IR 05000333/2007002: Difference between revisions

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{{Adams|number = ML071340409}}
{{Adams
| number = ML071340409
| issue date = 05/11/2007
| title = IR 05000333-07-002, on 01/01/2007 - 03/31/2007, Entergy Nuclear Northeast (Entergy); James A. FitzPatrick Nuclear Power Plant; Maintenance Risk Assessment and Emergent Work Control - NRC Integrated Inspection Report
| author name = Cobey E W
| author affiliation = NRC/RGN-I/DRP/PB2
| addressee name = Dietrich P T
| addressee affiliation = Entergy Nuclear Northeast
| docket = 05000333
| license number = DPR-059
| contact person = Cobey, Eugene W. RI/DRP/PB2/610-337-5171
| document report number = IR-07-002
| document type = Inspection Report, Letter
| page count = 26
}}


{{IR-Nav| site = 05000333 | year = 2007 | report number = 002 }}
{{IR-Nav| site = 05000333 | year = 2007 | report number = 002 }}
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=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
...................................................iv
IR 05000333/2007-002; 01/01/2007 - 03/31/2007; James A. FitzPatrick Nuclear Power Plant;Maintenance Risk Assessment and Emergent Work Control.The report covered a three-month period of inspection by resident inspectors and region-basedinspectors. Two Green findings were identified, both of which were determined to be non-cited violations. 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000. A.
 
===NRC-Identified and Self-Revealing Findings===
 
===Cornerstone: Mitigating Systems===
: '''Green.'''
A Green, self-revealing, non-cited violation (NCV) of Title 10 of the Codeof Federal Regulations (CFR), Part 50, Appendix B, Criterion V, "Instructions,
Procedures, and Drawings," was identified when Entergy failed to properly implement a torus exhaust valve maintenance procedure. As a result, on February 25, 2007, valve 27AOV-118 did not open on demand to vent the torus and maintain drywell to torus differential pressure. Entergy entered this issue into their corrective action program and performed an extent of condition review.The inspectors determined that this finding more than minor because it wasassociated with the Barrier Performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Failure of the valve to operate remotely from the relay room would have required operators to open the valve locally using the manual operator in accordance with procedure Emergency Procedure 6, "Post-Accident Containment Venting and Gas Control."  The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A,
"Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance (Green)because it did not represent an actual open pathway in the physical integrity of reactor containment, or involve an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. (Section 1R13)*Green. A Green, self-revealing, NCV of 10 CFR Part 50.65(a)(4), "Requirementsfor Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" was identified when Entergy did not perform a risk assessment for planned maintenance activities when a tagout was applied on the 'B' electro-hydraulic control (EHC) pump, in conjunction with a previous emergent failure of torus exhaust outer isolation valve 27AOV-118. Entergy performed a risk assessment and entered the deficiency into their corrective action program.
 
vThe inspectors determined that this finding affected the initiating eventscornerstone; and it was was more than minor because it was similar to Example 7(f) in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that, the emergent failure of 27AOV-118, in combination with the subsequent removal of the 'B' electro hydraulic control pump availability resulted in the plant being in a higher risk category, which required risk management actions, under Entergy's on-line risk management procedure. The inspectors evaluated this finding using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management SDP," Flowchart 1, "Assessment of Risk Deficit," and determined the finding to be of very low safety significance (Green)because the finding resulted in an increase in the incremental core damage probability deficit of less than 1 x 10
-6 (actual increase was in the high 10
-8range). The inspectors determined that this finding had a cross-cutting aspect in the areaof human performance because Entergy did not incorporate appropriate risk insights into planned work activities. (Section 1R13)
 
===B.Licensee-Identified Violations===
 
None.


=REPORT DETAILS=
=REPORT DETAILS=
..........................................................1
Summary of Plant StatusThe James A. FitzPatrick Nuclear Power Plant began the inspection period operating at fullpower. On February 17, 2007, the 'B' feedwater pump inboard seal exhibited increased leakage; and, as a result, the licensee elected to downpower to approximately 50 percent power to remove the feedwater pump from service. Following repairs, the plant was returned to full power on February 21, 2007, and continued to operate at or near full power for the remainder of the inspection period.1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 1 sample)


==REACTOR SAFETY==
====a. Inspection Scope====
.........................................................11R01Adverse Weather Protection .......................................1
In mid-February, Oswego county experienced significant lake effect snowfall exceedingtotals in excess of eight feet in some areas with snowfall rates of several inches per hour. The inspectors reviewed Entergy's preparations and response to these conditions including actions specified in Supplemental Action Procedure 19, "Severe Weather," and Administrative Procedure (AP) 12.04, "Seasonal Weather Preparations."  In addition, the inspectors verified that Entergy took action to ensure that adequate operator and onsite staff were available during the storm. The inspectors also verified the operability of offsite and onsite emergency power supplies and that control room operators communicated with the transmission system operators in accordance with AP 12.13, "345/115 kV [kilovolt] Transmission Line Operations and Interface."  This inspection satisfied one inspection sample for the onset of adverse weather.


====b. Findings====
No findings of significance were identified.
{{a|1R04}}
{{a|1R04}}
==1R04 Equipment Alignment ............................................1==
==1R04 Equipment Alignment (71111.04Q - 3 samples, 71111.04S - 1 sample).1Partial System Walkdown (3 samples)==
 
====a. Inspection Scope====
The inspectors performed three partial system walkdowns to verify the operability ofredundant or diverse trains and components during periods of system train unavailability or following periods of maintenance. The inspectors referenced the system procedures, the Updated Final Safety Analysis Report (UFSAR), and system drawings in order to verify that the alignment of the available train was proper to support its required safety functions. The inspectors also reviewed applicable condition reports (CR) and work orders to ensure that Entergy had identified and properly addressed equipment discrepancies that could potentially impair the capability of the available train, as required by 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action."  The 2Enclosuredocuments reviewed are listed in the Attachment. The inspectors performed a partialwalkdown on the following systems which represented three inspection samples:*Train 'A' low pressure coolant injection (LPCI) independent power supply systemduring testing of the LPCI battery and battery charger;*Train 'B' emergency service water and emergency diesel generator systemsduring planned maintenance on the 'A' emergency service water train; and*Train 'B' residual heat removal system when the 'A' residual heat removal systemwas out of service for testing.
 
====b. Findings====
No findings of significance were identified..2Complete System Walkdown (1 sample)
 
====a. Inspection Scope====
The inspectors performed a complete system alignment inspection of the containmentatmosphere control and dilution system to identify any discrepancies between the existing equipment lineup and the required lineup. During the inspection, system drawings and operating procedures were used to verify proper equipment alignment and operational status. The inspectors reviewed the open maintenance work orders associated with the system for any deficiencies that could affect the ability of the system to perform its function. Documentation associated with unresolved design issues such as temporary modifications, operator workarounds, and items tracked by plant engineering were also reviewed to assess their collective impact on system operation.
 
In addition, the inspectors reviewed the CR database to verify that equipment problems were being identified and appropriately resolved. The documents reviewed during this inspection are listed in the Attachment. The inspection represented one inspection
 
sample.


====b. Findings====
No findings of significance were identified.
{{a|1R05}}
{{a|1R05}}
==1R05 Fire Protection .................................................2==
==1R05 Fire Protection (71111.05Q - 9 samples)==


====a. Inspection Scope====
The inspectors conducted a tour of the areas listed below to assess the materialcondition and operational status of fire protection features. The inspectors verified that:
combustibles and ignition sources were controlled in accordance with Entergy's administrative procedures;  fire detection and suppression equipment was available for use; passive fire barriers were maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire plan. The inspectors used procedure ENN-DC-161, 3Enclosure"Transient Combustible Program," in performing the inspection. The inspectorsevaluated the fire protection program against the requirements of License Condition 2.C.3. The documents reviewed are listed in the Attachment. This inspection represented nine inspection samples for fire protection tours and were conducted in the following areas:: *Fire Area/Zone 1E/TB-1 North, elevation 252 foot;*Fire Area/Zone 1E-TB-1 South, elevation 252 foot;
*Fire Area/Zone 1E/TB-1 North, elevation 272 foot;
*Fire Area/Zone IX/SG-1;
*Fire Area/Zone IX/RB-1A, elevation 272 foot;
*Fire Area/Zone II/SW-1, elevation 272 foot;
*Fire Area/Zone IC/SW-1, elevation 272 foot;
*Fire Area/Zone V/EG-1 South, elevation 272 foot; and
*Fire Area/Zone VII/CS-1, elevation 272 foot.
====b. Findings====
No findings of significance were identified.
{{a|1R06}}
{{a|1R06}}
==1R06 Flood Protection Measures ........................................31R07Heat Sink Performance ...........................................3==
==1R06 Flood Protection Measures Internal Flooding (71111.06 -1 sample)==


{{a|1R11}}
====a. Inspection Scope====
==1R11 Licensed Operator Requalification Program............................4==
The inspectors reviewed selected risk-important plant design features and Entergy'sprocedures intended to protect the cable spreading and relay rooms and associated safety-related equipment from internal flooding events. The inspectors reviewed flood analysis and design documents, including the Individual Plant Examination and the UFSAR, engineering calculations, and abnormal operating procedures. The documents reviewed are listed in the Attachment. These activities represented one inspection
 
sample.


