Mode of Operation
M231, Fuel Pool Cooling & Demineralizer system,
2.2.85.1, Augmented Fuel Pool Cooling (With Shutdown Cooling) Mode 1
2.2.85.2, Augmented Fuel Pool Cooling (Without Shutdown Cooling) Mode 2
2.1.1, Startup From Shutdown
M242, Core Spray System
2.2.20, Core Spray System
M215, Reactor Building Closed Loop Cooling Water System, Sheets 1,2&3
2.2.30, Reactor Building Closed Loop Cooling Water System
References for Section 1R05
5.5.2, Special Fire Procedure
A-7
Attachment
Pilgrim Nuclear Power Station Fire Hazards Analysis
A319, Reactor & Turbine Building Floor Plan El. 51' - 0" & 74' - 3" Fire Barrier System
References for Section 1R08
PIL-R15-05-032, Examination summary sheet for UT of Merid Head Weld RPV-TH-M4
including sizing data and indication evaluation
PIL-R15-05-033, Examination summary sheet fo UT of RPV-TH-M5 weld including sizing data
and indication evaluation
PIL-R15-05-001 UT Examination summary sheets (UT-011, 012, 023 and 024) for Vessel
head to flange weld RPV-HF-240-360
PIL-R15-05-001 Magnetic Particle examination report (MT-001) for RPV-HF-240-360
IR 03-0145, Ultrasonic Examinations of the torus shell
NDE-10.02 R0, VT-2 Examination of annulus drains (IWE exam)
IR 03-0283, VT-3 Examination - IWE general visual walkdown
VT-50-05001 thru VT-3, Examination of containment and support surfaces 50-05009
05-M-307-ISI, MT examination of weld GB-14-VBWJ36A-1
05-M-308-ISI, MT examination of weld GB-14-VBW36A-2
05-M-323-ISI, MT examination of weld HL-23-4-1B
UT-038, UT calibration and examination record for procedure TP04-032
UT-039, UT calibration and examination record for procedure TP04-032
UT-001, 002, UT calibration and examination record for procedure ENN-NDE-9.11
015 & 016
UT-011, 012,UT calibration and examination record for procedure TP04-019
023 & 024
2.1.8.7, ASME Visual Examination of Primary Containment
ENN-NDE-9.31, Magnetic Particle Testing
ENN-NDE-10.02 R0, VT-2 ASME Section XI VT-2 examination of components
ENN-NDE-9.10 R0, UT of dissimilar metal piping welds (ASME XI)
ENN-NDE-9.11 R0, Manual UT of RPV welds (ASME XI)
TP04-014, Invessel Visual Inspection of BWR 3 RPV Internals
MR 05108176, Replace socket weld B-11-79 in SBLC system
ENN-DC-126, Evaluation of embedded flaw in N2F nozzle
PNPS-21Q-310, Wall thickness evaluation for Recirc inlet nozzle N2
1272 Calculation for Recirc inlet nozzle N2
CR-2005-01914, UT of socket weld revealed lack of fusion at weld root location
CR-2005-01888, Jet Pump Wedges #16 and 17 are in the full down position
CR-2005-01890, Metallic debris found in RPV annulus on shroud support ledge
CR-2005-01608, Four tie bars at top of steam dryer noted to be cracked (3, 4, 5 and 9)
CR-2005-01895, Nozzle N2G and N2J weld surface profile conditions
CR-2005-01827, Nozzle N2G weld crown surface conditions
CR-2005-01830, Nozzle N2J wall thickness and weld surface condition
CR-2005-01837, Loose insulation during nozzle work inside the drywell
CR-2005-01860, Unsecured insulation during drywell work
CR-2005-01870, Indication in top head meridional weld RPV-TH-M4
CR-2005-01871, Indication in top head meridional weld RPV-TH-M5
CR-2003-01095, Reduced wall thickness identified at some locations of torus shell
CR-2005-01368, Correction to frequency for refuel floor liner drains
A-8
Attachment
CR-2005-01657, Reduced wall thickness measured on nozzle N2E
CR-2003-01618, IWE containment examination identified some coating failure
CR-2005-01980, Loose nuts/studs on pressure relief valves of personnel air lock
CR-2005-01916, Indication identified in N2F nozzle to safe end weld
CR-2005-02322, Socket weld B-11-79 requires additional evaluation
CR-2005-01770, UT thickness readings on 2R-N2G-1 noted two areas below min wall
