ML090400437
| ML090400437 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/09/2009 |
| From: | Jack Giessner NRC/RGN-III/DRP/RPB4 |
| To: | Schwartz C Entergy Nuclear Operations |
| References | |
| IR-05-005 | |
| Download: ML090400437 (40) | |
See also: IR 05000255/2008005
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE ROAD, SUITE 210
LISLE, IL 60532-4352
February 9, 2009
Mr. Christopher J. Schwarz
Site Vice President
Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant
27780 Blue Star Memorial Highway
Covert, MI 49043-9530
SUBJECT: PALISADES NUCLEAR PLANT INTEGRATED INSPECTION
REPORT 05000255/2008-005
Dear Mr. Schwarz:
On December 31, 2008, the U. S. Nuclear Regulatory Commission (NRC) completed an
inspection at your Palisades Nuclear Plant. The enclosed report documents the inspection
findings, which were discussed on January 22, 2009, with 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.
Based on the results of this inspection, two NRC-identified violations of very low safety
significance were identified. The findings involved violations of NRC requirements. However,
because of their very low safety significance, and because the issues were entered into your
corrective action program, the NRC is treating the issues as Non-Cited Violations (NCVs)in
accordance with Section VI.A.1 of the NRC Enforcement Policy. Additionally, one licensee
identified violation is listed in Section 4OA7 of this report.
If you contest the subject or severity of an NCV, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the
U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington,
DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory
Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the
Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; and the Resident Inspector Office at the Palisades Nuclear Plant.
C. Schwarz -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
John B. Giessner, Chief
Branch 4
Division of Reactor Projects
Docket No. 50-255
License No. DPR-20
Enclosure: Inspection Report 05000255/2008-005
w/Attachment: Supplemental Information
cc w/encl: Senior Vice President
Vice President Oversight
Senior Manager, Nuclear Safety & Licensing
Senior Vice President and COO
Assistant General Counsel
Manager, Licensing
W. DiProfio
W. Russell
G. Randolph
Supervisor, Covert Township
Office of the Governor
T. Strong, State Liaison Officer
Michigan Department of Environmental Quality
Michigan Office of the Attorney General
C. Schwarz -2-
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its
enclosure will be available electronically for public inspection in the NRC Public Document
Room or from the Publicly Available Records (PARS) component of NRC's document system
(ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
John B. Giessner, Chief
Branch 4
Division of Reactor Projects
Docket No. 50-255
License No. DPR-20
Enclosure: Inspection Report 05000255/2008-005
w/Attachment: Supplemental Information
cc w/encl: Senior Vice President
Vice President Oversight
Senior Manager, Nuclear Safety & Licensing
Senior Vice President and COO
Assistant General Counsel
Manager, Licensing
W. DiProfio
W. Russell
G. Randolph
Supervisor, Covert Township
Office of the Governor
T. Strong, State Liaison Officer
Michigan Department of Environmental Quality
Michigan Office of the Attorney General
DOCUMENT NAME: G:\1-Secy\1-Work In Progress\Pali 2008-005.doc
Publicly Available Non-Publicly Available Sensitive Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII
NAME RLerch:dtp JGiessner
DATE 02/06/09 02/06/09
OFFICIAL RECORD COPY
Letter to C. Schwarz from J. Giessner dated February 9, 2009
SUBJECT: PALISADES NUCLEAR PLANT INTEGRATED INSPECTION
REPORT 05000255/2008-005
DISTRIBUTION:
RidsNrrPMPalisades
RidsNrrDorlLpl3-1
RidsNrrDirsIrib Resource
Mark Satorius
Kenneth Obrien
Cynthia Pederson
DRPIII
DRSIII
Patricia Buckley
ROPreports@nrc.gov
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No: 50-255
License No: DPR-20
Report No: 05000255/2008-005
Licensee: Entergy Nuclear Operations, Inc.
Facility: Palisades Nuclear Plant
Location: Covert, MI
Dates: October 1, 2008, to December 31, 2008
Inspectors: J. Ellegood, Senior Resident Inspector
T. Taylor, Resident Inspector
J. Cassidy, Senior Health Physicist
A. Dahbur, Senior Reactor Inspector
R. Jickling, Senior Emergency Preparedness Inspector
R. Winter, Reactor Inspector
Approved by: J. Giessner, Chief
Branch 4
Division of Reactor Projects
Enclosure
TABLE OF CONTENTS
SUMMARY OF FINDINGS ......................................................................................................... 1
REPORT DETAILS..................................................................................................................... 3
Summary of Plant Status......................................................................................................... 3
1. REACTOR SAFETY ..................................................................................................... 3
1R04 Equipment Alignment (71111.04) ....................................................................... 3
1R05 Fire Protection (71111.05) ................................................................................. 4
1R11 Licensed Operator Requalification Program (71111.11)..................................... 5
1R12 Maintenance Effectiveness (71111.12) .............................................................. 5
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13).......... 6
1R15 Operability Evaluations (71111.15) .................................................................... 6
1R18 Plant Modifications (71111.18) ........................................................................... 8
1R19 Post-Maintenance Testing (71111.19) ............................................................... 9
1R22 Surveillance Testing (71111.22)....................................................................... 10
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04) ............... 11
1EP6 Drill Evaluation (71114.06) ............................................................................... 11
2. RADIATION SAFETY ................................................................................................. 12
2OS1 Access Control to Radiologically Significant Areas (71121.01) ........................ 12
4. OTHER ACTIVITIES .................................................................................................. 16
4OA1 Performance Indicator Verification (71151) ...................................................... 16
4OA2 Identification and Resolution of Problems (71152) ........................................... 17
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153) .............. 22
4OA5 Other Activities................................................................................................. 24
4OA6 Management Meetings .................................................................................... 25
4OA7 Licensee-Identified Violations .......................................................................... 26
SUPPLEMENTAL iNFORMATION ............................................................................................. 1
KEY POINTS OF CONTACT .................................................................................................. 1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED ....................................................... 2
LIST OF DOCUMENTS REVIEWED ....................................................................................... 3
LIST OF ACRONYMS USED .................................................................................................. 8
Enclosure
SUMMARY OF FINDINGS
IR 05000255/2008-005; 10/01/2008 - 12/31/2008; Palisades Power Plant; Integrated Inspection
Report; Operability Evaluations; Follow-up of Events.
This report covers a 3-month period of inspection by resident inspectors and announced
baseline inspections by regional inspectors. Two Green findings were identified by the
inspectors. The findings were considered Non-Cited Violations of NRC regulations. 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 4,
dated December 2006.
A. NRC-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a finding of very low safety significance (Green) and
an associated NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for
the inadequate testing of the heat removal capacity of the Control Room Heating,
Ventilation, and Air Conditioning (CR HVAC) system. Specifically, the licensee isolated
refrigerant hot gas bypass flow during the test which increased the heat removal
capability of the chiller. The licensee entered the issue into their corrective action
program as condition report (CR) PLP-2008-3993 and re-performed portions of the
engineering basis calculation to demonstrate margin to account for the hot gas bypass
flow.
The finding is more than minor because, in accordance with IMC 0612, Appendix E,
Examples of Minor Issues, the inspectors determined that the finding was similar to
Example E.3.j and resulted in a reasonable doubt as to the operability of the chiller.
Based upon a review of the licensees revised calculation for the CR HVAC system
acceptance criteria and the Technical Specification (TS) requirements, the finding
screens as very low safety significance (Green) using the Phase 1 SDP worksheets.
The inspectors determined that the finding included a cross-cutting aspect in the area of
human performance, resources, and complete and accurate procedures (H.2(c))
because the surveillance procedure unacceptably preconditioned the chiller. (1R15)
- Green. A self-revealed finding of very low safety significance (Green) and an associated
NCV for failure to comply with TS 3.8.1 requirements when metal fragments were found
in the valve assembly area of the 1-2 Emergency Diesel Generator (EDG) cylinder 2L.
The source of the fragments was a failed spring lock for one of the exhaust valves. This
resulted in the EDG being inoperable for a period greater than allowed by TSs.
Subsequently, the licensee inspected the remaining spring locks on the 1-2 EDG.
Inspections of the 1-1 EDG spring locks are planned.
The finding is more than minor because it affected the equipment performance attribute
of the mitigating system cornerstone and adversely affected the objective of ensuring the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. A failure analysis performed by the vendor in
1 Enclosure
conjunction with an apparent cause analysis by the licensee led to an evaluation that the
diesel could perform its safety function for at least the 24-hour Probabilistic Risk
Assessment mission time. In consultation with the regional Senior Risk Analyst, the
finding screens as Green using the significance determination process phase 1. No
cross-cutting aspect was assigned because this issue is not indicative of current plant
performance. (4OA3)
B. Licensee-Identified Violations
A violation of very low safety significance that was identified by the licensee has been
reviewed by inspectors. Corrective actions planned or taken by the licensee have been
entered into the licensees corrective action program. This violation and corrective
action tracking numbers are listed in Section 4OA7 of this report.
2 Enclosure
REPORT DETAILS
Summary of Plant Status
Throughout the inspection period, the plant operated at or near 100 percent power.