{{a|1R12}}
====b. Findings====
==1R12 Maintenance Effectiveness ........................................41R13Maintenance Risk Assessments and Emergent Work Control .............51R15Operability Evaluations ...........................................91R19Post Maintenance Testing.........................................91R22Surveillance Testing ............................................10==
No findings of significance were identified.
{{a|1R07}}
==1R07 Heat Sink Performance (71111.07A - 1 sample)==


1EP6Drill Evaluation
====a. Inspection Scope====
The inspectors reviewed the testing and evaluation of test results for the residual heatremoval system heat exchanger 10E-2B performed in accordance with Entergy's response to NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment."  Heat removal measurements and heat exchanger capacity calculations were reviewed to verify that cooler performance was consistent with design 4Enclosurecalculations and the UFSAR. The documents reviewed are listed in the Attachment. These activities represented one inspection sample.


==RADIATION SAFETY==
====b. Findings====
.......................................................112OS2ALARA Planning and Controls
No findings of significance were identified.
{{a|1R11}}
==1R11 Licensed Operator Requalification Program.1Resident Inspector Quarterly Review (71111.11Q - 1 sample)==


==OTHER ACTIVITIES==
====a. Inspection Scope====
........................................................124OA2Identification and Resolution of Problems............................124OA3Event Followup ................................................13 4OA5Other Activities ................................................13 4OA6Meetings, Including Exit..........................................14ATTACHMENT:
On March 1, 2007, the inspectors observed licensed operator simulator training toassess operator performance during several scenarios to verify that operator performance was adequate and evaluators were identifying and documenting crew performance problems. The inspectors evaluated the performance of risk significant operator actions, including the use of emergency operating procedures. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance of timely control board operation and manipulation, and the oversight and direction provided by the shiftmanager. The inspectors also reviewed simulator fidelity to evaluate the degree of similarity to the actual control room. Licensed operator training was evaluated against the requirements of 10 CFR Part 55, "Operators' Licenses."  The documents reviewed are listed in the Attachment. This observation of operator simulator training represented one inspection sample.


=SUPPLEMENTAL INFORMATION=
====b. Findings====
No findings of significance were identified.
{{a|1R12}}
==1R12 Maintenance Effectiveness (71111.12Q - 2 samples)==


==KEY POINTS OF CONTACT==
====a. Inspection Scope====
................................................A-1
The inspectors reviewed performance-based problems involving selected in-scopestructures, systems, or components (SSCs) to assess the effectiveness of the maintenance program. The reviews focused on:*Proper Maintenance Rule scoping in accordance with 10 CFR Part 50.65;*Characterization of reliability issues;
==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
...........................A-1
==LIST OF DOCUMENTS REVIEWED==
..........................................A-1
==LIST OF ACRONYMS==
......................................................A-5
ivSUMMARY
: [[OF]] [[]]
FINDINGSIR 05000333/2007-002; 01/01/2007 - 03/31/2007; James A. FitzPatrick Nuclear Power Plant;Maintenance Risk Assessment and Emergent Work Control.The report covered a three-month period of inspection by resident inspectors and region-basedinspectors. Two Green findings were identified, both of which were determined to be non-cited
violations. 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]]
NRC's program for overseeing the safe operation of
commercial nuclear power reactors is described in
: [[NUR]] [[]]
EG-1649, "Reactor Oversight Process,"
Revision 3, dated July 2000.
: [[A.N]] [[]]
RC-Identified and Self-Revealing FindingsCornerstone: Mitigating Systems
*Green. A Green, self-revealing, non-cited violation (NCV) of Title 10 of the Codeof Federal Regulations (CFR), Part 50, Appendix B, Criterion V, "Instructions,
Procedures, and Drawings," was identified when Entergy failed to properly
implement a torus exhaust valve maintenance procedure. As a result, on
February 25, 2007, valve 27AOV-118 did not open on demand to vent the torus
and maintain drywell to torus differential pressure. Entergy entered this issue
into their corrective action program and performed an extent of condition review.The inspectors determined that this finding more than minor because it wasassociated with the Barrier Performance attribute of the Barrier Integrity
cornerstone; and it impacted the cornerstone objective of providing reasonable
assurance that physical design barriers protect the public from radionuclide
releases caused by accidents or events. Failure of the valve to operate remotely
from the relay room would have required operators to open the valve locally
using the manual operator in accordance with procedure Emergency
Procedure 6, "Post-Accident Containment Venting and Gas Control."  The
inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A,
"Significance Determination of Reactor Inspection Findings for At-Power
Situations," and determined it to be of very low safety significance (Green)
because it did not represent an actual open pathway in the physical integrity of
reactor containment, or involve an actual reduction in defense-in-depth for the
atmospheric pressure control or hydrogen control functions of the reactor
containment.  (Section 1R13)*Green. A Green, self-revealing,
: [[NCV]] [[of 10]]
CFR Part 50.65(a)(4), "Requirementsfor Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" was
identified when Entergy did not perform a risk assessment for planned
maintenance activities when a tagout was applied on the 'B' electro-hydraulic
control (EHC) pump, in conjunction with a previous emergent failure of torus
exhaust outer isolation valve 27AOV-118. Entergy performed a risk assessment
and entered the deficiency into their corrective action program.
vThe inspectors determined that this finding affected the initiating eventscornerstone; and it was was more than minor because it was similar to Example
7(f) in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor
Issues," in that, the emergent failure of 27AOV-118, in combination with the
subsequent removal of the 'B' electro hydraulic control pump availability resulted
in the plant being in a higher risk category, which required risk management
actions, under Entergy's on-line risk management procedure. The inspectors
evaluated this finding using IMC 0609, Appendix K, "Maintenance Risk
Assessment and Risk Management SDP," Flowchart 1, "Assessment of Risk
Deficit," and determined the finding to be of very low safety significance (Green)
because the finding resulted in an increase in the incremental core damage
probability deficit of less than 1 x 10-6 (actual increase was in the high 10-8range). The inspectors determined that this finding had a cross-cutting aspect in the areaof human performance because Entergy did not incorporate appropriate risk
insights into planned work activities. (Section 1R13)B.Licensee-Identified ViolationsNone.
: [[REPORT]] [[]]
DETAILSSummary of Plant StatusThe James A. FitzPatrick Nuclear Power Plant began the inspection period operating at fullpower. On February 17, 2007, the 'B' feedwater pump inboard seal exhibited increased
leakage; and, as a result, the licensee elected to downpower to approximately 50 percent power
to remove the feedwater pump from service. Following repairs, the plant was returned to full
power on February 21, 2007, and continued to operate at or near full power for the remainder of
the inspection period.1.REACTOR
: [[SAFET]] [[]]
YCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 1 sample)  a.Inspection ScopeIn mid-February, Oswego county experienced significant lake effect snowfall exceedingtotals in excess of eight feet in some areas with snowfall rates of several inches per
hour. The inspectors reviewed Entergy's preparations and response to these conditions
including actions specified in Supplemental Action Procedure 19, "Severe Weather," and
Administrative Procedure (AP) 12.04, "Seasonal Weather Preparations."  In addition, the
inspectors verified that Entergy took action to ensure that adequate operator and onsite
staff were available during the storm. The inspectors also verified the operability of
offsite and onsite emergency power supplies and that control room operators
communicated with the transmission system operators in accordance with AP 12.13,
"345/115 kV [kilovolt] Transmission Line Operations and Interface."  This inspection
satisfied one inspection sample for the onset of adverse weather.
b.FindingsNo findings of significance were identified.1R04Equipment Alignment (71111.04Q - 3 samples, 71111.04S - 1 sample).1Partial System Walkdown (3 samples)  a.Inspection Scope The inspectors performed three partial system walkdowns to verify the operability ofredundant or diverse trains and components during periods of system train unavailability
or following periods of maintenance. The inspectors referenced the system procedures,
the Updated Final Safety Analysis Report (UFSAR), and system drawings in order to
verify that the alignment of the available train was proper to support its required safety
functions. The inspectors also reviewed applicable condition reports (CR) and work
orders to ensure that Entergy had identified and properly addressed equipment
discrepancies that could potentially impair the capability of the available train, as
required by
: [[10 CFR]] [[Part 50, Appendix B, Criterion]]
XVI, "Corrective Action."  The
2Enclosuredocuments reviewed are listed in the Attachment. The inspectors performed a partialwalkdown on the following systems which represented three inspection samples:*Train 'A' low pressure coolant injection (LPCI) independent power supply systemduring testing of the
: [[LP]] [[]]
CI battery and battery charger;*Train 'B' emergency service water and emergency diesel generator systemsduring planned maintenance on the 'A' emergency service water train; and*Train 'B' residual heat removal system when the 'A' residual heat removal systemwas out of service for testing. b.FindingsNo findings of significance were identified..2Complete System Walkdown (1 sample)  a.Inspection Scope The inspectors performed a complete system alignment inspection of the containmentatmosphere control and dilution system to identify any discrepancies between the
existing equipment lineup and the required lineup. During the inspection, system
drawings and operating procedures were used to verify proper equipment alignment and
operational status. The inspectors reviewed the open maintenance work orders
associated with the system for any deficiencies that could affect the ability of the system
to perform its function. Documentation associated with unresolved design issues such
as temporary modifications, operator workarounds, and items tracked by plant
engineering were also reviewed to assess their collective impact on system operation.
In addition, the inspectors reviewed the CR database to verify that equipment problems
were being identified and appropriately resolved. The documents reviewed during this
inspection are listed in the Attachment. The inspection represented one inspection
sample. b.FindingsNo findings of significance were identified.1R05Fire Protection (71111.05Q - 9 samples)  a.Inspection ScopeThe inspectors conducted a tour of the areas listed below to assess the materialcondition and operational status of fire protection features. The inspectors verified that:
combustibles and ignition sources were controlled in accordance with Entergy's
administrative procedures;  fire detection and suppression equipment was available for
use; passive fire barriers were maintained; and compensatory measures for out-of-
service, degraded, or inoperable fire protection equipment were implemented in
accordance with Entergy's fire plan. The inspectors used procedure
: [[ENN]] [[-]]
DC-161,
3Enclosure"Transient Combustible Program," in performing the inspection. The inspectorsevaluated the fire protection program against the requirements of License Condition
2.C.3. The documents reviewed are listed in the Attachment. This inspection
represented nine inspection samples for fire protection tours and were conducted in the
following areas:: *Fire Area/Zone 1E/TB-1 North, elevation 252 foot;*Fire Area/Zone 1E-TB-1 South, elevation 252 foot;
*Fire Area/Zone 1E/TB-1 North, elevation 272 foot;
*Fire Area/Zone
: [[IX]] [[/]]
SG-1;
*Fire Area/Zone
: [[IX]] [[/]]
RB-1A, elevation 272 foot;
*Fire Area/Zone
: [[II]] [[/]]
SW-1, elevation 272 foot;
*Fire Area/Zone
: [[IC]] [[/]]
SW-1, elevation 272 foot;
*Fire Area/Zone V/EG-1 South, elevation 272 foot; and
*Fire Area/Zone
: [[VII]] [[/]]
CS-1, elevation 272 foot. b.FindingsNo findings of significance were identified.1R06Flood Protection Measures Internal Flooding (71111.06 -1 sample)  a.Inspection ScopeThe inspectors reviewed selected risk-important plant design features and Entergy'sprocedures intended to protect the cable spreading and relay rooms and associated
safety-related equipment from internal flooding events. The inspectors reviewed flood
analysis and design documents, including the Individual Plant Examination and the
: [[UFS]] [[]]
AR, engineering calculations, and abnormal operating procedures. The documents
reviewed are listed in the Attachment. These activities represented one inspection
sample. b.FindingsNo findings of significance were identified.1R07Heat Sink Performance (71111.07A - 1 sample)  a.Inspection ScopeThe inspectors reviewed the testing and evaluation of test results for the residual heatremoval system heat exchanger 10E-2B performed in accordance with Entergy's
response to NRC Generic Letter 89-13, "Service Water System Problems Affecting
Safety-Related Equipment."  Heat removal measurements and heat exchanger capacity
calculations were reviewed to verify that cooler performance was consistent with design
4Enclosurecalculations and the
: [[UFS]] [[]]
AR. The documents reviewed are listed in the Attachment. These activities represented one inspection sample. b.FindingsNo findings of significance were identified.1R11Licensed Operator Requalification Program.1Resident Inspector Quarterly Review (71111.11Q - 1 sample)  a.Inspection ScopeOn March 1, 2007, the inspectors observed licensed operator simulator training toassess operator performance during several scenarios to verify that operator
performance was adequate and evaluators were identifying and documenting crew
performance problems. The inspectors evaluated the performance of risk significant
operator actions, including the use of emergency operating procedures. The inspectors
assessed the clarity and effectiveness of communications, the implementation of
appropriate actions in response to alarms, the performance of timely control board
operation and manipulation, and the oversight and direction provided by the shiftmanager. The inspectors also reviewed simulator fidelity to evaluate the degree of
similarity to the actual control room. Licensed operator training was evaluated against
the requirements of 10 CFR Part 55, "Operators' Licenses."  The documents reviewed
are listed in the Attachment. This observation of operator simulator training represented
one inspection sample. b.FindingsNo findings of significance were identified.1R12Maintenance Effectiveness (71111.12Q - 2 samples)  a.Inspection ScopeThe inspectors reviewed performance-based problems involving selected in-scopestructures, systems, or components (SSCs) to assess the effectiveness of the
maintenance program. The reviews focused on:*Proper Maintenance Rule scoping in accordance with 10 CFR Part 50.65;*Characterization of reliability issues;
* Changing system and component unavailability;
* Changing system and component unavailability;
* 10 CFR Part 50.65 (a)(1) and (a)(2) classifications;
* 10 CFR Part 50.65 (a)(1) and (a)(2) classifications;
*Identifying and addressing common cause failures;
*Identifying and addressing common cause failures;
* Trending of system flow and temperature values;
* Trending of system flow and temperature values;
* Appropriateness of performance criteria for SSCs classified (a)(2); and  
* Appropriateness of performance criteria for SSCs classified (a)(2); and
* Adequacy of goals and corrective actions for SSCs classified (a)(1).
* Adequacy of goals and corrective actions for SSCs classified (a)(1).
5EnclosureThe inspectors reviewed system health reports, maintenance backlogs, andMaintenance Rule basis documents. The inspectors evaluated the maintenance
 