INR 015-05-01 R1, Steam dryer tie bars 3, 4, 5 and 9
INR 015-05-02, Steam dryer leveling screws, comparison with exam in 2003
JXEF4-04 RA, Steam dryer repair instructions (tie bar replacement)
JXEF4-07 RB, Revised steam dryer repair instructions (tie bar replacement)
ENN-DC-115 R4, Process for development of engineering requests response
OEN-2005-00110, Tee quencher bolting, moisture separator and decon line leakage
NE15.14, IWE Containment Inspection Program
References for Section 1R11
LORT Training Module #O-RQ-06-02-102, Loss of Feed Water Heating Heater Tube Leak
Procedure 2.1.144, Degraded Voltage
Procedure 2.1.6, Reactor Trip
Procedure 2.4.150, Loss of Feedwater Heating
Procedure 2.2.152, Feedwater Heater, Extraction Steam, and Heater Drains
Procedure 2.1.14, Station Power Changes
Procedure 2.4.49, Feedwater Malfunction
2.4.16, Distribution Alignment Electrical system Malfunction
References for Section 1R15
OE and REO CR 200501136, Potential Safety Limit Violation for Analyzed Operational
Occurrences (GE Part 21 Report SC05-03)
GE Part 21 Report SC05-03 dated March 29, 2005
Condition Report CR-WPO-2005-0071
Condition Reports 200503168, 200503140, 200503151
MR 05109337, Potential Fuse Failure Alarm During B EDG Test Run
Drawings 41100-0443, 41200-3127 and 41100-0427
Procedure 2.2.8, Standby AC Power Systems (Diesel Generators)
2.2.108, Diesel Generator Cooling and Ventilation System
Calculations M991, X-107A/B High Temperature Design Limit
Technical Specifications 3.5.F.1 and 3.9.B.3
UFSAR Sections 8.3 and 8.5
EN#41799, Both Emergency Diesel Generators Inoperable
References for Section 1R16
Procedure 1.3.34.4, Compensatory Measures (CM)
Operator Compensatory Measure Log
CM Evaluation #314, 341, 342, 345, 349, 350, 351, 352, 354, 356, 357, 358, 359, 360, 361
Maintenance Request 04105084, Clean & Inspect ACB 102 Insulator Bushings
Condition Reports 200500826, 200501192, 200501482, 200502888, 200503034, 200503142
A-9
Attachment
ENN-LI-100, Att 9.1, SE Screen for CM #349, 4/14/05
ENN-LI-101, Att 9.1, 5059 Screen for CM #349, 4/20/05
References for Section 1R20
3.M.1-45, Outage Shutdown Risk Assessment
RFO 15 Shutdown Risk Review Report
TP05-002, RFO15 Compensatory Measures
Power Maneuvering Plan PMP-MAN.C15-39
2.1.5, Controlled Shutdown from Power, Rev 89
2.2.19.1, Residual Heat Removal System - Shutdown Cooling Mode of Operation (Rev.13)
2.1.7, Vessel Heatup and Cooldown (Rev 46)
2.4.25, Loss of Shutdown Cooling (Rev 27)
2.2.93, Main Condenser Vacuum System, Rev 50
4.3, Fuel Handling, Rev 99
4.5, Reactor Core Fuel Verification, Rev 19
LCO-1-05-0044, Inoperable Rod Worth Minimizer
LCO-OUT-1-05-0014 and LCO-ACT-1-05-084 for Standby Liquid Control
LCO-OUT-1-05-0075 for Structural Integrity of Primary System Boundary
OPER-07, RPV metal Temperatures and Pressures, 4/18/05
OPER-13, Daily Refueling Checklist
OPER-14, Shift Refueling Checklist
OPER-25, Fuel Movement Within the Spent Fuel Pool Checklist
2.4.31, Reactor Basin an/or Spent Fuel Pool Draindown
2.2.85.1, Augmented Fuel Pool Cooling (With Shutdown Cooling) Mode 1 (Rev 6)
Technical Specification 3.10.A and 4.10 A, Refueling Interlocks
Technical Specification 3.10.B, Core Monitoring
UFSAR Section 7.5.4, Source Range Monitoring
UFSAR Section 7.6, Refuelng Interlocks
INR RFO15-05-04 Foreign Material Indication Notification Report
Procedure 2.1.36, Object Retrieval from Reactor Cavity and Spent Fuel Pool
License Amendment No. 215, Alternate Source Term for Fuel Handling Accident, 4/28/05
NEA-03-052, Revised SRM Quadrant Definitions including Rotation of Quadrants, 4/21/03
GE-NE-0000-0014-5292, Pilgrim SRM Quadrant Definition Analysis, 4/18/03
MR 03108794, Secondary Containment Leak Rate Test