1. REACTOR SAFETY
1R01 Adverse Weather Protection (71111.01)
a. Inspection Scope
Since extreme cold conditions were forecast in the vicinity of the facility for
December 2008, the inspectors reviewed the licensees overall preparations/protection
for the expected weather conditions. On December 17, the inspectors walked down the
condensate storage tanks and safety injection and refueling water tank system(s)
because their safety-related functions could be affected or required as a result of the
extreme cold conditions forecast for the facility. The inspectors observed insulation,
heat trace circuits, and weatherized enclosures to ensure operability of affected
systems. The inspectors reviewed licensee procedures. Specific documents reviewed
during this inspection are listed in the Attachment.
This inspection constituted one readiness for impending adverse weather condition
sample as defined in IP 71111.01-05.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment (71111.04)
a. Inspection Scope
The inspectors performed partial system walkdowns of the following risk-significant
systems:
- High Pressure Safety Injection (HPSI)-A with HPSI-B out of service for
maintenance;
- Service Water System with Service Water Pump C out of service for
maintenance;
- HPSI-B with HPSI-A out of service for maintenance;
The inspectors selected these systems based on their risk significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors attempted
to identify any discrepancies that could impact the function of the system, and, therefore,
potentially increase risk. The inspectors reviewed applicable operating procedures,
system diagrams, Updated Final Safety Analysis Report (UFSAR), Technical
Specification (TS) requirements, outstanding work orders, condition reports, and the
impact of ongoing work activities on redundant trains of equipment in order to identify
conditions that could have rendered the systems incapable of performing their intended
3 Enclosure
functions. The inspectors also walked down accessible portions of the systems to verify
system components and support equipment were aligned correctly and operable. The
inspectors examined the material condition of the components and observed operating
parameters of equipment to verify that there were no obvious deficiencies. The
inspectors also verified that the licensee had properly identified and resolved equipment
alignment problems that could cause initiating events or impact the capability of
mitigating systems or barriers and entered them into the Corrective Action program
(CAP) with the appropriate significance characterization. Documents reviewed are listed
in the Attachment.
These activities constituted four partial system walkdown samples as defined in
IP 71111.04-05.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05)
a. Inspection Scope
The inspectors conducted fire protection walkdowns which were focused on availability,
accessibility, and the condition of firefighting equipment in the following risk-significant
plant areas:
- 1-D Switchgear Room (during/following emergent Post Maintenance Testing for
Auxiliary Feed Water (AFW) -C power supply);
- Component Cooling Water Room;
- North and Southwest Penetration Rooms;
- Charging Pumps Room.
The inspectors reviewed areas to assess if the licensee had implemented a fire
protection program that adequately controlled combustibles and ignition sources within
the plant, effectively maintained fire detection and suppression capability, maintained
passive fire protection features in good material condition, and had implemented
adequate compensatory measures for out of service, degraded, or inoperable fire
protection equipment, systems, or features in accordance with the licensees fire plan.
The inspectors selected fire areas based on their overall contribution to internal fire risk
as documented in the plants Individual Plant Examination of External Events with later
additional insights, their potential to impact equipment which could initiate or mitigate a
plant transient, or their impact on the plants ability to respond to a security event. Using
the documents listed in the Attachment, the inspectors verified that fire hoses and
extinguishers were in their designated locations and available for immediate use; that
fire detectors and sprinklers were unobstructed; that transient material loading was
within the analyzed limits; and fire doors, dampers, and penetration seals appeared to
be in satisfactory condition. The inspectors also verified that minor issues identified
during the inspection were entered into the licensees CAP. Documents reviewed are
listed in the Attachment to this report.
4 Enclosure
These activities constituted five quarterly fire protection inspection samples as defined in
IP 71111.05-05.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program (71111.11)
a. Inspection Scope
On November 17, the inspectors observed a crew of licensed operators in the plants
simulator during licensed operator requalification examinations to verify that operator
performance was adequate, evaluators were identifying, and documenting crew
performance problems, and training was being conducted in accordance with licensee
procedures. The inspectors evaluated the following areas:
- licensed operator performance;
- crews clarity and formality of communications;
- ability to take timely actions in the conservative direction;
- prioritization, interpretation, and verification of annunciator alarms;
- correct use and implementation of abnormal and emergency procedures;
- control board manipulations;
- oversight and direction from supervisors; and
- ability to identify and implement appropriate TS actions and Emergency Plan
actions and notifications.
The crews performance in these areas was compared to pre-established operator action
expectations and successful critical task completion requirements. Documents reviewed
are listed in the Attachment to this report.
This inspection constituted one quarterly licensed operator requalification program
sample as defined in IP 71111.11.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12)
a. Inspection Scope
The inspectors evaluated degraded performance issues involving the following risk
significant systems:
The inspectors assessed performance issues with respect to the reliability, availability,
and condition monitoring of the system. In addition, the inspectors verified maintenance
effectiveness issues were entered into the corrective action program with the appropriate
significance characterization. Documents reviewed are listed in the Attachment.
5 Enclosure
This inspection constitutes one quarterly maintenance effectiveness samples as defined
in IP 71111.12-05.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13)
a. Inspection Scope
The inspectors reviewed the licensee's evaluation and management of plant risk for the
maintenance and emergent work activities affecting risk-significant and safety-related
equipment listed below to verify that the appropriate risk assessments were performed
prior to removing equipment for work:
- HPSI-B bearing oil change with hot leg injection valve breaker work and
switchyard work (yellow risk);
- Unplanned AFW right train inoperability.
These activities were selected based on their potential risk significance relative to the
reactor safety cornerstones. As applicable for each activity, the inspectors verified that
risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate
and complete. When emergent work was performed, the inspectors verified that the
plant risk was promptly reassessed and managed. The inspectors reviewed the scope
of maintenance work, discussed the results of the assessment with the licensee's
probabilistic risk analyst or shift technical advisor, and verified plant conditions were
consistent with the risk assessment. The inspectors also reviewed TS requirements and
walked down portions of redundant safety systems, when applicable, to verify risk
analysis assumptions were valid and applicable requirements were met.
These maintenance risk assessments and emergent work control activities constituted
two samples as defined in IP 71111.13-05.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15)
a. Inspection Scope
The inspectors reviewed the following issues:
- CR HVAC chiller capacity with hot-gas bypass isolated;
- CR HVAC fittings found not in-accordance with design;
- Containment Spray header inoperable after pressure gauge replacement;
- EDG 1-2 due to increased load from Containment Air Coolers.
The inspectors selected these potential operability issues based on the risk-significance
of the associated components and systems. The inspectors evaluated the technical
6 Enclosure
adequacy of the evaluations to ensure that TS operability was properly justified, and the
subject component or system remained available such that no unrecognized increase in
risk occurred. The inspectors compared the operability and design criteria in the
appropriate sections of the TS and UFSAR to the licensees evaluations, to determine
whether the components or systems were operable. Where compensatory measures
were required to maintain operability, the inspectors determined whether the measures
in place would function as intended and were properly controlled. The inspectors
determined, where appropriate, compliance with bounding limitations associated with the
evaluations. Additionally, the inspectors also reviewed a sampling of corrective action
documents to verify that the licensee was identifying and correcting any deficiencies
associated with operability evaluations. Documents reviewed are listed in the
Attachment.
This inspection constitutes four samples as defined in IP 71111.15.-05.
b. Findings
Introduction: The inspectors identified a finding of very low safety significance (Green)
and an associated NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, for
the inadequate testing of the heat removal capacity of the CR HVAC system.
Specifically, the licensee isolated refrigerant hot gas bypass flow during the test which
increased the heat removal capability of the chiller.
Description: On August 26, 2008, the licensee performed a surveillance test, RT-202, to
determine the heat removal capability of the A chiller. The test assessed the capability
of chiller to remove heat load during design basis conditions. The inspectors reviewed
the procedure and conduct of the test and noted that step 5.1.3.d of the procedure
manually isolated the hot gas bypass valve of the chiller. Under normal and design
conditions, the hot gas bypass flow control valve regulates automatically to open and
bypass refrigerant from the condensing unit to maintain compressor suction pressure
under low load condition. At full load conditions the valve is closed, but by design there
is a small amount of refrigerant that flows. Therefore, isolating the hot gas bypass
increases the heat removal capability of the chiller. Since hot gas bypass would not be
isolated during normal operation nor during an event, the inspectors concluded that
isolating hot gas bypass was preconditioning. As part of the corrective actions, the
licensee reviewed test data going back approximately 3 years and adjusted the data to
account for hot gas bypass flow. For the July 2007 test on the A train of the CR HVAC
system, taking hot gas bypass flow into consideration would have resulted in a
measured capacity approximately 9200 BTU/hr below the acceptance criteria of RT-202.
Based on discussions with the system engineer and a review of previous revisions of
RT-202, the inspectors determined that RT-202 was being performed with hot gas
bypass flow isolated since the year 2000.
In order to determine current operability of the CR HVAC system, the licensee adjusted
the most recent results of RT-202 to account for hot gas bypass flow. The licensee
determined that the B CR HVAC train was operable and that the A CR HVAC train
would remain operable below an outside air temperature of 87 degrees. Upon further
questioning by the inspectors regarding the TS requirements, the licensee later declared
the A train of CR HVAC inoperable. As part of the licensees corrective actions, the
RT-202 basis calculation was revised. By using margin included in the analysis
regarding allowable control room temperature, the licensee provided a basis to restore
7 Enclosure
the A chiller to operable status and demonstrate operability for the previous tests that
were reviewed.