program against the requirements of 10 CFR Part 50.65. The documents reviewed are
5EnclosureThe inspectors reviewed system health reports, maintenance backlogs, andMaintenance Rule basis documents. The inspectors evaluated the maintenance program against the requirements of 10 CFR Part 50.65. The documents reviewed are listed in the Attachment. The following Maintenance Rule samples were reviewed and represent two inspection samples:*125 volt station and 419 volt low pressure coolant injection batteries, chargers,and inverters; and*Residual heat removal service water system.
listed in the Attachment. The following Maintenance Rule samples were reviewed and
 
represent two inspection samples:*125 volt station and 419 volt low pressure coolant injection batteries, chargers,and inverters; and*Residual heat removal service water system. b.FindingsNo findings of significance were identified.1R13Maintenance Risk Assessments and Emergent Work Control (71111.13 - 5 samples) a.Inspection Scope The inspectors reviewed maintenance activities to verify that the appropriate riskassessments were performed prior to removing equipment for work. The inspectors
====b. Findings====
verified that risk assessments were performed as required by 10 CFR 50.65(a)(4), and
No findings of significance were identified.
were accurate and complete. When emergent work was performed, the inspectors
{{a|1R13}}
verified that the plant risk was promptly reassessed and managed. The documents
==1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 5 samples)==
reviewed are listed in the Attachment. The review of the following activities represented
 
five inspection samples:*Week of January 8, 2007, which included biennial maintenance on the 'A' trainlow pressure coolant injection inverter and full load surveillance testing of the 'A'
====a. Inspection Scope====
train emergency diesel generator; *Week of January 22, 2007, which included full load testing of 'B' train emergencydiesel generators, quarterly inservice testing of the high pressure coolant
The inspectors reviewed maintenance activities to verify that the appropriate riskassessments were performed prior to removing equipment for work. 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 documents reviewed are listed in the Attachment. The review of the following activities represented five inspection samples:*Week of January 8, 2007, which included biennial maintenance on the 'A' trainlow pressure coolant injection inverter and full load surveillance testing of the 'A' train emergency diesel generator; *Week of January 22, 2007, which included full load testing of 'B' train emergencydiesel generators, quarterly inservice testing of the high pressure coolant injection system, and service testing of the 'B' train low pressure coolant injection battery and inverter; *Week of February 19, 2007, which included emergent work on the 'B' feedwaterpump due to a degraded inboard seal and 'C' condensate pump packing replacement;*Week of February 26, 2007, which included emergent work on the torus exhaustouter isolation valve, 27-AOV-118, emergency diesel generator surveillance testing, and high pressure coolant injection system instrument surveillance testing; and*Week of March 12, 2007, which included 'B' reactor water cleanup pump sealfailure and system isolation with the 'A' steam packing exhauster and 'C' residual heat removal service water pump out of service.
injection system, and service testing of the 'B' train low pressure coolant injection
 
battery and inverter; *Week of February 19, 2007, which included emergent work on the 'B' feedwaterpump due to a degraded inboard seal and 'C' condensate pump packing
====b. Findings====
replacement;*Week of February 26, 2007, which included emergent work on the torus exhaustouter isolation valve, 27-AOV-118, emergency diesel generator surveillance
.1Introduction:  A Green, self-revealing, non-cited violation (NCV) of 10 CFR 50,Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified when Entergy failed to properly implement a torus exhaust valve maintenance. As a result, on February 25, 2007, valve 27AOV-118 did not open on demand to vent the torus to maintain drywell-to-torus differential pressure.Description:  Valve 27AOV-118 is a normally closed 20-inch air-operated containmentisolation valve in the containment atmosphere control system. During normal operation the valve is remotely operated from a relay room control panel to vent the torus via the standby gas treatment system to maintain drywell-to-torus differential pressure. During loss of normal decay heat removal events, the valve is used for containment pressure control and reactor decay heat removal. The system is manually initiated from the relay room in accordance with emergency operating procedures when drywell pressure reaches 44 psig. The valve can also be operated using a separately mounted manual operator.On February 25, 2007, while attempting to vent the torus, Entergy identified that valve27AOV-118 could not be opened from the relay room. Investigation revealed that the key connecting the air actuator coupling hub to the valve coupling hub had sheared and that the gap between the hubs appeared to be excessive. A similar key failure occurred in 1996 and had been attributed to an excessive coupling hub gap which overstressed the key. Based on the results of calculation JAF-CALC-CAD-02766, "Containment Purge Actuator to Valve Coupling Gap for Valves 27AOV-101A/B and 111 Through 118," an allowable range of 0.0625 to 0.125 inches was established and incorporated on December 19, 2002, into the valve actuator Maintenance Procedure (MP) 060.02, "GH Bettis Pneumatic Valve Actuator Maintenance," Revision 5. Following the February 25, 2007 key failure, Entergy found the coupling hub gap to be 0.625 inches.
testing, and high pressure coolant injection system instrument surveillance
 
testing; and*Week of March 12, 2007, which included 'B' reactor water cleanup pump sealfailure and system isolation with the 'A' steam packing exhauster and 'C' residual
=====Analysis:=====
heat removal service water pump out of service.
The inspectors determined that the performance deficiency was that Entergyfailed to implement the actuator-to-valve coupling gap limit specified in the maintenance procedure. This was reasonably within Entergy's ability to foresee and prevent.
6Enclosure  b.Findings.1Introduction:  A Green, self-revealing, non-cited violation (NCV) of 10 CFR 50,Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified when
 