SRM Neutron Flux Response (MR03117386, 03117387, 03117388, 03117389), 4/17/05
MR 03108819, FCU and FME Cover Installation and Removal
Condition Reports 200501482, 200501498, 200501503, 200501666, 200501673, 200501676,
200501688, 200501702, 200501857, 200501890, 200501965, 200501979, 200501981,
200502024, 200502056, 200502139, 200502322, 200502302, 200502356, 200502357,
200502466, 200502468, 200502470, 200502471, 200502472
MR 05107627, Reactor Vessel Debris Removal
OSRC Meeting Minutes 05-06,07, 08, 09, 10 and 11
MAN.C16-01, Power Maneuvering Plan Cycle 16 Startup
References for Section 1R22
Procedure 4.3, Fuel Handling dated 4/20/05
8.10.1, Attachment 1, refueling Interlock Functional Test, 4/19/05 and 4/20/05
A-10
Attachment
8.10.1, Attachment 7, Refueling Interlocks Logic Functional Test, 4/20/05
1.3.34, Attachment 9, Surveillance Test Review, 4/20/05
Technical Specification 3/4.10 and Bases, Core Alterations
UFSAR 7.6, Refueling Interlocks
References for Section 1EP2
EP-AD-417, Annual Siren Test Program, Rev 3
EP-AD-418, Monthly Testing of the Prompt Alert and Notification System, Rev 5
EP-AD-419, Annual Maintenance of the PANS Two-Way System, Rev 2
References for Section 1EP3
PNPS Emergency Plan Section O, Emergency Response Training
Nuclear Training Manual Section 5.5, Emergency Plan Training, Rev 28
NOP88A4, Assignment of Responsibilities in Support of the PNPS EP Program, Rev 8
SCBA EP Qualifications Notification Forms (5/31/05 & 6/23/05)
RP-STD-28, Maintenance of SCBA Qualifications for the EP Program, Rev 1
Procedure No. 6.7-002, Respiratory Protection Program, Rev 9
Procedure No. 6.7.1-104, Issue Use and Return of Respiratory Protection Equipment, Rev 12
EP-IP-100, Emergency Classification and Notification, Rev 23
EP-IP-220, TSC Activation and Response, Rev 13
EP-IP-230, OSC Activation and Response, Rev 4
EP-IP-231, Onsite Radiation Protection, Rev 6
EP-IP-240, Emergency Security Organization Activation and Response, Rev 10
EP-AD-110, Emergency Preparedness Organization and Responsibilities, Rev 3A
EP-AD-122, Maintenance of the Emergency Telephone Directory, Rev 7
EP-AD-125, Maintenance of the ERO, Rev 3
EP-AD-410, Maintenance of the Computerized Automated Notification System, Rev 3
EP-AD-411, Testing of the CANS, Rev 6
Emergency Telephone Directory, Rev 70
NRC Inspection Reports 05000293/2003-007 & 010
References for Section 1EP4
EP-IP-501, Transport of Contaminated Injured Personnel, RETIRED
10CFR50.54(q) Effectiveness Review for EP-IP-501
10CFR50.54(q) Effectiveness Review for EAL 5.3.1.1
10CFR50.54(q) Effectiveness Review for Revision 28 of the PNPS Emergency Plan
Revised Distribution List of EPPI Public Information Emergency Response Procedure Manual
CR-PNP-2005-03109
References for Section 1EP5
Audit Report 03-10, Emergency Preparedness Program
Audit Report QA-07-2004-01, Emergency Preparedness Program
CR-PNP-2003-04584
A-11
Attachment
CR-PNP-2004-01136
CR-PNP-2004-01142
CR-PNP-2004-01144
CR-PNP-2004-01146
CR-PNP-2004-01148
CR-PNP-2004-01150
CR-PNP-2004-01153
CR-PNP-2004-03882
CR-PNP-2005-00581
CR-PNP-2005-01298
CR-PNP-2005-02741
CR-PNP-2005-02815
WRT-053506
References for Section 2OS1
RWP No. 05-0054, Rev. 00, Electrical motor-operated valve work (drywell
elevations 23 and 41)
RWP No. 05-0071, Rev. 00, Dryer upgrade (underwater welding in the
separator/dryer pool)
RWP No. 05-0080, Rev. 00, ISI exams and support for N6A & B core spray
nozzle areas (drywell elevations 74 and 83)
RWP No. 05-0082, Rev. 00, ISI exams and support for N1 & N2 nozzle areas
(drywell elevations 23, 41, and 51)
RWP No. 05-0101, Rev.00, Control rod drive exchange and support (under
vessel)
Procedure EN-RP-104, Rev. 0, Personnel contamination events