Analysis: The inspectors concluded that the failure to properly demonstrate the
capability of the chillers warranted an evaluation using the SDP. Specifically, the
manual isolation of the hot gas bypass line provided an erroneous value for chiller
capacity. On one occasion, isolation of the flow path masked a condition where the
chiller would not have been able to meet surveillance requirements. Although the
licensee was able to revise an underlying calculation to show the chiller could maintain
control room temperatures within the TS requirements, the failure to meet the
requirements prior to the calculation markup created a reasonable doubt of the
operability of the chiller. Using IMC 0612, Appendix E, Examples of Minor Issues, the
inspectors determined that the finding was similar to Example E.3.j and was more than
minor because the error resulted in reasonable doubt as to the operability of the chiller.
This finding affects the Procedure Quality attribute of the Mitigating Systems cornerstone
objective of ensuring the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences in that the licensee used
inadequate testing to demonstrate operability of the system. To further assess the
significance of the finding, the inspectors used IMC 0609, Appendix A, Determining the
Significance of Reactor Inspection Findings for At-Power Situations. Based upon a
review of the licensees revised calculation for the CR HVAC system acceptance criteria
and the TS requirements, the finding screens as very low safety significance (Green)
using the Phase 1 worksheets due to answering no to each of the screening questions.
The inspectors determined that the finding included a cross-cutting aspect in the area of
human performance, resources, and complete and accurate procedures (H.2(c))
because the surveillance procedure unacceptably preconditioned the chiller. The finding
reflects current performance because the licensee revised the steps in question in
July 2007 but failed to identify and correct the preconditioning.
Enforcement: Appendix B of 10 CFR Part 50, Criterion XI, Test Control, requires, in
part, that a test program be established to assure that all testing required to
demonstrate that structures, systems, and components will perform satisfactorily in
service is identified and performed in accordance with written test procedures which
incorporate the requirements and acceptance limits contained in the applicable design
documents. Contrary to this requirement, during the performance of TS Surveillance
Procedure RT-202, Control Room HVAC Heat Removal Capability in August 2008, the
inspectors identified that the procedure does not adequately test the chillers to ensure
bypass flow of refrigerant is accounted for in the capacity of the unit. In addition,
potential degradation in the hot gas bypass flow line is not analyzed. Because the
finding is of very low safety significance and has been entered into the licensees
CAP as CR-2008-03993, this violation of 10 CFR 50, Appendix B, Criterion XI, is being
treated as a NCV, consistent with Section VI.A of the NRC Enforcement Policy.
(NCV 05000255/2008005 01, Inadequate Testing of Control Room Chillers)
1R18 Plant Modifications (71111.18)
a. Inspection Scope
The inspectors reviewed the following temporary modification(s):
- EC-10638, Disable Fast Transfer.
8 Enclosure
The inspectors compared the temporary configuration changes and associated
10 CFR 50.59 screening and evaluation information against the design basis, the
UFSAR, and the TS, as applicable, to verify that the modification did not affect the
operability or availability of the affected system(s). The inspectors also compared the
licensees information to operating experience information to ensure that lessons learned
from other utilities had been incorporated into the licensees decision to implement the
temporary modification. The inspectors, as applicable, performed field verifications to
ensure that the modifications were installed as directed; the modifications operated as
expected; modification testing adequately demonstrated continued system operability,
availability, and reliability; and that operation of the modifications did not impact the
operability of any interfacing systems. Lastly, the inspectors discussed the temporary
modification with operations, engineering, and training personnel to ensure that the
individuals were aware of how extended operation with the temporary modification in
place could impact overall plant performance.
This inspection constituted one temporary modification sample as defined in
IP 71111.18-05.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing (71111.19)
a. Inspection Scope
The inspectors reviewed the following post-maintenance activities to verify that
procedures and test activities were adequate to ensure system operability and functional
capability:
- Uninterruptible power supply following installation;
- Primary rod position indication following power supply replacement;
- AFW Right Train Power Supply Replacement.
These activities were selected based upon the structure, system, or component's ability
to impact risk. The inspectors evaluated these activities for the following (as applicable):
the effect of testing on the plant had been adequately addressed; testing was adequate
for the maintenance performed; acceptance criteria were clear and demonstrated
operational readiness; test instrumentation was appropriate; tests were performed as
written in accordance with properly reviewed and approved procedures; equipment was
returned to its operational status following testing (temporary modifications or jumpers
required for test performance were properly removed after test completion); and test
documentation was properly evaluated. The inspectors evaluated the activities against
TS, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various
NRC generic communications to ensure that the test results adequately ensured that the
equipment met the licensing basis and design requirements. In addition, the inspectors
reviewed corrective action documents associated with post-maintenance tests to
determine whether the licensee was identifying problems and entering them in the CAP
and that the problems were being corrected commensurate with their importance to
safety. Documents reviewed are listed in the Attachment to this report.
9 Enclosure
This inspection constituted three post-maintenance testing samples as defined in
IP 71111.19-05.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22)
a. Inspection Scope
The inspectors reviewed the test results for the following activities to determine whether
risk-significant systems and equipment were capable of performing their intended safety
function and to verify testing was conducted in accordance with applicable procedural
and TS requirements:
- QO-1, Safety Injection Test;
- RO-52, Fire Pump Capacity Test;
- QO-15, Component Cooling Water-B Pump In-Service Test.
The inspectors observed in-plant activities and reviewed procedures and associated
records to determine the following:
- did preconditioning occur;
- were the effects of the testing adequately addressed by control room personnel
or engineers prior to the commencement of the testing;
- were acceptance criteria clearly stated, demonstrated operational readiness, and
consistent with the system design basis;
- plant equipment calibration was correct, accurate, and properly documented;
- as-left setpoints were within required ranges; and the calibration frequency were
in accordance with TSs, the UFSAR, procedures, and applicable commitments;
- measuring and test equipment calibration was current;
- test equipment was used within the required range and accuracy; applicable
prerequisites described in the test procedures were satisfied;
- test frequencies met TS requirements to demonstrate operability and reliability;
tests were performed in accordance with the test procedures and other
applicable procedures; jumpers and lifted leads were controlled and restored
where used;
- test data and results were accurate, complete, within limits, and valid;
- test equipment was removed after testing;
- where applicable for inservice testing activities, testing was performed in
accordance with the applicable version of Section XI, American Society of
Mechanical Engineers Code, and reference values were consistent with the
system design basis;
- where applicable, test results not meeting acceptance criteria were addressed
with an adequate operability evaluation or the system or component was
declared inoperable;
- where applicable for safety-related instrument control surveillance tests,
reference setting data were accurately incorporated in the test procedure;
10 Enclosure
- where applicable, actual conditions encountering high resistance electrical
contacts were such that the intended safety function could still be accomplished;
- prior procedure changes had not provided an opportunity to identify problems
encountered during the performance of the surveillance or calibration test;
- equipment was returned to a position or status required to support the
performance of its safety functions; and
- all problems identified during the testing were appropriately documented and
dispositioned in the CAP.
Documents reviewed are listed in the Attachment to this report.
This inspection constituted four routine surveillance testing sample(s) and one inservice
testing sample.
b. Findings
No findings of significance were identified.
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)
a. Inspection Scope
Since the last NRC inspection of this program area, Emergency Plan Revision 17 and
Implementing Procedure EI-1, "Emergency Classification and Actions," Revision 49,
were implemented based on your determination, in accordance with 10 CFR 50.54(q),
that the changes resulted in no decrease in effectiveness of the Plan, and that the
revised Plan as changed continues to meet the requirements of 10 CFR 50.47(b) and
Appendix E to 10 CFR Part 50. The inspectors conducted a sampling review of the
Emergency Plan changes and a review of the Emergency Action Level changes to
evaluate for potential decreases in effectiveness of the Plan. However, this review does
not constitute formal NRC approval of the changes. Therefore, these changes remain
subject to future NRC inspection in their entirety.
This Emergency Action Level and Emergency Plan Changes inspection constituted one
sample as defined in IP 71114.04-05.
b. Findings
No findings of significance were identified.
1EP6 Drill Evaluation (71114.06)
a. Inspection Scope
The inspectors evaluated the conduct of a routine licensee emergency drill on
November 5, 2008, and a table top drill for the emergency operations facility on
December 17, 2008, to identify any weaknesses and deficiencies in classification,
notification, and protective action recommendation development activities. The
inspectors observed, as applicable, emergency response operations in the simulator
control room, technical support center, and emergency operations facility to determine
whether the event classification, notifications, and protective action recommendations
11 Enclosure
were performed in accordance with procedures. The inspectors also attended the
licensee drill critique to compare any inspector-observed weakness with those identified
by the licensee staff in order to evaluate the critique and to verify whether the licensee
staff was properly identifying weaknesses and entering them into the corrective action
program. As part of the inspection, the inspectors reviewed the drill package and other
documents listed in the Attachment to this report.
These emergency preparedness drill inspections constituted two samples as defined in
IP 71114.06-05.
b. Findings
No findings of significance were identified.