Entergy failed to properly implement a torus exhaust valve maintenance. As a result, on
Traditional enforcement does not apply because the issue did not have an actual safety consequence or a potential for impacting the NRC's regulatory function, and it was not the result of any willful violation of NRC requirements.The inspectors determined that this finding was more than minor because it wasassociated with the Barrier Performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or events. Failure of the valve to operate remotely from the relay room would have required operators to open the valve locally using the manual operator in accordance with procedure Emergency Procedure 6, "Post-Accident Containment Venting and Gas Control."  The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance (Green)7Enclosurebecause it did not represent an actual open pathway in the physical integrity of reactorcontainment, or involve an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment.Enforcement:  10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, andDrawings," requires that activities affecting quality shall be accomplished in accordance with instructions, procedures, or drawings. Entergy Maintenance Procedure (MP)060.02, "GH Bettis Pneumatic Valve Actuator Maintenance," Revision 5, requires adjustment of the valve stem coupling hub to hub gap be 0.0625 to 0.125 inches on both the manual and air operator side of valve 27AOV-118 as measured at the key interface locations. Contrary to the above, in October 2004, Entergy did not maintain the valve stem coupling hub to hub gap on valve 27AOV-118 in accordance with maintenance procedure MP-060.02 resulting in failure of the actuator-to-valve coupling hub key on February 25, 2007. Because the issue was of very low safety significance (Green) and was entered into Entergy's corrective action program as condition report CR-JAF-2007-00752, this violation is being treated as an NCV consistent with Section VI.A.1 of theNRC Enforcement Policy:  (NCV 05000333/2007002-01, Inadequate Maintenance onContainment Atmosphere Control Valve.).
February 25, 2007, valve 27AOV-118 did not open on demand to vent the torus to
 
maintain drywell-to-torus differential pressure.Description:  Valve 27AOV-118 is a normally closed 20-inch air-operated containmentisolation valve in the containment atmosphere control system. During normal operation
=====Introduction:=====
the valve is remotely operated from a relay room control panel to vent the torus via the
A Green, self-revealing NCV of 10 CFR Part 50.65 (a)(4), "Requirementsfor Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" was identified when Entergy did not perform a risk assessment for planned maintenance activities when a tagout was applied on the 'B' electro-hydraulic control (EHC) pump, in conjunction with a previous emergent failure of torus exhaust outer isolation valve 27AOV-118.
standby gas treatment system to maintain drywell-to-torus differential pressure. During
 
loss of normal decay heat removal events, the valve is used for containment pressure
=====Description:=====
control and reactor decay heat removal. The system is manually initiated from the relay
On February 25, 2007, at 11:10 p.m., 27AOV-118 would not open remotelyto vent the torus. 27AOV-118 is a primary containment isolation valve, and, as required by TS 3.6.1.3, operators isolated the affected containment penetration flow path by closing and deactivating the inboard isolation valve. Subsequently, a tagout was applied on the 'B' EHC pump for scheduled maintenance. A risk assessment that considered the emergent failure of 27AOV-118 and the impact on containment venting prior to application of the tagout on the 'B' EHC pump was not performed. Specifically, Entergy's risk assessment failed to consider risk significant systems, structures, andcomponents, as well as support systems that were unavailable during the maintenance.
room in accordance with emergency operating procedures when drywell pressure
 
reaches 44 psig. The valve can also be operated using a separately mounted manual
Administrative Procedure (AP) 10.10, "On-Line Risk Assessment," assigns a risk category color in risk significant order from Green, Yellow, Orange or Red based on core damage frequency. Risk management actions are implemented depending on the risk category color. With 27AOV-118 inoperable, combined with the impact of the tagout on the EHC system, the color increased from Yellow to Orange. On February 26, at approximately 6:46 a.m., the plant staff recognized the condition and the 'B' EHC pump was returned to standby.Analysis:  The inspectors determined that the finding was a performance deficiencybecause Entergy did not perform a risk assessment following the emergent failure of 27AOV-118, and subsequently continued scheduled work activities including tagout of 8Enclosurethe 'B' EHC pump. The inspectors determined that this finding impacted the InitiatingEvents cornerstone due to the increased likelihood of a plant transient with the loss of the normal heat sink. It was reasonable that Entergy should have identified the condition and updated the risk assessment, because Entergy procedure AP 10.10, "On-Line Risk Assessment," specifies that plant risk must be reassessed when plant conditions change. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Entergy's procedures. The inspectors determined that this finding was more than minor because it was similarto Example 7(f) in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that, the emergent failure of 27AOV-118, in combination with the subsequent removal of the 'B' electro hydraulic control pump availability resulted in the plant being in a higher risk category, which required risk management actions, under Entergy's on-line risk management procedure. The ability to vent the containment from the torus is a risk important action to prevent containment failure and core damage for situations that involve the inability to remove core decay heat from the torus water. The tagout of the
operator.On February 25, 2007, while attempting to vent the torus, Entergy identified that valve27AOV-118 could not be opened from the relay room. Investigation revealed that the
'B' electro hydraulic control pump increased the chance of a plant transient with the loss of the normal heat sink.Using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk ManagementSDP," Flowchart 1, "Assessment of Risk Deficit," the inspectors determined the incremental core damage probability deficit from Entergy's core damage frequency as a result of the actual duration of the 27AOV-118 maintenance combined with the time the
key connecting the air actuator coupling hub to the valve coupling hub had sheared and
'B' EHC system was not available due to the tagout (two hours). The inspectors calculated the incremental core damage probability deficit to be in the high 10
that the gap between the hubs appeared to be excessive. A similar key failure occurred
-8 range. Because the calculated risk deficit was not greater than 1 x 10
in 1996 and had been attributed to an excessive coupling hub gap which overstressed
-6 incremental coredamage probability deficit, the inspectors determined that this finding was of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance because Entergy did not incorporate appropriate risk insights into planned work activities.Enforcement:  10 CFR 50.65 (a)(4), requires, in part, that before performingmaintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities.
the key. Based on the results of calculation
 
: [[JAF]] [[-]]
Contrary to the above, on February 26, 2007, Entergy did not assess and manage the increase in risk for planned maintenance activities when a tagout was applied on the 'B' EHC pump, following an emergent failure of torus exhaust outer isolation valve 27AOV-118. Because this finding was of very low safety significance and was entered into Entergy's corrective action program as condition report CR-JAF-2007-00755, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000333/2007002-02, Failure to Perform a Risk Assessment When Required by 10 CFR 50.65(a)(4).
CALC-CAD-02766, "Containment
 
Purge Actuator to Valve Coupling Gap for Valves 27AOV-101A/B and 111 Through
9Enclosure1R15Operability Evaluations (71111.15 - 6 samples)
118," an allowable range of 0.0625 to 0.125 inches was established and incorporated on
 
December 19, 2002, into the valve actuator Maintenance Procedure (MP) 060.02, "GH
====a. Inspection Scope====
Bettis Pneumatic Valve Actuator Maintenance," Revision 5. Following the February 25,
The inspectors reviewed operability determinations to assess the acceptability of theevaluations; when needed, the use and control of compensatory measures; and compliance with TS. The inspectors' review included a verification that the operability determinations were made as specified by ENN-OP-104, "Operability Determinations."
2007 key failure, Entergy found the coupling hub gap to be 0.625 inches. Analysis: The inspectors determined that the performance deficiency was that Entergyfailed to implement the actuator-to-valve coupling gap limit specified in the maintenance
 
procedure. This was reasonably within Entergy's ability to foresee and prevent.
The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR, and associated design basis documents. The documents reviewed are listed in the Attachment. The following evaluations were reviewed and represented six inspection samples:*CR-2007-00151 concerning a commercial grade gate timing circuit card installedin safety-related low pressure coolant injection inverter 71INV-3A;*CR-2007-00293 concerning installation of nonsafety-related contacts insafety-related circuit breaker 71MCC-163-OG-1 for containment isolation valve 10MOV-34B; *CR-2007-00392 concerning an out of calibration condition of a local leak ratemonitor panel that affected the test results of 12 containment penetrations;*CR-2007-00647 concerning inaccurate valve position indication of vacuumbreaker 27VB-2 while performing surveillance test ST-15J, "Torus to Drywell Vacuum Breakers Quarterly Test IST;"*CR-2007-00281 concerning 'D' safety relief valve tailpipe temperature increase; and*CR-2007-00752 concerning extent of condition reviews following failure ofcontainment atmosphere control isolation valve 27AOV-118.
Traditional enforcement does not apply because the issue did not have an actual safety
 
consequence or a potential for impacting the NRC's regulatory function, and it was not
====b. Findings====
the result of any willful violation of NRC requirements.The inspectors determined that this finding was more than minor because it wasassociated with the Barrier Performance attribute of the Barrier Integrity cornerstone;
No findings of significance were identified.
and it impacted the cornerstone objective of providing reasonable assurance that
{{a|1R19}}
physical design barriers protect the public from radio nuclide releases caused by
==1R19 Post Maintenance Testing (71111.19 - 6 samples)==
accidents or events. Failure of the valve to operate remotely from the relay room would
 
have required operators to open the valve locally using the manual operator in
====a. Inspection Scope====
accordance with procedure Emergency Procedure 6, "Post-Accident Containment
The inspectors reviewed six post-maintenance test procedures and associated testingactivities for selected risk significant mitigating systems to assess whether the effect of maintenance on plant systems was adequately addressed by control room and engineering personnel. The inspectors verified:  test acceptance criteria were clear, demonstrated operational readiness and were consistent with Design Basis Documents; test instrumentation had current calibrations and adequate range and accuracy for the application; and tests were performed, as written, with applicable prerequisites satisfied.
Venting and Gas Control."  The inspectors evaluated this finding using Phase 1 of IMC
 