RFO-15 drywell field guide by Radiation Protection Department, April 2005
Radiological surveys for elevation 74 of the drywell, 04/18 -27/2005
Skin dose assessment for a personnel contamination on 04/25/2005
Procedure No. 1.3.114, Rev. 19, Conduct of radiological operations
Procedure No. 1.16.1, Rev. 7, Spent fuel pool non-SNM inventory control
Procedure No. 6.1-009, Rev. 9, Radiological controls for handling highly radioactive objects
and refuel floor activities
Procedure No. 6.1-014, Rev. 16, High radiation area control
Procedure No. 6.1-031, Rev. 18, Radiation work permits
A-12
Attachment
Procedure No. 6.3-061, Rev. 17, Radiological survey techniques
Pilgrim Station daily dose report for June 13, 14, 15, and 16, 2005
RWP No. 05-0054, Rev. 00, Electrical motor-operated valve work (drywell elevations 23 and
41)
RWP No. 05-0065, Rev. 04, Reactor disassembly/reassembly and associated support
RWP No. 05-0071, Rev. 00, Dryer upgrade (underwater welding in the separator/dryer pool)
RWP No. 05-0080, Rev. 00, ISI exams and support for N6A & B core spray nozzle areas
(drywell elevations 74 and 83)
RWP No. 05-0082, Rev. 00, ISI exams and support for N1 & N2 nozzle areas (drywell
elevations 23, 41, and 51)
References for Section 2OS2
RWP No. 05-0054, Rev. 00, Electrical motor-operated valve work (drywell
elevations 23 and 41)
RWP No. 05-0071, Rev. 00, Dryer upgrade (underwater welding in the
separator/dryer pool)
RWP No. 05-0080, Rev. 00, ISI exams and support for N6A & B core spray
nozzle areas (drywell elevations 74 and 83)
RWP No. 05-0082, Rev. 00, ISI exams and support for N1 & N2 nozzle areas
(drywell elevations 23, 41, and 51)
RWP No. 05-0101, Rev.00, Control rod drive exchange and support (under
vessel)
Specific RWP dose by date report for 04/18 to 04/25/2005 (listing RWP/task-
description, actual hours to date by RWP, and actual dose to date by RWP)
Radiological engineering spreadsheet for RFO 15 as of 04/25/2005 (listing
component, description, RWP, maintenance request, location, zone, total estimated hours for
completion by RWP, total estimated rem for completion by RWP, and engineering controls)
Daily dose report for RFO-15 for 04/29/2005 (listing work activity description,
actual dose to date by work activity, total estimated dose for completion by work activity, and
percent of estimate)
Pre-job ALARA reviews for RWPs 05-0054, -0071, -0080, -0082, and -0101
RFO-15 steam dryer tie rod repair dose estimate as of 04/26/2005
In-progress ALARA reviews for RWPs 05
-0054 (Electrical motor-operated valve work (drywell elevations 23 and41)),
-0080 (ISI exams and support for N6A & B core spray nozzle areas
(drywell elevations 74 and 83)),
-0082 (ISI exams and support for N1 & N2 nozzle areas (drywell
elevations 23, 41, and 51)),
-0099 (VAC 206B-2 motor/fan replacement (drywell elevation 9)),
-0101 (Control rod drive exchange and support (under vessel))
- ALARA review checklist for jobs greater than one rem
Procedure No. 6.10-020, Rev. 9, ALARA work reviews
Procedure No. 6.10-021, Rev. 6, Station ALARA performance
Procedure No. 6.10-022, Rev. 8, ALARA engineering controls
Procedure No. 6.10-023, Rev. 3, ALARA planning assessments
RWP No. 05-0054, Rev. 00, Electrical motor-operated valve work (drywell elevations 23 and
41)
A-13
Attachment
RWP No. 05-0065, Rev. 04, Reactor disassembly/reassembly and associated support
RWP No. 05-0071, Rev. 00, Dryer upgrade (underwater welding in the separator/dryer pool)
RWP No. 05-0080, Rev. 00, ISI exams and support for N6A & B core spray nozzle areas
(drywell elevations 74 and 83)
RWP No. 05-0082, Rev. 00, ISI exams and support for N1 & N2 nozzle areas (drywell
elevations 23, 41, and 51)
ALARA review packages for RWP Nos. 05-0054, 05-0065, 05-0071, 05-0080, and 05-0082