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01)
.1 Plant Walkdowns and Radiation Work Permit (RWP) Reviews
a. Inspection Scope
The inspectors reviewed licensee controls and surveys in the following radiologically
significant work areas within radiation areas, high radiation areas, and airborne
radioactivity areas in the plant to determine if radiological controls including surveys,
postings, and barricades were acceptable:
- Auxiliary Building;
- Containment Building; and
- Spent Fuel Pool Area.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors reviewed the RWPs and work packages used to access these areas and
other high radiation work areas. The inspectors assessed the work control instructions
and control barriers specified by the licensee. Electronic dosimeter alarm set points for
both integrated dose and dose rate were evaluated for conformity with survey indications
and plant policy. The inspectors interviewed workers to verify that they were aware of
the actions required if their electronic dosimeters noticeably malfunctioned or alarmed.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors walked down and surveyed (using an NRC survey meter) these areas to
verify that the prescribed RWP, procedure, and engineering controls were in place; that
licensee surveys and postings were complete and accurate; and that air samplers were
properly located.
This inspection constitutes one sample as defined in IP 71121.01-5.
12 Enclosure
The inspectors also reviewed the licensees physical and programmatic controls for
highly activated and/or contaminated materials (non-fuel) stored within the spent fuel
pool or other storage pools.
This inspection constitutes one sample as defined in IP 71121.01-5.
b. Findings
No findings of significance were identified.
.2 Problem Identification and Resolution
a. Inspection Scope
The inspectors reviewed a sample of the licensees self-assessments, audits, Licensee
Event Reports (LERs), and Special Reports related to the access control program to
verify that identified problems were entered into the CAP for resolution.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors evaluated the licensees process for problem identification,
characterization, and prioritization and verified that problems were entered into the
CAP and resolved. For repetitive deficiencies and/or significant individual deficiencies
in problem identification and resolution, the inspectors verified that the licensees
self-assessment activities were capable of identifying and addressing these deficiencies.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors reviewed licensee documentation packages for all performance indicator
(PI) events occurring since the last inspection to determine if any of these PI events
involved dose rates in excess of 25 R/hr at 30 centimeters or in excess of 500 R/hr at
1 meter. Barriers were evaluated for failure and to determine if there were any barriers
left to prevent personnel access. Unintended exposures exceeding 100 millirem total
effective dose equivalent (or 5 rem shallow dose equivalent or 1.5 rem lens dose
equivalent) were evaluated to determine if there were any regulatory overexposures or if
there was a substantial potential for an overexposure.
This inspection constitutes one sample as defined in IP 71121.01-5.
b. Findings
No findings of significance were identified.
.3 Job-In-Progress Reviews
a. Inspection Scope
The inspectors observed the following two jobs that were being performed in radiation
areas, airborne radioactivity areas, or high radiation areas for observation of work
activities that presented the greatest radiological risk to workers:
13 Enclosure
- Troubleshooting Pressurizer Heater Breaker; and
- Negative Reactivity Testing of the Spent Fuel Pool Racks.
The inspectors reviewed radiological job requirements for these activities, including
RWP requirements and work procedure requirements.
This inspection constitutes one sample as defined in IP 71121.01-5.
Job performance was observed with respect to the radiological control requirements to
assess whether radiological conditions in the work area were adequately communicated
to workers through pre-job briefings and postings. The inspectors evaluated the
adequacy of radiological controls, including required radiation, contamination, and
airborne surveys for system breaches; radiation protection job coverage, including any
applicable audio and visual surveillance for remote job coverage; and contamination
controls.
This inspection constitutes one sample as defined in IP 71121.01-5.
a. Findings
No findings of significance were identified.
.4 High Risk Significant, High Dose Rate, High Radiation Area, and Very High Radiation
Area Controls
a. Inspection Scope
The inspectors held discussions with the Radiation Protection Manager concerning high
dose rate, high radiation area, and very high radiation area controls and procedures,
including procedural changes that had occurred since the last inspection, in order to
assess whether any procedure modifications substantially reduced the effectiveness and
level of worker protection.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors discussed with radiation protection supervisors the controls that were in
place for special areas of the plant that had the potential to become very high radiation
areas during certain plant operations. The inspectors assessed if plant operations
required communication beforehand with the radiation protection group, so as to allow
corresponding timely actions to properly post and control the radiation hazards.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors conducted plant walkdowns to assess the posting and locking of
entrances to high dose rate, high radiation areas, and very high radiation areas.
This inspection constitutes one sample as defined in IP 71121.01-5.
b. Findings
No findings of significance were identified
14 Enclosure
.5 Radiation Worker Performance
a. Inspection Scope
During job performance observations, the inspectors evaluated radiation worker
performance with respect to stated radiation safety work requirements. The inspectors
evaluated whether workers were aware of any significant radiological conditions in their
workplace, of the RWP controls and limits in place, and of the level of radiological
hazards present. The inspectors also observed worker performance to determine if
workers accounted for these radiological hazards.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors reviewed radiological problem reports for which the cause of the event
was due to radiation worker errors to determine if there was an observable pattern
traceable to a similar cause and to determine if this perspective matched the corrective
action approach taken by the licensee to resolve the reported problems. Problems or
issues with planned or completed corrective actions were discussed with the Radiation
Protection Manager.
This inspection constitutes one sample as defined in IP 71121.01-5.
b. Findings
No findings of significance were identified.
.6 Radiation Protection Technician Proficiency
a. Inspection Scope
During job performance observations, the inspectors evaluated radiation protection
technician performance with respect to radiation safety work requirements. The
inspectors evaluated whether technicians were aware of the radiological conditions in
their workplace, the RWP controls and limits in place, and if their performance was
consistent with their training and qualifications with respect to the radiological hazards
and work activities.
This inspection constitutes one sample as defined in IP 71121.01-5.
The inspectors reviewed radiological problem reports for which the cause of the event
was radiation protection technician error to determine if there was an observable pattern
traceable to a similar cause and to determine if this perspective matched the corrective
action approach taken by the licensee to resolve the reported problems.
This inspection constitutes one sample as defined in IP 71121.01-5.
b. Findings
No findings of significance were identified.
15 Enclosure
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Reactor Coolant System Leakage
a. Inspection Scope
The inspectors sampled licensee submittals for the Reactor Coolant System
Leakage performance indicator for the period from the fourth quarter 2007 through
the third quarter 2008. To determine the accuracy of the PI data reported during those
periods, PI definitions and guidance contained in the Nuclear Energy Institute (NEI)
Document 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 5, was used. The inspectors reviewed the licensees operator logs and verified
the accuracy of a sample of calculations for the period of October 2007 through
September 2008 to validate the licensees submittals. Documents reviewed are listed in
the Attachment to this report.
This inspection constituted one reactor coolant system leakage sample as defined in
IP 71151-05.
b. Findings
No findings of significance were identified.
.2 Unplanned Transients per 7000 Critical Hours
a. Inspection Scope
The inspectors sampled licensee submittals for the Unplanned Transients per
7000 Critical Hours performance indicator for the period from the fourth quarter 2007
through the third quarter of 2008. To determine the accuracy of the PI data reported
during those periods, PI definitions and guidance contained in the NEI Document 99-02,
Regulatory Assessment Performance Indicator Guideline, Revision 5, was used. The
inspectors reviewed the licensees operator narrative logs and NRC Integrated
Inspection Reports for the period of October 2007 through September 2008 to validate
the accuracy of the submittals. Documents reviewed are listed in the Attachment to this
report.
This inspection constituted one unplanned transients per 7000 critical hours sample as
defined in IP 71151-05.
b. Findings
No findings of significance were identified.
.3 Safety System Functional Failures
a. Inspection Scope
The inspectors sampled licensee submittals for the Safety System Functional Failures
performance indicator for the period from the fourth quarter 2007 through the third
16 Enclosure
quarter 2008 to determine the accuracy of the PI data reported during those periods,
PI definitions and guidance contained in the NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 5, and NUREG-1022, Event
Reporting Guidelines 10 CFR 50.72 and 50.73," definitions and guidance, were used.
The inspectors reviewed licensee event reports for the period of October 2007 through
September 2008 to validate the accuracy of the submittals. Documents reviewed are
listed in the Attachment to this report.
This inspection constituted one safety system functional failures sample as defined in
IP 71151-05.
.4 Mitigating Systems Performance Index - Cooling Water Systems
a. Inspection Scope
The inspectors sampled licensee submittals for the Mitigating Systems Performance
Index (MSPI) - Cooling Water Systems performance indicator for the period from the
fourth quarter 2007 through the third quarter of 2008 to determine the accuracy of the
PI data reported during those periods, PI definitions and guidance contained in the
NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 5, were used. The inspectors reviewed the licensees operator narrative logs,
condition reports, MSPI derivation reports, event reports, and NRC Integrated Inspection
Reports for the period of October 2007 through September 2008 to validate the accuracy
of the submittals. The inspectors reviewed the MSPI component risk coefficient to
determine if it had changed by more than 25 percent in value since the previous
inspection, and if so, that the change was in accordance with applicable NEI guidance.
The inspectors also reviewed the licensees issue report database to determine if any
problems had been identified with the PI data collected or transmitted for this indicator
and none were identified. Documents reviewed are listed in the Attachment to this
report.