0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-
Upon completion, the inspectors verified that equipment was returned to the proper alignment necessary to perform its safety function. Post-maintenance testing was evaluated against the requirements of 10 CFR 50, Appendix B, Criterion XI, "Test 10EnclosureControl."  The documents reviewed are listed in the Attachment. The following post-maintenance test activities were reviewed and represented six inspection samples:*Work request JAF-03-07188-01, involving repair of a water leak in service waterpump 46P-1B motor oil reservoir during the week of June 16, 2007;*Work request JAF-05-17157, involving inspection and repair of a residual heatremoval service water keep-full check valve;*Work request JAF-05-22322, involving cleaning and replacement of two-inchemergency service water piping to the west electric bay and east cable tunnel coolers;*Work request JAF-07-15650, involving repair of the air operator coupling of torusexhaust outboard containment isolation valve 27AOV-118;*Work request 51104288, involving replacement of residual heat removal servicewater pump 10P-1C due to high vibration; and*Work request  51178018, involving maintenance and repair of stator watercooling pump 94P-15B due to high vibration.
Power Situations," and determined it to be of very low safety significance (Green)
 
7Enclosurebecause it did not represent an actual open pathway in the physical integrity of reactorcontainment, or involve an actual reduction in defense-in-depth for the atmospheric
====b. Findings====
pressure control or hydrogen control functions of the reactor containment.Enforcement:  10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, andDrawings," requires that activities affecting quality shall be accomplished in accordance
No findings of significance were identified.
with instructions, procedures, or drawings. Entergy Maintenance Procedure (MP)
{{a|1R22}}
060.02, "GH Bettis Pneumatic Valve Actuator Maintenance," Revision 5, requires
==1R22 Surveillance Testing (71111.22 - 7 samples)==
adjustment of the valve stem coupling hub to hub gap be 0.0625 to 0.125 inches on both
 
the manual and air operator side of valve 27AOV-118 as measured at the key interface
====a. Inspection Scope====
locations. Contrary to the above, in October 2004, Entergy did not maintain the valve
The inspectors witnessed performance of surveillance tests and/or reviewed test data ofselected risk-significant SSCs to assess whether the SSCs satisfied TS, UFSAR, Technical Requirements Manual, and Entergy procedure requirements. The inspectors verified:  test acceptance criteria were clear, demonstrated operational readiness, and were consistent with Design Basis Documents; test instrumentation had current calibrations and adequate range and accuracy for the application; and tests were performed, as written, with applicable prerequisites satisfied. Upon surveillance test (ST) completion, the inspectors verified that equipment was returned to the status specified to perform its safety function. The inspectors evaluated the tests against the requirements in TS. The following STs were reviewed and represented seven inspection program samples:*ST-2AM, "Residual Heat Removal Loop B Quarterly Operability Test;"*ST-4N, "High Pressure Coolant Injection Quick Start, Inservice, and TransientMonitoring Test;"  *MST-071.30, "Low Pressure Coolant Injection Charger-Inverter Performance andLow Pressure Coolant Injection Battery Service Surveillance Test;" *ST-7F, "Standby Gas Treatment Fan B and Valve Exercising Test IST;"
stem coupling hub to hub gap on valve 27AOV-118 in accordance with maintenance
procedure MP-060.02 resulting in failure of the actuator-to-valve coupling hub key on
February 25, 2007. Because the issue was of very low safety significance (Green) and
was entered into Entergy's corrective action program as condition report
: [[CR]] [[-]]
JAF-2007-
00752, this violation is being treated as an
: [[NCV]] [[consistent with Section]]
: [[VI.A.]] [[1 of theNRC Enforcement Policy:  (NCV 05000333/2007002-01, Inadequate Maintenance onContainment Atmosphere Control Valve.).2Introduction: A Green, self-revealing]]
: [[NCV]] [[of 10]]
CFR Part 50.65 (a)(4), "Requirementsfor Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" was identified
when Entergy did not perform a risk assessment for planned maintenance activities
when a tagout was applied on the 'B' electro-hydraulic control (EHC) pump, in
conjunction with a previous emergent failure of torus exhaust outer isolation valve
: [[27AOV]] [[-118. Description: On February 25, 2007, at 11:10 p.m., 27]]
AOV-118 would not open remotelyto vent the torus. 27AOV-118 is a primary containment isolation valve, and, as required
by TS 3.6.1.3, operators isolated the affected containment penetration flow path by
closing and deactivating the inboard isolation valve. Subsequently, a tagout was applied
on the 'B' EHC pump for scheduled maintenance. A risk assessment that considered
the emergent failure of 27AOV-118 and the impact on containment venting prior to
application of the tagout on the 'B' EHC pump was not performed. Specifically,
Entergy's risk assessment failed to consider risk significant systems, structures, andcomponents, as well as support systems that were unavailable during the maintenance.
Administrative Procedure (AP) 10.10, "On-Line Risk Assessment," assigns a risk
category color in risk significant order from Green, Yellow, Orange or Red based on
core damage frequency. Risk management actions are implemented depending on the
risk category color. With 27AOV-118 inoperable, combined with the impact of the
tagout on the EHC system, the color increased from Yellow to Orange. On February 26,
at approximately 6:46 a.m., the plant staff recognized the condition and the 'B' EHC
pump was returned to standby.Analysis:  The inspectors determined that the finding was a performance deficiencybecause Entergy did not perform a risk assessment following the emergent failure of
27AOV-118, and subsequently continued scheduled work activities including tagout of
8Enclosurethe 'B' EHC pump. The inspectors determined that this finding impacted the InitiatingEvents cornerstone due to the increased likelihood of a plant transient with the loss of
the normal heat sink. It was reasonable that Entergy should have identified the
condition and updated the risk assessment, because Entergy procedure AP 10.10, "On-
Line Risk Assessment," specifies that plant risk must be reassessed when plant
conditions change. Traditional enforcement does not apply since there were no actual
safety consequences or potential for impacting the NRC's regulatory function, and the
finding was not the result of any willful violation of NRC requirements or Entergy's
procedures. The inspectors determined that this finding was more than minor because it was similarto Example 7(f) in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor
Issues," in that, the emergent failure of 27AOV-118, in combination with the subsequent
removal of the 'B' electro hydraulic control pump availability resulted in the plant being in
a higher risk category, which required risk management actions, under Entergy's on-line
risk management procedure. The ability to vent the containment from the torus is a risk
important action to prevent containment failure and core damage for situations that
involve the inability to remove core decay heat from the torus water. The tagout of the
'B' electro hydraulic control pump increased the chance of a plant transient with the loss
of the normal heat sink.Using
: [[IMC]] [[0609, Appendix K, "Maintenance Risk Assessment and Risk Management]]
SDP," Flowchart 1, "Assessment of Risk Deficit," the inspectors determined the
incremental core damage probability deficit from Entergy's core damage frequency as a
result of the actual duration of the 27AOV-118 maintenance combined with the time the
'B' EHC system was not available due to the tagout (two hours). The inspectors
calculated the incremental core damage probability deficit to be in the high 10-8 range. Because the calculated risk deficit was not greater than 1 x 10-6 incremental coredamage probability deficit, the inspectors determined that this finding was of very low
safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance because Entergy did not incorporate appropriate risk insights into
planned work activities.Enforcement:  10 CFR 50.65 (a)(4), requires, in part, that before performingmaintenance activities (including but not limited to surveillance, post-maintenance
testing, and corrective and preventive maintenance), the licensee shall assess and
manage the increase in risk that may result from the proposed maintenance activities.
Contrary to the above, on February 26, 2007, Entergy did not assess and manage the
increase in risk for planned maintenance activities when a tagout was applied on the 'B'
: [[EHC]] [[pump, following an emergent failure of torus exhaust outer isolation valve 27]]
AOV-
118. Because this finding was of very low safety significance and was entered into
Entergy's corrective action program as condition report
: [[CR]] [[-]]
JAF-2007-00755, this
violation is being treated as an
: [[NCV]] [[, consistent with Section]]
VI.A.1 of the NRC
Enforcement Policy: NCV 05000333/2007002-02, Failure to Perform a Risk Assessment
When Required by 10 CFR 50.65(a)(4).
9Enclosure1R15Operability Evaluations (71111.15 - 6 samples) a.Inspection Scope The inspectors reviewed operability determinations to assess the acceptability of theevaluations; when needed, the use and control of compensatory measures; and
compliance with TS. The inspectors' review included a verification that the operability
determinations were made as specified by
: [[ENN]] [[-]]
OP-104, "Operability Determinations."
The technical adequacy of the determinations was reviewed and compared to the
: [[TS]] [[s,]]
: [[UFS]] [[]]
AR, and associated design basis documents. The documents reviewed are listed in
the Attachment. The following evaluations were reviewed and represented six
inspection samples:*CR-2007-00151 concerning a commercial grade gate timing circuit card installedin safety-related low pressure coolant injection inverter
: [[71INV]] [[-3A;*]]
: [[CR]] [[-2007-00293 concerning installation of nonsafety-related contacts insafety-related circuit breaker]]
: [[71MCC]] [[-163-]]
: [[OG]] [[-1 for containment isolation valve]]
: [[10MOV]] [[-34B; *]]
: [[CR]] [[-2007-00392 concerning an out of calibration condition of a local leak ratemonitor panel that affected the test results of 12 containment penetrations;*CR-2007-00647 concerning inaccurate valve position indication of vacuumbreaker]]
: [[27VB]] [[-2 while performing surveillance test]]
ST-15J, "Torus to Drywell
Vacuum Breakers Quarterly Test
: [[IST]] [[;"*]]
CR-2007-00281 concerning 'D' safety relief valve tailpipe temperature increase;and*CR-2007-00752 concerning extent of condition reviews following failure ofcontainment atmosphere control isolation valve 27AOV-118. b.FindingsNo findings of significance were identified.1R19Post Maintenance Testing (71111.19 - 6 samples) a.Inspection ScopeThe inspectors reviewed six post-maintenance test procedures and associated testingactivities for selected risk significant mitigating systems to assess whether the effect of
maintenance on plant systems was adequately addressed by control room and
engineering personnel. The inspectors verified:  test acceptance criteria were clear,
demonstrated operational readiness and were consistent with Design Basis Documents;
test instrumentation had current calibrations and adequate range and accuracy for the
application; and tests were performed, as written, with applicable prerequisites satisfied.
Upon completion, the inspectors verified that equipment was returned to the proper
alignment necessary to perform its safety function. Post-maintenance testing was
evaluated against the requirements of
: [[10 CFR]] [[50, Appendix B, Criterion]]
XI, "Test
10EnclosureControl."  The documents reviewed are listed in the Attachment. The following post-maintenance test activities were reviewed and represented six inspection samples:*Work request
: [[JAF]] [[-03-07188-01, involving repair of a water leak in service waterpump 46P-1B motor oil reservoir during the week of June 16, 2007;*Work request]]
JAF-05-17157, involving inspection and repair of a residual heatremoval service water keep-full check valve;*Work request JAF-05-22322, involving cleaning and replacement of two-inchemergency service water piping to the west electric bay and east cable tunnel
coolers;*Work request
: [[JAF]] [[-07-15650, involving repair of the air operator coupling of torusexhaust outboard containment isolation valve 27]]
: [[AOV]] [[-118;*Work request 51104288, involving replacement of residual heat removal servicewater pump 10P-1C due to high vibration; and*Work request  51178018, involving maintenance and repair of stator watercooling pump 94P-15B due to high vibration. b.FindingsNo findings of significance were identified.1R22Surveillance Testing (71111.22 - 7 samples) a.Inspection ScopeThe inspectors witnessed performance of surveillance tests and/or reviewed test data ofselected risk-significant]]
: [[SSC]] [[s to assess whether the]]
: [[SSC]] [[s satisfied]]
: [[TS]] [[,]]
UFSAR,
Technical Requirements Manual, and Entergy procedure requirements. The inspectors
verified:  test acceptance criteria were clear, demonstrated operational readiness, and
were consistent with Design Basis Documents; test instrumentation had current
calibrations and adequate range and accuracy for the application; and tests were
performed, as written, with applicable prerequisites satisfied. Upon surveillance test
(ST) completion, the inspectors verified that equipment was returned to the status
specified to perform its safety function. The inspectors evaluated the tests against the
requirements in
: [[TS.]] [[The following]]
STs were reviewed and represented seven
inspection program samples:*ST-2AM, "Residual Heat Removal Loop B Quarterly Operability Test;"*ST-4N, "High Pressure Coolant Injection Quick Start, Inservice, and TransientMonitoring Test;"  *MST-071.30, "Low Pressure Coolant Injection Charger-Inverter Performance andLow Pressure Coolant Injection Battery Service Surveillance Test;" *ST-7F, "Standby Gas Treatment Fan B and Valve Exercising Test IST;"
*ST-8Q, "Testing of the Emergency Service Water System;"
*ST-8Q, "Testing of the Emergency Service Water System;"
*ST-9BB, "Emergency Diesel Generator 'B' and 'D' Full Load Test and EmergencyService Water Pump Operability Test;" and*ST-5BB, "APRM System 'B' Channel Functional Test."
*ST-9BB, "Emergency Diesel Generator 'B' and 'D' Full Load Test and EmergencyService Water Pump Operability Test;" and*ST-5BB, "APRM System 'B' Channel Functional Test."
11Enclosure  b.Findings No findings of significance were identified.Cornerstone:  Emergency Preparedness1EP6Drill Evaluation (71114.06 - 1 sample)   a.Inspection Scope The inspectors observed emergency response organization activities during the fullparticipation drill that was conducted on January 11, 2007. The inspectors verified that
 