Refueling outage - 15 dose report for June 14, 2005 by project
Refueling outage -15 dose report by RWP
References for Section 2OS3
Procedure No. 6.4-331, Rev. 16, Operation of common radiation detectors and air samplers
Procedure No. 6.7.1-106, Rev. 11, Inspection and testing of respiratory protection equipment
Procedure No. 6.7.1-201, Rev. 8, Operation of the SCBA air compressor
Monthly SCBA pressure check surveillance status as of June 16, 2005
Weekly emergency respirator inspection report as of June 16, 2005
SCBA EP qualification status notification form
CR-PNP-2005-02990, no supervisory review of SCBA surveillance records per 6.7.1-106, step
8.6.3(3)
CR-PNP-2005-02991, SCBA bottles staged and ready for use with expired hydrostatic testing
dates
CR-PNP-2005-03002, possible shortfall of electrical pool ERO responders with SCBA
qualifications
References Section 4OA1
EP-AD-150, Emergency Preparedness Indicator Tracking Guideline, Rev 2
PNPS Emergency Telephone Directory Rev 70
CR-PNP-2005-03127
References for Section 4OA5
Temporary Instruction 2515/163, Operational Readiness of Offsite Power
Procedure 2.1.144, Degraded Voltage
Procedure 2.1.15, Daily Surveillance Log
Procedure EN-WM-101, On-Line Work Management Process
Procedure EN-WM-100, Work Request (WR) Generation, Screening, and Classification
Procedure 5.3.31, Station Blackout
Procedure 2.2.146, Station Blackout Diesel Generator
Procedure 1.5.22, Risk Assessment Process
NUREG 1022, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73
ISONE Procedure Master/Local control Center Procedure #1, Nuclear Plant Transmission
Operations
SPEC-01-1524, Revision 3, Packaging certification documentation, 7A Type A, Container
Products Corporation, December 21, 2004
USA DOT-7A Type A Container Certification Nos. 1340, 1341, 1342, 1343, 1344, 1347, 1348,
1349, 1350, 1351, and 1352
A-14
Attachment
Radioactive shipment records for reactor control rod drives: RSR Nos.03-321 (two boxes), 03-
323 (four boxes),03-325 (two boxes),05-325 (four boxes), and 05-326 (four boxes)
C-FB-02-02-01, Rev. 5, Self-contained breathing apparatus instructor guide
LIST OF ACRONYMS
As Low As Reasonable Achievable
Alert and Notification System
American Society of Mechanical Engineers
Corrective Action Program
CCWS
Closed Cooling Water System
CFR
Code of Federal Regulations
CR
Condition Report
Condensate Storage Tank
Drill and Exercise Performance
Department Of Transportation
Emergency Action Level
Electronic Personnel Dosimeter
ER
Engineering Request
Emergency Response Organization
Heat Exchange
INR
Indication Notification Report
IR
Inspection Report
Inservice Inspection
LER
Licensee Event Report
Loss of Coolant Accident
MO
Motor Operated
Motor-Operated Valve
Maintenance Request
Main Stream Isolation Valve
Magnetic Particle
NRC
Nuclear Regulatory Commission
OA
Other Activities
Operating Experience
OS
On Site Review Committee
Performance Demonstration Initiative
Performance Indicator
A-15
Attachment
Problem Identification and Resolution
Pilgrim Nuclear Power Station
Penetrant
Post Work Test
QASR
Quality Assurance Surveillance Report
Reactor Building Closed Cooling Water
Radiologically-Controlled Area
ReFueling Outage
Risk Significant Planning Standard
Radiation Work Permit
Self-Contained-Breathing Apparatus
Significant Determination Process
SDT
Shutdown Transformer
Safety Evaluation
SEE
Substitution Equivalency Evaluation
Spent Fuel Pool
Safety Review Committee
Structure, System, and Component
Salt Service Water
TBCCW
Turbine Building Closed Cooling Water
TI
Temporary Instruction
TS
Technical Specification
Transmission System Operators
Updated Final Safety Analysis Report
Ultrasonic Testing
Very High Radiation Area
Visual Test