This inspection constituted one MSPI cooling water system sample as defined in
IP 71151-05.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems (71152)
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency
Preparedness, Public Radiation Safety, Occupational Radiation Safety, and
.1 Routine Review of items Entered Into the Corrective Action Program
a. Inspection Scope
As part of the various baseline inspection procedures discussed in previous sections of
this report, the inspectors routinely reviewed issues during baseline inspection activities
and plant status reviews to verify that they were being entered into the licensees CAP at
17 Enclosure
an appropriate threshold, that adequate attention was being given to timely corrective
actions, and that adverse trends were identified and addressed. Attributes reviewed
included: the complete and accurate identification of the problem; that timeliness was
commensurate with the safety significance; that evaluation and disposition of
performance issues, generic implications, common causes, contributing factors, root
causes, extent of condition reviews, and previous occurrences reviews were proper and
adequate; and that the classification, prioritization, focus, and timeliness of corrective
actions were commensurate with safety and sufficient to prevent recurrence of the issue.
Minor issues entered into the licensees CAP as a result of the inspectors observations
are included in the attached List of Documents Reviewed.
These routine reviews for the identification and resolution of problems did not constitute
any additional inspection samples. Instead, by procedure they were considered an
integral part of the inspections performed during the quarter and documented in
Section 1 of this report.
b. Findings
No findings of significance were identified.
.2 Daily Corrective Action Program Reviews
a. Inspection Scope
In order to assist with the identification of repetitive equipment failures and specific
human performance issues for follow-up, the inspectors performed a daily screening of
items entered into the licensees CAP. This review was accomplished through
inspection of the stations daily condition report packages.
These daily reviews were performed by procedure as part of the inspectors daily plant
status monitoring activities and, as such, did not constitute any separate inspection
samples.
b. Findings
No findings of significance were identified.
.3 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a review of the licensees CAP and associated documents to
identify trends that could indicate the existence of a more significant safety issue. The
inspectors review was focused on repetitive equipment issues, but also considered the
results of daily inspector CAP item screening discussed in Section 4OA2.2 above,
licensee trending efforts, departmental performance reports and metrics, and licensee
human performance results. The inspectors review nominally considered the 6 month
period of July 2008 through December 2008, although some examples expanded
beyond those dates where the scope of the trend warranted.
18 Enclosure
The inspectors compared and contrasted their results with the results contained in the
licensees assessments. Corrective actions associated with a sample of the issues
identified in the licensees trending reports were reviewed for adequacy.
This review constituted a single semi-annual trend inspection sample as defined in
IP 71152-05.
b. Findings
No findings of significance were identified.
.4 Annual Sample: Review of Operator Workarounds
a. Inspection Scope
The inspectors evaluated the licensees implementation of their process used to identify,
document, track, and resolve operational challenges. Inspection activities included, but
were not limited to, a review of the cumulative effects of the Operator Work Arounds on
system availability and the potential for improper operation of the system, for potential
impacts on multiple systems, and on the ability of operators to respond to plant
transients or accidents.
The inspectors performed a review of the cumulative effects of Operator Work Arounds.
The documents listed in the Attachment were reviewed to accomplish the objectives of
the inspection procedure. The inspectors reviewed both current and historical
operational challenge records to determine whether the licensee was identifying operator
challenges at an appropriate threshold, had entered them into their CAP, and proposed
or implemented appropriate and timely corrective actions which addressed each issue.
Reviews were conducted to determine if any operator challenge could increase the
possibility of an Initiating Event, if the challenge was contrary to training, required a
change from long-standing operational practices, or created the potential for
inappropriate compensatory actions. Additionally, all temporary modifications were
reviewed to identify any potential effect on the functionality of Mitigating Systems,
impaired access to equipment, or required equipment uses for which the equipment was
not designed. Daily plant and equipment status logs, degraded instrument logs, and
operator aids or tools being used to compensate for material deficiencies were also
assessed to identify any potential sources of unidentified operator workarounds.
This review constituted one operator workaround annual inspection sample as defined in
IP 71152-05.
b. Findings
No findings of significance were identified.
.5 Selected Issue Follow-Up Inspection: Criticality Controls in the Spent Fuel Pool
a. Inspection Scope
In July of 2008, the licensee tested some spent fuel storage rack locations to determine
if the neutron poison built into the storage rack continued to meet assumptions in the
19 Enclosure
criticality analysis. The testing revealed that the neutron absorption capability of the
Spent Fuel Pool racks had degraded and in some cases no longer met assumptions
contained within the criticality analysis. The licensee implemented additional criticality
controls in the spent fuel pool and informed the NRC of the controls by letter dated
August 27, 2008. After reviewing the controls, the NRC issued Confirmatory Action
Letter (CAL) RIII-08-003 to confirm the commitments made by Entergy Nuclear
Operations.
During this inspection period, the inspectors validated that the licensee implemented the
requirements of the CAL. Specific actions included:
- review of the licensees basis for criticality safety for proposed fuel moves;
- verification that fuel moves complied with CAL requirements;
- observation of additional testing of the spent fuel pool neutron absorption
capability.
The inspectors concluded that the licensee complied with the requirements of the CAL.
This review constituted one in-depth problem identification and resolution sample as
defined in IP 71152-05.
b. Findings
No findings of significance were identified.
.6 Selected Issue Follow-Up Inspection: Deviation between Estimated and Actual Critical
Positions for the August 9, 2008, Startup
a. Inspection Scope
The inspectors reviewed the data and associated procedures for the August 9, 2008,
reactor startup from a forced outage. The inspectors also interviewed licensee staff and
searched for relevant operating experience and past similar issues identified by the NRC
at other plants. Corrective action documents related to the issue were reviewed for
appropriate categorization and action. During the August 9, 2008, reactor startup, the
reactor attained criticality with rods withdrawn to a position of 0.28 percent delta rho
above that predicted by the estimated critical position. Although the deviation was below
the TS limit of 1 percent delta rho and the licensee administrative limit of 0.5 percent
delta rho, the deviation was abnormally high. The licensee reviewed the error in
condition report CR-PLP-2008-3439. The cause evaluation concluded the error
occurred due to boron 10 depletion in the primary coolant system. B-10 depletion
occurs due to B-10 burnup via neutron absorption. Naturally occurring boron contains
19.8 atom percent B-10 with the remainder B-11. As B-10 (which has a much larger
neutron absorption cross-section) absorbs neutrons, the ratio of B-10 to B-11 decreases.
This change alters the effective boron poison strength of the primary coolant for a given
boron concentration. When fresh boron is added, the effective boron concentration
changes. In this instance, the licensee added fresh boron during the shutdown but did
not account for it while determining an estimated critical position. As part of the
corrective actions, the licensee identified several items to help address this issue. The
licensee did have some guidance in EM-04-24, their Critical Prediction and Critical
Approach procedure, in regards to when to apply a B-10 correction. The licensee plans
20 Enclosure
to enhance this portion of the procedure for clarity and ease of conducting a B-10
correction, if deemed necessary. In addition, the licensee benchmarked other fleet
plants to determine an optimum isotopic sampling frequency to track the depletion of
B-10. The licensee is proceeding with plans to obtain and ship samples for analysis
more frequently.
Based on data provided by the licensee, the last time there was a deviation of this
magnitude or greater was June 1998 with a 0.393 percent delta rho error. Beyond that,
in October 1997 there was a 0.437 percent deviation. The October 1997 deviation was
determined to be mostly due to B-10 depletion. A search for relevant operating
experience and previous findings was conducted with no relevant results. The
inspectors reviewed the licensees reactivity management event classification guidelines
and corrective action process procedure to validate appropriate categorization of the
issue. No deficiencies were identified.
This review constituted one in-depth problem identification and resolution sample as
defined in IP 71152-05.
.7 Selected Issue Follow-Up Inspection: Licensee Compliance with the Station Blackout
(SBO) Rule Assumptions
a. Inspection Scope
While performing an operability evaluation on inadequate testing of the CR HVAC
system (discussed in Section 1R15 of this report), the inspectors noted that the
licensees revision to the underlying calculation for the systems surveillance test may
not have taken into account SBO Rule (10 CFR 50.63) requirements. The inspectors
interviewed licensee personnel and reviewed design basis documents, the UFSAR, and
docketed correspondence between the NRC and the licensee. The inspectors review
included the analysis performed by the licensee to demonstrate compliance with the
SBO rule. Based on the review, the inspectors concluded the licensee remained
compliant with the license condition for SBO but noted weaknesses in the licensees
program. Noted weaknesses included:
- Lack of operations staff awareness of impacts on SBO rule compliance with
elevated temperatures in the control room
- Lack of programmatic controls for maintaining compliance with the plants SBO
assumptions
The licensee generated two condition reports (CR-PLP-2008-5023 and
CR-PLP-2008-5074) to address the issue.
This review constituted one in-depth problem identification and resolution sample as
defined in IP 71152-05.
b. Findings
No findings of significance were identified.
21 Enclosure
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153)
.1 (Closed) LER 05000255/2008-05-00, Completion of Plant Shutdown Required by TSs
On August 5, 2008, the licensee began a planned shutdown to replace leaking Control
Rod Drive Mechanism seals. At 97 percent power, a relief valve, RV-2006, in the
letdown system lifted resulting in primary coolant leakage in excess of the TS limit of
1 gallon per minute of unidentified leakage. The licensee made a 4-hour report to the
NRC and completed a plant shutdown as required by TSs. The licensee determined that
the relief valve lifted when a second charging pump started per the pressurizer level
control program. Historically, the licensee has experienced difficulties with control of
pressure in the letdown system during operation of charging pumps due to performance
of controllers in the system. In addition, the licensee determined that the setpoint of the
relief valve will drift as a result of previous relief valve lifts. During the outage, the
licensee replaced the relief valve. In this event, leak rates of approximately 4 gallons per
minute occurred when the relief valve lifted. Because the line discharged into a quench
tank with multiple sources, the licensee treated the leakage as unidentified leakage. The
licensee exited the Limiting Condition for Operation (LCO) when the plant entered
Mode 5 and the LCO no longer applied. Documents reviewed as part of this inspection
are listed in the Attachment. The inspectors did not identify any additional safety issues.