emergency classification declarations, notifications, and protective action
====b. Findings====
recommendations were properly completed. The inspectors evaluated the drill against
No findings of significance were identified.Cornerstone:  Emergency Preparedness1EP6Drill Evaluation (71114.06 - 1 sample)
the requirements of 10 CFR Part 50, Appendix E, "Emergency Planning and
 
Preparedness for Production and Utilization Facilities."  This observation constituted one
====a. Inspection Scope====
inspection program sample. b.FindingsNo findings of significance were identified.
The inspectors observed emergency response organization activities during the fullparticipation drill that was conducted on January 11, 2007. The inspectors verified that emergency classification declarations, notifications, and protective action recommendations were properly completed. The inspectors evaluated the drill against the requirements of 10 CFR Part 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities."  This observation constituted one inspection program sample.
: [[2.RADIA]] [[]]
 
: [[TION]] [[]]
====b. Findings====
: [[SAFETY]] [[Cornerstone: Occupational Radiation Safety 2]]
No findings of significance were identified.
OS2ALARA Planning and Controls (71121.02 - 4 samples) a.Inspection ScopeThe inspector conducted the following activities to verify that Entergy was properlymaintaining individual and collective radiation exposures as low as is reasonably
 
achievable (ALARA). Implementation of the
==RADIATION SAFETY==
: [[ALA]] [[]]
 
RA program was reviewed against the
===Cornerstone:===
criteria contained in 10 CFR 20.1101(b) and Entergy's procedures.(1)The following five highest exposure work activities from the Fall 2006 refuelingoutage were selected for review:*Reactor disassembly/reassembly;*In-Service inspection/ erosion-corrosion/intergranular stress corrosioncracking;*Reactor vessel visual inspection and defueling activities;
Occupational Radiation Safety 2OS2ALARA Planning and Controls (71121.02 - 4 samples)
 
====a. Inspection Scope====
The inspector conducted the following activities to verify that Entergy was properlymaintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). Implementation of the ALARA program was reviewed against the criteria contained in 10 CFR 20.1101(b) and Entergy's procedures.(1)The following five highest exposure work activities from the Fall 2006 refuelingoutage were selected for review:*Reactor disassembly/reassembly;*In-Service inspection/ erosion-corrosion/intergranular stress corrosioncracking;*Reactor vessel visual inspection and defueling activities;
*Control rod drive replacement; and
*Control rod drive replacement; and
*Preventive maintenance on motor-operated valves.
*Preventive maintenance on motor-operated valves.
2Enclosure(2)With respect to the work activities listed above, the
 
: [[ALA]] [[]]
12Enclosure(2)With respect to the work activities listed above, the ALARA evaluations,exposure estimates, and applicable exposure mitigation requirements were reviewed. This included a review of exposure mitigation procedures and engineering and work controls to achieve exposures that are ALARA. These work activities were also reviewed to determine if they were reasonably grouped into work activities, based on historical precedence or industry standard groupings.(3)The actual results achieved were compared with the intended dose establishedin the ALARA planning for the work activities. ALARA post-job reviews were reviewed and interviews were conducted to evaluate the adequacy of ALARA controls as implemented, and to identify any significant performance deficiencies that may have resulted in unintended dose consequences.(4)The methodology for adjusting work activity exposure estimates were evaluatedwith respect to the work activities. The reasons for the exposure estimate adjustments were determined and evaluated with respect to sound radiation protection and ALARA principles and to ensure the revised exposure estimates provided an effective ALARA performance measure.
RA evaluations,exposure estimates, and applicable exposure mitigation requirements were
 
reviewed. This included a review of exposure mitigation procedures and
====b. Findings====
engineering and work controls to achieve exposures that are
No findings of significance were identified.4.OTHER ACTIVITIES4OA2Identification and Resolution of Problems.1Annual Sample: Operator Workaround Program (71152 - 1 sample)
: [[ALA]] [[]]
 
RA. These
====a. Inspection Scope====
work activities were also reviewed to determine if they were reasonably grouped
The inspectors reviewed the cumulative effects of operator workarounds on thereliability, availability, and potential for mis-operation of a system and on the operator's ability to implement abnormal or emergency operating procedures. The inspectors reviewed the results of Entergy Surveillance Test ST-99H, "Operations Cumulative Impact Assessment," and the resolution of items identified in the assessment. The inspectors reviewed Entergy's program for identifying operator workarounds at an appropriate threshold and for entering them into the corrective action program. In addition, inspectors reviewed operation department records including standing orders for operational decision-making issues and operability evaluations.
into work activities, based on historical precedence or industry standard
 
groupings.(3)The actual results achieved were compared with the intended dose establishedin the
b.Assessment and ObservationsNo findings of significance were identified. The corrective action program waseffectively used to identify and resolve operator workarounds. The resolution of operator workaround items has been appropriately prioritized.
: [[ALARA]] [[planning for the work activities.]]
 