This LER is closed.
This event follow-up review constituted one sample as defined in IP 71153-05.
.2 (Closed) LER 05000255/2008-06-00, Emergency Diesel Generator Inoperable in Excess
of TSs Requirements
a. Inspection Scope
On February 19, 2008, during performance of planned maintenance on the 1-2 EDG,
licensee personnel discovered foreign material (metal fragments) in the valve assembly
area of the 1-2 EDG cylinder 2L. The metal fragments were identified to be broken
pieces of the valve seat spring lock associated with the cylinder 2L inboard exhaust
valve. These fragments were discovered by workers during an unrelated maintenance
activity involving the snubber valves. Due to some physical interference between
components during work, the valve cover for cylinder 2L needed to be removed to
complete some of the snubber valve maintenance. When the cover was removed,
workers noted the spring lock material described above. The licensee performed an
apparent cause evaluation that included a failure analysis performed by Fairbanks
Morse Engine. The licensee determined that the spring lock was damaged due to a
condition discovered in March 2000. While performing maintenance on March 21, 2000,
licensee personnel discovered that the valve yoke retaining lock nut had fallen off one of
the valve assemblies for cylinder 2L. The licensee determined that because of this, the
adjusting screw backed out such that the valve yoke was able to strike the spring seat
instead of the adjusting nut striking the valve stem, as designed. This resulted in cyclical
side loading of the valve assembly which initiated a fatigue crack on the inside diameter
of the spring lock. Eventually it failed, resulting in the pieces discovered on
February 19, 2008. The licensee inspected the other spring locks on the 1-2 EDG and
found no discrepancies. An extent-of-condition review was performed for the 1-1 EDG
by the licensee. The licensee concluded that there was high assurance that a similar
22 Enclosure
condition does not exist on the 1-1 EDG. The licensee plans to inspect valve seat spring
locks for the 1-1 EDG during the next scheduled preventative maintenance outage.
The licensee reviewed the corrective actions for the March 2000 discovery of the lock
nut that fell off and the associated backing-out of the adjusting screw. To resolve the
issue, the nut was returned to the proper position and torqued. In addition, the other
retaining nuts were inspected and found to be acceptable. There were no documented
inspections of any other components in the valve assembly area under the valve cover.
Following the discovery, the licensee did not change the routine preventative
maintenance requirements. The inspectors identified one finding. This LER is closed.
This event follow-up review constituted one sample as defined in IP 71153-05.
b. Findings
Introduction: On February 19, 2008, a Green self-revealed finding and associated NCV
of TS requirement 3.8.1.b became evident when licensee personnel discovered metal
fragments in the valve assembly area of the 1-2 EDG cylinder 2L. Subsequent
evaluation by the licensee determined that the diesel would not be able to function for its
30-day mission time and was inoperable for a period of time greater than allowed by TS.
Description: While performing maintenance on the 1-2 EDG, licensee personnel noted
metal fragments in the valve cover for the 1-2 EDG. The licensee investigated the
source of the metal fragments and discovered that the spring lock for an exhaust valve
had failed. The licensee reviewed maintenance history for the diesel and determined
that a lock nut had backed off an adjustment screw in 2000. The licensee contracted the
diesel vendor to evaluate the condition of the cylinder head in order to determine the
cause of the spring lock failure. The vendor concluded that the loss of the lock nut
caused the adjusting screw to drift and allowed the yoke arm to contact the valve spring
seat. The abnormal operating condition likely created cracks in the spring lock that
propagated even after replacement of the lock nut. In 2000, the licensee limited
corrective actions to replacing the lock nut and checking the lock nuts on other cylinders.
The licensee did not evaluate potential degradation of other components due to the
abnormal operation of the valve.
The licensee evaluated the as-found condition and concluded that a sufficient amount of
the spring lock remained on the valve to assure continued operation of the EDG for at
least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. However, the wear rates could not be determined to assure the EDG
would operate for 30 days. The licensee identified that a potential failure mode was loss
of the valve into the associated cylinder and subsequent damage to the turbo charger
through transport of generated debris. As corrective action, the licensee replaced the
spring lock and inspected the others on the 1-2 EDG. An extent of condition was
performed for the 1-1 EDG, and inspections are planned for the 1-1 EDG as well. The
condition was entered into the licensees CAP as CR-PLP-2008-0822.
Analysis: The inspectors determined that the inoperability of the diesel generator for
longer than TS allowed outage time represented a performance deficiency that
warranted a significance determination. Through review of the condition reports related
to the missing lock nut in 2000, the inspectors determined that the licensee could
reasonably be expected to foresee and correct the condition of the cylinder/valve
operating mechanism. In 2000, the licensee did not evaluate potential adverse effects
23 Enclosure
on the valve and related components. Corrective actions at the time also indicated the
remaining lock nuts were only checked hand-tight. The inspectors determined the issue
is more than minor because it affected the mitigating system cornerstone objective of
ensuring the reliability and capability of mitigating systems. The inspectors performed a
phase 1 screen on the issue in accordance with IMC 0609 and determined that since the
EDG was inoperable for greater than the TS allowed outage time, support from the
region-based senior risk analyst was needed to determine risk significance. Based on
vendor analysis, the licensee concluded that the EDG would likely operate in excess of
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but not for the 30-day mission time required by the license. Since the EDG
would function for the Probable Risk Assessment mission time, the senior risk analyst
determined the issue would screen as Green using the phase 1 worksheets. No cross-
cutting aspect was assigned because the review dated back to 2000, therefore, this
issue is not indicative of current plant performance.
Enforcement: Technical Specification LCO 3.8.1 states, in part, that two diesel
generators each capable of supplying one train of the Class 1E AC electrical power
distribution system shall be operable. The TS action requirement requires a plant
shutdown if the associated EDG is inoperable for greater than 7 days. Contrary to the
required action statements, on February 19, 2008, a failed spring lock for one of the
exhaust valves on the 1-2 EDG was discovered. This resulted in an EDG inoperability
beyond the allowed required action times. The failure was determined to have been the
result of an issue discovered in 2000 involving a lock nut that had fallen off of the
exhaust valve assembly. Further analysis led to a conclusion by the licensee that the
condition would likely allow the EDG to run for the 24-hour Probabilistic Risk
Assessment mission time but not the 30-day mission time. Because the finding is of
very low safety significance and has been entered into the licensees CAP as
CR-2008-0822, this violation of TS requirement 3.8.1.b is being treated as an NCV,
consistent with Section VI.A of the NRC Enforcement Policy.
(NCV 05000255/2008005-02 Emergency Diesel Generator Inoperable in Excess of
TS Requirements).
4OA5 Other Activities
.1 Observations of Security
a. Inspection Scope
During the inspection period, the inspectors conducted observations of security force
personnel and activities to ensure that the activities were consistent with licensee
security procedures and regulatory requirements relating to nuclear plant security.
These observations took place during both normal and off-normal plant working hours.
These quarterly resident inspector observations of security force personnel and activities
did not constitute any additional inspection samples. Rather, they were considered an
integral part of the inspectors' normal plant status review and inspection activities.
b. Findings
No findings of significance were identified.
24 Enclosure
.2 Implementation of Temporary Instruction (TI) 2515/176, Emergency Diesel generator
Technical Specification Surveillance Requirements Regarding Endurance and Margin
testing
a. Inspection Scope
The objective of TI 2515/176 was to gather information to assess the adequacy of
nuclear power plant emergency diesel generator endurance and margin testing as
prescribed in plant-specific TSs. The inspectors reviewed the licensee's TS,
procedures, and calculations, and interviewed licensee personnel to complete the TI.
The information gathered for this TI was forwarded to the Office of Nuclear Reactor
Regulation for further review and evaluation on December 17, 2008. This TI is
complete at Palisades Nuclear Plant; however, this TI 2515/176 will not expire until
August 31, 2009. Additional information may be required after review by the Office of
Nuclear Reactor Regulation.
b. Findings
No findings of significance were identified.
4OA6 Management Meetings
.1 Exit Meeting Summary
On January 22, 2009, the inspectors presented the inspection results to Thomas Kirwin
and other members of the licensee staff. The licensee acknowledged the issues
presented. The inspectors confirmed that none of the potential report input discussed
was considered proprietary. A follow-up phone exit meeting was held by the NRC with
Ms. B. Dotson and Mr. C. Sherman on February 2, 2009 to discuss additional inspection
results.
.2 Interim Exit Meetings
Interim exits were conducted for:
- The annual review of emergency action level and emergency plan changes with
the licensee's Senior Emergency Planning Coordinator, Mr. M. Sweet via
telephone on December 30, 2008.
- The results of the access control to radiologically significant areas inspection with
the Plant Manager, Mr. T. Kirwin, and other members of your staff, on
December 12, 2008.
- A telephone exit for TI 2515/176 was conducted with John Broschak,
Entergy/Engineering Director and other Licensee staff on December 2, 2008.