ALARA post-job reviews were
13Enclosure.2Review of Items Entered into the Corrective Action Program
reviewed and interviews were conducted to evaluate the adequacy of
 
: [[ALA]] [[]]
====a. Inspection Scope====
RA
As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's corrective action program. The review was accomplished by accessing Entergy's computerized database for CRs and attending CR screening meetings.In accordance with the baseline inspection modules, the inspectors selected correctiveaction program items across the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for additional follow-up and review. Additionally, an NRC specialist inspector reviewed six CRs associated with the Occupational Radiation Safety Cornerstone that were initiated between October 2006 and January 2007. The inspectors assessed Entergy's threshold for problem identification, the adequacy of the cause analyses, extent of condition review, operability determinations, and the timeliness of the specified corrective actions. The CRs reviewed are listed in the
controls as implemented, and to identify any significant performance deficiencies
. b.Assessment and Observations  No findings of significance were identified.
that may have resulted in unintended dose consequences.(4)The methodology for adjusting work activity exposure estimates were evaluatedwith respect to the work activities. The reasons for the exposure estimate
 
adjustments were determined and evaluated with respect to sound radiation
4OA3Event Followup (71153 - 1 sample).1(Closed) LER 05000333/2006002-00, High Pressure Coolant Injection System DeclaredInoperable Due to Turbine Speed Oscillations.On November 4, 2006, with the plant operating in Mode 1, Entergy identified that thehigh pressure coolant injection system was inoperable due to turbine speed oscillations.
protection and
 
: [[ALA]] [[]]
The condition was discovered during post-maintenance testing following a refueling outage, and was caused by connecting two turbine hydraulic actuator oil lines to the incorrect oil ports. The enforcement aspects of this violation of maintenance procedures were documented in section
RA principles and to ensure the revised exposure estimates
{{a|4OA7}}
provided an effective
==4OA7 of NRC Inspection Report 05000333/2006005.==
: [[ALARA]] [[performance measure. b.FindingsNo findings of significance were identified.4.]]
 
: [[OTHER]] [[]]
Entergy entered the event into its corrective action program as CR-2006-04754. This LER is closed. 4OA5Other Activities.1Closed: URI 2006006-03, "Simulator Transient Testing"This URI was opened because the inspectors determined that the facility process forreviewing simulator transient testing was not adequate to comply with 10 CFR Part 55.46(d)(1) and Regulatory Guide 1.149, "Nuclear Power Plant Simulation Facilities for use in Operator Training and License Examinations," Revision 1. The inspectors 14Enclosureconsidered this issue unresolved pending an appropriate review of test results todetermine if significant discrepancies existed in simulator performance.Entergy put together an "expert panel" of six individuals from operations, engineering,and training. This use of an expert panel satisfies regulatory requirements for test review. These individuals reviewed the most recent simulator transient tests and identified discrepancies. Two discrepancies were identified by the expert panel review which resulted in simulatormodel changes. One instance involved solid plant pressure response at high pressure, the other involved indications of recirculation flow under certain conditions with allrecirculation pumps tripped.The inspectors evaluated the impact of these discrepancies against the criteriadescribed in NRC Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance significance determination process," and determined that these discrepancies were minor. Accordingly, this URI is closed.4OA6Meetings, Including ExitOn March 27, 2007, the inspectors presented the inspection results toMr. Peter T. Dietrich and other members of his staff. The inspectors asked Entergy whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.ATTACHMENT:  
: [[ACTIVI]] [[]]
 
TIES4OA2Identification and Resolution of Problems.1Annual Sample: Operator Workaround Program (71152 - 1 sample) a.Inspection ScopeThe inspectors reviewed the cumulative effects of operator workarounds on thereliability, availability, and potential for mis-operation of a system and on the operator's
=SUPPLEMENTAL INFORMATION=
ability to implement abnormal or emergency operating procedures. The inspectors
 
reviewed the results of Entergy Surveillance Test ST-99H, "Operations Cumulative
==KEY POINTS OF CONTACT==
Impact Assessment," and the resolution of items identified in the assessment. The
Entergy Personnel
inspectors reviewed Entergy's program for identifying operator workarounds at an
: [[contact::P. Dietrich]], Site Vice President
appropriate threshold and for entering them into the corrective action program. In
: [[contact::C. Adner]], Manager, Operations
addition, inspectors reviewed operation department records including standing orders
: [[contact::N. Avrakotos]], Manager, Emergency Preparedness  
for operational decision-making issues and operability evaluations. b.Assessment and ObservationsNo findings of significance were identified. The corrective action program waseffectively used to identify and resolve operator workarounds. The resolution of
: [[contact::S. Bono]], Director Engineering
operator workaround items has been appropriately prioritized.
: [[contact::J. Costedio]], Manager, Regulatory Compliance
13Enclosure.2Review of Items Entered into the Corrective Action Program   a.Inspection ScopeAs required by Inspection Procedure 71152, "Identification and Resolution of Problems,"and in order to help identify repetitive equipment failures or specific human performance
: [[contact::M. Durr]], Manager, System Engineering
issues for follow-up, the inspectors performed a daily screening of all items entered into
: [[contact::B. Finn]], Director, Nuclear Safety Assurance
Entergy's corrective action program. The review was accomplished by accessing
: [[contact::D. Johnson]], Manager, Training
Entergy's computerized database for
: [[contact::J. LaPlante]], Manager, Security
: [[CR]] [[s and attending]]
: [[contact::K. Mulligan]], General Manager, Plant Operations
CR screening meetings.In accordance with the baseline inspection modules, the inspectors selected correctiveaction program items across the Initiating Events, Mitigating Systems, and Barrier
: [[contact::J. Pechacek]], Manager, Programs and Components Engineering
Integrity cornerstones for additional follow-up and review. Additionally, an NRC
: [[contact::W. Rheaume]], Manager, CA&A
specialist inspector reviewed six CRs associated with the Occupational Radiation Safety
: [[contact::J. Solowski]], Radiation Protection
Cornerstone that were initiated between October 2006 and January 2007. The
inspectors assessed Entergy's threshold for problem identification, the adequacy of the
cause analyses, extent of condition review, operability determinations, and the
timeliness of the specified corrective actions. The CRs reviewed are listed in the
Attachment. b.Assessment and Observations  No findings of significance were identified.
: [[4OA]] [[3Event Followup (71153 - 1 sample).1(Closed)]]
LER 05000333/2006002-00, High Pressure Coolant Injection System DeclaredInoperable Due to Turbine Speed Oscillations.On November 4, 2006, with the plant operating in Mode 1, Entergy identified that thehigh pressure coolant injection system was inoperable due to turbine speed oscillations.
The condition was discovered during post-maintenance testing following a refueling
outage, and was caused by connecting two turbine hydraulic actuator oil lines to the
incorrect oil ports. The enforcement aspects of this violation of maintenance procedures
were documented in section
: [[4OA]] [[7 of]]
NRC Inspection Report 05000333/2006005.
Entergy entered the event into its corrective action program as
: [[CR]] [[-2006-04754. This]]
: [[LER]] [[is closed. 4]]
: [[OA]] [[5Other Activities.1Closed:]]
: [[URI]] [[2006006-03, "Simulator Transient Testing"This]]
URI was opened because the inspectors determined that the facility process forreviewing simulator transient testing was not adequate to comply with 10 CFR Part
55.46(d)(1) and Regulatory Guide 1.149, "Nuclear Power Plant Simulation Facilities for
use in Operator Training and License Examinations," Revision 1. The inspectors
14Enclosureconsidered this issue unresolved pending an appropriate review of test results todetermine if significant discrepancies existed in simulator performance.Entergy put together an "expert panel" of six individuals from operations, engineering,and training. This use of an expert panel satisfies regulatory requirements for test
review. These individuals reviewed the most recent simulator transient tests and
identified discrepancies. Two discrepancies were identified by the expert panel review which resulted in simulatormodel changes. One instance involved solid plant pressure response at high pressure,
the other involved indications of recirculation flow under certain conditions with allrecirculation pumps tripped.The inspectors evaluated the impact of these discrepancies against the criteriadescribed in NRC Inspection Manual Chapter 0609, Appendix I, "Operator
Requalification Human Performance significance determination process," and
determined that these discrepancies were minor. Accordingly, this
: [[URI]] [[is closed.4]]
OA6Meetings, Including ExitOn March 27, 2007, the inspectors presented the inspection results toMr. Peter T. Dietrich and other members of his staff. The inspectors asked Entergy
whether any of the material examined during the inspection should be considered
proprietary. No proprietary information was identified.ATTACHMENT:
: [[SUPPLE]] [[]]
: [[MENTAL]] [[]]
: [[INFORM]] [[]]
: [[ATION]] [[A-1AttachmentSUPPLEMENTAL]]
: [[INFORM]] [[]]
: [[ATIONK]] [[EY]]
: [[POINTS]] [[]]
: [[OF]] [[]]
: [[CONTAC]] [[]]
TEntergy PersonnelP. Dietrich, Site Vice PresidentC. Adner, Manager, Operations
N. Avrakotos, Manager, Emergency Preparedness
S. Bono, Director Engineering
J. Costedio, Manager, Regulatory Compliance
M. Durr, Manager, System Engineering
B. Finn, Director, Nuclear Safety Assurance
D. Johnson, Manager, Training
J. LaPlante, Manager, Security
K. Mulligan, General Manager, Plant Operations
: [[J.]] [[Pechacek, Manager, Programs and Components Engineering]]
: [[W.]] [[Rheaume, Manager,]]
CA&A
J. Solowski, Radiation Protection
John Pircsuk, Senior Operations Instructor
John Pircsuk, Senior Operations Instructor
Darren Deritz, Regulatory Compliance
Darren Deritz, Regulatory Compliance
Greg Pitts, TrainingLIST
Greg Pitts, Training
: [[OF]] [[]]
==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
: [[ITEMS]] [[]]
Opened and
: [[OPENED]] [[,]]
===Closed===
: [[CLOSED]] [[,]]
: [[Closes LER::05000333/LER-2006-002]]-00LERHigh Pressure Coolant Injection SystemDeclared Inoperable Due to Turbine Speed
: [[AND]] [[]]
: Oscillations (Section 4OA3)
DISCUSSEDOpened and Closed05000333/2006002-00LERHigh Pressure Coolant Injection SystemDeclared Inoperable Due to Turbine Speed
: [[Closes finding::05000333/FIN-2007002-01]]NCVInadequate Maintenance on ContainmentAtmosphere Control Valve (Section 1R13)
Oscillations (Section
: [[Closes finding::05000333/FIN-2007002-02]]NCVFailure to Perform a Risk AssessmentWhen Required by 10 CFR 50.65(a)(4)
: [[4OA]] [[3)05000333/2007002-01]]
(Section 1R13)
NCVInadequate Maintenance on ContainmentAtmosphere Control Valve (Section 1R13)05000333/2007002-02NCVFailure to Perform a Risk AssessmentWhen Required by 10 CFR 50.65(a)(4)
 