The inspectors confirmed that none of the potential report input discussed was
considered proprietary.
25 Enclosure
4OA7 Licensee-Identified Violations
The following violation of very low significance (Green) were identified by the licensee
and are violations of NRC requirements which meet the criteria of Section VI of the
NRC Enforcement Policy, NUREG-1600, for being dispositioned as an NCV.
- Technical Specification 5.7.1 requires areas with dose rates greater than
100 mR/hr to be posted as a High Radiation Area and barricaded. Contrary to this,
on October 18, 2008, the high radiation area posting and barricade did not provide
an adequate barricade around areas of the room having dose rates greater than
100 mR/hr in the west engineered safeguards room. This was identified in the
licensees corrective action program as CR-PLP-2008-04310. The finding was
determined to be of very low safety significance because it was not an ALARA
planning issue, there was no overexposure nor substantial potential for
overexposure, and the licensees ability to assess dose was not compromised.
ATTACHMENT: SUPPLEMENTAL INFORMATION
26 Enclosure
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
C. Schwarz, Site Vice President
V. Beilfuss, Project Manager
J. Broschak, Engineering Director
N. Brott, Emergency Preparedness Coordinator
J. Burnett, RETS/REMP Specialist
T. Davis, Regulatory Compliance
B. Dotson, Regulatory Compliance
J. Fontaine, Senior Emergency Planning Coordinator
J. Ford, Corrective Action Manager
T. Kirwin, Plant General Manager
L. Lahti, Licensing Manager
B. Nixon, Assistant Operations Manager
T. Shewmaker, Chemistry Manager
C. Sherman, Radiation Protection Manager
M. Sicard, Operations Manager
G. Sleeper, Assistant Operations Manager
E. Williams, Radiation Protection Supervisor
Nuclear Regulatory Commission
J. Giessner, Chief, Branch 4
1 Attachment
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000255/2008005-01 NCV Inadequate Testing of Control Room Chillers (Section 1R15)05000255/2008005-02 NCV Emergency Diesel Generator Inoperable in Excess of
Technical Specification Requirements (Section 4OA3)
Closed
05000255/2008005-01 NCV Inadequate Testing of Control Room Chillers (Section 1R15)05000255/2008005-02 NCV Emergency Diesel Generator Inoperable in Excess of
Technical Specification Requirements (Section 4OA3)
05000255/2008005 LER Completion of Plant Shutdown Required by Technical
Specifications (Section 4OA3)
05000255/2008006 LER Emergency Diesel Generator Inoperable in Excess of
Technical Specification Requirements (Section 4OA3)
2 Attachment
LIST OF DOCUMENTS REVIEWED
The following is a partial list of documents reviewed during the inspection. Inclusion on this list
does not imply that the NRC inspector reviewed the documents in their entirety, but rather that
selected sections or portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
1R01 Adverse Weather Protection
- CR-PLP-2007-5672, 1-3 DG cold weather vent covers fall out of the vents, November 7, 2007
- CR-PLP-2007-5934, The Insulating Box for SIRW Tank Level Instrument LT-0331 is Bent,
November 21, 2007
- SOP-23, Plant Heating System, Revision 27
1R04 Equipment Alignment
- Drawing M-204, Safety Injection, Containment Spray, and Shutdown Cooling System
- Drawing M-208, Service Water System
- Drawing M-213, Service Water, Screen Structure, and Chlorinator
- FSAR Chapter 9.1.1, Service Water System, Revision 25
- Procedure No. 4.02, Control of Equipment, Revision 47
- SOP-15, Service Water System, Revision 47
- SOP-22, Emergency Diesel Generators, Revision 45
- SOP-3, High Pressure Safety Injection System, Revision 75
1R05 Fire Protection
- Fire Hazard Analysis, Palisades Plant, Revision 7
1R11 Licensed Operator Requalification Program
- Procedure SEP, Site Emergency Plan, Revision 16
- Simulator Exercise Guide for November 17, 2008 Licensed Operator Requalification Drill
1R12 Maintenance Effectiveness
- CR-PLP-2007-01807, K-6B Placed in Maintenance Rule Category a(1), May 3, 2007
- CR-PLP-2007-01145, K-6B exceeds Maintenance Rule Unavailability Hours, March 13, 2007
- CR-PLP-2008-01424, EG-30A, 1-2 Emergency Diesel Generator Alarm Panel Failure (K-6B),
March 28, 2008
- CR-PLP-2008-03141, During the Performance of MO-7A-2, Emergency Diesel Generator 1-2,
Difficulties were Encountered in Attempting to Match Generator Output Voltage to Match Bus
1D Voltage, July 21, 2008
- CR-PLP-2008-03142, While Increasing Load on 1-2 Diesel Generator, the Cyclinder Petcock
for 9R Came Open, July 21, 2008
- CR-PLP-2008-0822, Broken Valve Keeper (Spring Seat Lock) Found on Cylinder 2-L of the 1-
2 Emergency Diesel Generator, February 19, 2008
- EGAD-EP-10, Maintenance Rule Scoping Document, Revision 5
- EM-20-01, Emergency Diesel Generator Reliability Program, Revision 2
3 Attachment
- Emergency Diesel Generator Related Plant Operating Log Entries from May 2007 thru
September 2008
- Emergency Diesel Generators Palisades Plant Health Report, 3rd Quarter 2008
- EN-DC-198, Emergency Diesel Generator Reliability Program, Revision 1
- EN-DC-203, Maintenance Rule Program, Revision 1
- EN-DC-204, Maintenance Rule Scope and Basis, Revision 1
- EN-DC-205, Maintenance Rule Monitoring, Revision 2
- EN-DC-206, Maintenance Rule (a)(1) Process, Revision 1
- NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at
Nuclear Power Plants, Revision 2
- Palisades Plant Maintenance Rule Unavailability Performance Indicator Data thru
September 2008
1R13 Maintenance Risk Assessments and Emergent Work Control
- Procedure No. 4.02, Control of Equipment, Revision 47
- SOP-3, High Pressure Safety Injection System, Revision 75
- FSAR Chapter 9.7, Auxiliary Feedwater System, Revision 26
1R15 Operability Determinations
- Calculation 022-M-001, Control Room Required Cooling Load, Revision 0 and markup
- CR-4580, Diesel Load Calculation, November 11, 2008
- CR-PLP-2008-03993, RT-202, Control Room Heat Removal Capability Test Operability Issue,
September 24, 2008
- CR-PLP-2008-04344, Control room HVAC mounting configuration of solenoid valves to their
positioners not in accordance with design drawings, October 21, 2008
- CR-PLP-2008-04580, Diesel Load Calculation did not Account for Worst Case Containment
Air Cooler Load, November 11, 2008
- CR-PLP-2008-4431, Containment Spray Header Pressure Indicator Indicating Below
Technical Specification Limit after Maintenance, October 28, 2008
- DBD 1.06, Control Room HVAC System, Revision 7
- DBD-2.03, Containment Spray System, Revision 7
- EA-SBO-1, Station Blackout Coping Evaluation for 10CFR50.63, Revision 2
- EC Markup EC-10988, Control Room Cooling Load, Revision 0
- ESSO-1, Containment Spray Header Fill, Revision 12
- FSAR Chapter 6.2, Containment Spray, Revision 26
- FSAR Chapter 9.8, Heating, Ventilation, and Air Conditioning System, Revision 26
- MO-29, Engineered Safety System Alignment, Revision 36
- RT-202, Control Room HVAC Heat Removal Capability, Revisions 3, 9, and 10
1R18 Plant Modifications
- EC-10638, Disable Fast Transfer
1R19 Post Maintenance Testing
- CR-PLP-2006-2485, Control Rods Failed RO-19, May 2, 2006
- CR-PLP-2480, CRD-8 Failed to Drive CRD-8, May 1, 2006
- DBD-1.03, Auxiliary Feedwater System Design Basis Document, Revision 7
- FSAR Chapter 9.7, Auxiliary Feedwater System, Revision 26
4 Attachment
- Procedure No. 9.20, Attachment 3, Process Control Sheet dtd 11/5/08 for Auxiliary Feedwater
System 18-Month Test Procedure, Revision 24
- RO-127, Auxiliary Feedwater System, 18-Month Test Procedure, Revision 7
- WO #51634185, P/S-0737A, Replace Power Supply During 1R20
- WO 00168249, MC1000-03 has Failed, October 12, 2008
- WO 0019168, RO-19 Control Rod Position Verification, May 2, 2006
- WO 51631052, Control Rod Position Verification, October 12, 2007
- WO002981980, Control Rod Position Verification, May 3, 2006
- EC-9282, Provide UPS Power for Perimeter Intrusion Detection and Monitoring Equipment,
October 17, 2008
- WO 00160408, Upgrade to Security UPS System, October 17, 2008
1R22 Surveillance Testing
- ASME OM Code 2001, Subsection ISTB
- EM-09-04, Inservice Testing of Selected Safety-Related Pumps, Revision 23
- QO-15 Basis Document for Inservice Test Procedure-Component Cooling Water Pumps,
Revision 14
- QO-15, Inservice Test Procedure, Component Cooling Water Pumps, Revision 26
- RO-146, Comprehensive Pump Test Procedure, Component Cooling System Pumps P-52A,
P-52B, and P-52C, Revision 1
- RO-52, Fire Suppression Water System Functional Test; Fire Pump Capacity Test,
Revision 27
- RPS-I-7, Anticipated Transient Without Scram (ATWS) Calibration/Functional Test,
October 22, 2008
- WO 51662292, Fire Suppress Wtr Sys Pump Capacity Func, September 23, 2008
1EP4 Emergency Action Level and Emergency Plan Changes
- Attachment 9.2, 10 CFR 50.54(q) Evaluation/Screening, Revisions 17 and 49
- EI-1, Emergency Classification and Actions, Revisions 48 and 49
- Palisades Nuclear Plant Site Emergency Plan, Revisions 16 and 17
1EP6 Drill Evaluation
- Procedure SEP, Site Emergency Plan, Revision 16
- NEI-99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5
- Fourth Quarter Integrated Training Drill, November 5, 2008
- Tabletop Drill, December 17, 2008
2OS1 Access Control to Radiologically Significant Areas
- CR-PLP-2007-04378, Ensure that all Requirements Associated with Control of Very High
Radiation Areas are Properly Assessed, dated September 23, 2007
- CR-PLP-2007-4383, Concerns Have Been Made Regarding Use of the New Fuel Elevator,
dated September 24, 2007
- CR-PLP-2008-04310, West Engineered Safeguards Room Dose Rates Warranted High
Radiation Area Posting, dated October 18, 2008
- EAR 2001-0518, Evaluate Current Mechanical Stop on New Fuel Elevator for Design
Function, Revision 0
- EN-RP-100, Radworker Expectations, Revision 2
5 Attachment
- EN-RP-101, Access Control For Radiologically Controlled Areas, Revision 4
- EN-RP-108, Radiation Protection Posting, Revision 7
- EN-RP-141, Job Coverage, Revision 4
- EN-RP-151, Radiological Diving Revision 2
- QA-14-2008-PLP-01, Quality Assurance Report, Radiation Protection, dated May 14, 2008
- Radiation Work Permit 20080015 and Associated ALARA File, Containment Activities During
Power Operations, Revision 3
- Radiation Work Permit 20080108 and Associated ALARA File, Badger Testing in the Spent
Fuel Pool, Revision 0
- SC-91-095-07, New Fuel Elevator Travel Limit Switch and Control Console Modification,
Revision 1
- Snapshot Assessment, Incorporation of Radiation Protection Hold Points and Radiation
Protection Notification into Palisades Operation Procedures, June 13, 2008
4OA1 Performance Indicator Verification
- Cooling Water System Mitigating System Performance Indicator Validation Packages, Fourth
Quarter 2007 thru Third Quarter 2008
- DWO-1, Operators Daily/Weekly Items Modes 1,2,3, and 4 Basis Document, Revision 53
- DWO-1, Operators Daily/Weekly Items Modes 1,2,3, and 4, Revision 84
- MSPI Margin Report for Cooling Water System, Period ending September 2008
- LER 05000255/2008-002-00, Breaker Cubicle Failure Results in High Pressure Safety
Injection Pump Inoperability
- LER 05000255/2007-007-00, Fuel Handling Area Ventilation System Inoperable
- LER 05000255/2007-008-00, Auxiliary Feedwater Pump Inoperable in excess of technical
Specification Limits Due to a Postulated steam Line Break
- Log Entries associated with Cooling Water System Equipment, 1 October 2007 thru 30
September 2008
- Selected Palisades Plant Operating Logs from Fourth Quarter 2007 thru Third Quarter 2008
- NEI-99-02, Regulatory Assessment Performance Indicator Guideline, Revision 5
4OA2 Problem Identification and Resolution
- 06-01 OWA, Auxiliary Feedwater Pumps P-8A, B, C Suction Source, May 23, 2008
- 07-02 OWA, 2400VAC Busses 1C, 1D, 1E, May 14, 2007
- 08-01 OB, 2400/4160VAC MOC Switch Required Inspections, May 14, 2008
- 08-01 OWA, Condensate Pump P-2B Degraded Pressure, December 6, 2007
- Corrective Action X, Basis for Fuel Assembly Moves in Palisades Spent Fuel Pool,
November 6 2008
- CR-PLP-2008-3439, The Estimated Critical Position, although Well Within Acceptance
Criteria, was not as Accurate as Typical for Palisades, August 9, 2008
- CR-PLP-2008-01748, 1-1 Diesel Fuel Oil Supply Line Rubbing, April 19, 2008
- CR-PLP-2008-01749, 1-1 Diesel Fuel Oil Supply Line Rubbing, April 19, 2008
- CR-PLP-2008-05023, NRC has questioned the basis, justification, and control of initial control
room temperature assumption in the heatup analysis of the Palisades station blackout (SBO)
event, December 12, 2008
- CR-PLP-2008-05074, An Administrative Issue was Identified During Review of Documents
Provided to the NRC Related to Palisades Compliance with SBO Rule, December 17, 2008
- EA-APR-95-023, Room heatup after loss of ventilation under Appendix R scenario in the
control room, 1C and 1D switchgear rooms, battery rooms, containment area, and EDG rooms
- EM-04-24, Palisades Critical Prediction and Critical Approach, Revision 8
6 Attachment
- EN-LI-102, Corrective Action Process, Revision 10
- EN-OP-103, Reactivity Management Program, Revision 3
- FSAR Chapter 8.1.5, Station Blackout, Revision 25
- Letter to Gerald Slade, Palisades Plant Station Blackout Analysis, Safety Evaluation
(TAC No. 68578), May 20, 1991
- Letter to Gerald Slade, Palisades Plant-Station Blackout Analysis-Safety Evaluation
(TAC No. M68578), June 25, 1992
- Letter to John Ellegood, Notification of Fuel Moves Planned -Palisades Spent Fuel Pool
Region 1 to Region 2, November 6, 2008
- Letter to the NRC, Palisades Plant-Station Blackout Analysis, Safety Evaluation
(TAC No. 68578), August 1, 1991
- NET-299-01, BADGER Test Campaign at Palisades Nuclear Plant, October 8, 2008
- NUMARC 87-00, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station
Blackout at Light Water Reactors, Revision 1
- ODMI Implementation Action Plan, P-2B Condensate Pump Degraded Performance,
November 18, 2007
- Operator Work Arounds / Burdens, October 14, 2008 Index
- Palisades Administrative Procedure 4.12, Operator Work-Around Program, Revision 5
- Palisades Operator Aggregate Index Reports, July through October 2008
- CR-PLP-2006-04702, Adverse Trend in Power Supply Reliability, October 3, 2006
- EN-LI-121, Entergy Trending Process Third Quarter Report, Revision 7
- CR-PLP-2008-04279, Adverse Trend in Power Supplies, October 16, 2008
- CR-PLP-2008-03203, Emerging Trend in Control of Equipment Positioning, July 25, 2008
- SOP-24, Ventilation and Air Conditioning System, Revision 52
- Letter to NRC, Commitments to Address degraded Spent Fuel Pool Storage Rack Neutron
Absorber, August 27, 2008
- Spent Fuel Pool Shuffle Sequence, November 7, 2008
- Letter to John Ellegood, Notification of Fuel Moves Planned -Palisades Spent Fuel Pool
region 1 to region 2, November 17, 2008
4OA3 Follow-up of Events and Notices of Enforcement Discretion
- CR-PLP-2008-03328, Plant Exceeded Technical Specification Limit for Unidentified Leak
Rate, August 5, 2008
- LER 05000255/2008-05-00, Completion of Plant Shutdown required by Technical
Specifications, Revision 0
- LER 05000255/2008-06-00, Emergency Diesel Generator Inoperable in Excess of Technical
Specification Requirements, Revision 0
- CR-PLP-2008-00822, Broken Valve Keeper (Spring Seat Lock) Found on Cylinder 2L of the
1-2 Emergency Diesel Generator, February 19, 2008
- CR-PLP-2000-00556, EDG 1-2 Valve Yoke Retaining Nut on Cylinder 2L Was Not Attached,
March 21, 2000
- EPS-M-14, Diesel Generator Periodic Maintenance, Revision 17
- EPS-M-15, Diesel Generator 1-2 Refueling Frequency Maintenance, Revision 1
4OA5 Other Activities
- Procedure No. RO-128-1; Diesel Generator 1-1 24 Hour Load Run; Revision 13
- Procedure No. RO-128-2; Diesel Generator 1-2 24 Hour Load Run; Revision 12
- Calculation No. EA-ELEC-LDTAB-005; Emergency Diesel Generator 1-1 & 1-2 Steady State
Loading; Revision 8
7 Attachment
LIST OF ACRONYMS USED
CAL Confirmatory Action Letter
CAP Corrective Action Program
CFR Code of Federal Regulations
CR Condition Report
CR HVAC Control Room Heating, Ventilation and Air Conditioning
EDG Emergency Diesel Generator
HPSI High Pressure Safety Injection
IMC Inspection Manual Chapter
LCO Limiting Condition for Operations
LER Licensee Event Report
MSPI Mitigating Systems Performance Index
NEI Nuclear Energy Institute
NCV Non-Cited Violation
NRC U.S. Nuclear Regulatory Commission
PI Performance Indicator
RWP Radiation Work Permits
SBO Station Black Out
SDP Significance Determination Process
TS Technical Specification
UFSAR Updated Final Safety Analysis Report
8 Attachment