(Section 1R13)Closed05000333/2006003-03URISimulator Transient Testing (Section
===Closed===
: [[4OA]] [[5)]]
: [[Closes LER::05000333/LER-2006-002]]-00LERHigh Pressure Coolant Injection SystemDeclared Inoperable Due to Turbine Speed
: [[LIST]] [[]]
: Oscillations (Section 4OA3)
: [[OF]] [[]]
: [[Closes finding::05000333/FIN-2007002-01]]NCVInadequate Maintenance on ContainmentAtmosphere Control Valve (Section 1R13)
: [[DOCUME]] [[NTS]]
: [[Closes finding::05000333/FIN-2007002-02]]NCVFailure to Perform a Risk AssessmentWhen Required by 10 CFR 50.65(a)(4)
: [[REVIEW]] [[]]
(Section 1R13)
EDSection 1R04: Equipment AlignmentOP-43C, "Low Pressure Coolant Injection Independent Power Supply System," Revision 17MST-071.11, "Low Pressure Coolant Injection Battery Quarterly Surveillance Test," Revision 19
 
A-2AttachmentDBD-027, "Design Basis Document for The Air Treatment Systems," Revision 10OP-37, "Containment Atmosphere Dilution System," Revision 73
==LIST OF DOCUMENTS REVIEWED==
FM-18A, "Flow Diagram - Drywell Inerting C.A.D. and Purge," Revision 57
==Section 1R04: Equipment AlignmentOP-43C, "Low Pressure Coolant Injection Independent Power Supply System," Revision 17MST-071.11, "Low Pressure Coolant Injection Battery Quarterly Surveillance Test," Revision 19==
FM-18B, "Flow Diagram - Drywell Inerting C.A.D. Purge and Containment Differential
: A-2AttachmentDBD-027, "Design Basis Document for The Air Treatment Systems," Revision 10OP-37, "Containment Atmosphere Dilution System," Revision 73
Pressurization," Revision 37
: FM-18A, "Flow Diagram - Drywell Inerting C.A.D. and Purge," Revision 57
14620.9011-US(N)-001, "Torque Calculation for Suppression Chamber (20") and Drywell (24")
: FM-18B, "Flow Diagram - Drywell Inerting C.A.D. Purge and Containment Differential Pressurization," Revision 37
Vent and Purge Butterfly Valves," Revision 2
: 14620.9011-US(N)-001, "Torque Calculation for Suppression Chamber (20") and Drywell (24")
: [[JAF]] [[-]]
: Vent and Purge Butterfly Valves," Revision 2
CALC-CAD-02782, "Nitrogen Flows and Volumes for Normal and Accident Conditions,"
: JAF-CALC-CAD-02782, "Nitrogen Flows and Volumes for Normal and Accident Conditions," Revision 0
Revision 0Section 1R05: Fire ProtectionPFP-PWR42, Fire Area/Zone 1E/TB-1PFP-PWR43, Fire Area/Zone 1E/TB-1
 
: [[PFP]] [[-]]
==Section 1R05: Fire ProtectionPFP-PWR42, Fire Area/Zone 1E/TB-1PFP-PWR43, Fire Area/Zone 1E/TB-1==
: [[PWR]] [[45, Fire Area/Zone 1E/TB-1]]
: PFP-PWR45, Fire Area/Zone 1E/TB-1
: [[PFP]] [[-]]
: PFP-PWR22, Fire Area/Zone IX/SG-1
: [[PWR]] [[22, Fire Area/Zone]]
: PFP-PWR20, Fire Area/Zone IX/RB-1A
: [[IX]] [[/]]
: PFP-PWR29, Fire Area/Zone II/SW-1
: [[SG]] [[-1]]
: PFP-PWR30, Fire Area/Zone IC/SW-1
: [[PFP]] [[-]]
: PFP-PWR31, Fire Area/Zone V/EG-1
: [[PWR]] [[20, Fire Area/Zone]]
: PFP-PWR11, Fire Area/Zone VII/CS-1
: [[IX]] [[/]]
: ST-76J13, "Heat Detector Functional Test - SGT Filter B," Revision 16
: [[RB]] [[-1A]]
 
: [[PFP]] [[-]]
==Section 1R06: Flood Protection MeasuresAOP-43, "Plant Shutdown from Outside the Control Room," Revision 32AOP-51, "Unexpected Fire Pump Start," Revision 5==
: [[PWR]] [[29, Fire Area/Zone]]
: AP-16.14, "Hazard Barrier Controls," Revision 3
: [[II]] [[/]]
: 11825-FE-1AS, "120V AC One Line Diagram, Emergency Bus A2 & B2 Distribution Panels
: [[SW]] [[-1]]
: 71ACA2 & 71ACB2," Revision 18
: [[PFP]] [[-]]
: LO-OEN-2005-00193, "Response to NRC Information Notice 2005-011, 'Internal Flooding/Spray-Down of Safety-Related Equipment Due to Unsealed Equipment Hatch Floor Plugs and/or Blocked Floor Drains'"
: [[PWR]] [[30, Fire Area/Zone]]
: LO-OEN-2005-00555, "Response to NRC Information Notice 2005-030, 'Safe Shutdown Potentially Challenged by Unanalyzed Internal Flooding Events and Inadequate Design'"
: [[IC]] [[/]]
: 0090-0006-C-003, "JAF NPP Fire Suppression Effects Analysis," Revision 0
: [[SW]] [[-1]]
: DBD-071 Tab 1, "Design Basis Document for the Electrical Distribution Systems 4160V and
: [[PFP]] [[-]]
: 600V AC Power Systems," Revision 2
: [[PWR]] [[31, Fire Area/Zone V/EG-1]]
: JAF-RPT-MULTI-02107, "IPE Update, Appendix H, Internal Flooding Analysis," Revision 2
: [[PFP]] [[-]]
 
: [[PWR]] [[11, Fire Area/Zone]]
==Section 1RO7: Heat Sink PerformanceAOP-30, "Loss of Shutdown Cooling," Revision 19AP-19.14, "Eddy Current Testing of Heat Exchanger Tubes," Revision 8==
: [[VII]] [[/]]
: EOP-4, "Primary Containment Control," Revision 7
: [[CS]] [[-1]]
: OP-13, "Residual Heat Removal System," Revision 93  
: [[ST]] [[-76J13, "Heat Detector Functional Test -]]
: A-3AttachmentST-2Y, "Residual Heat Removal Heat Exchanger Performance Test," Revision 7, datedJanuary 24, 2007
SGT Filter B," Revision 16Section 1R06: Flood Protection MeasuresAOP-43, "Plant Shutdown from Outside the Control Room," Revision 32AOP-51, "Unexpected Fire Pump Start," Revision 5
: JAF-CALC-MISC-02634, "Torus Station Blackout Torus Heat-Up for Power Uprate Conditions," Revision 0
AP-16.14, "Hazard Barrier Controls," Revision 3
: JAF-CALC-Residual Heat Removal-01903, "Instrument Indication Uncertainty for Residual HeatRemoval Heat Exchanger Performance Test," Revision 1
11825-FE-1AS, "120V
: JAF-CALC-Residual Heat Removal-02953, "Residual Heat Removal Heat Exchanger K-value with Reduced Tube Side Fouling Factor," Revision 0
: [[AC]] [[One Line Diagram, Emergency Bus A2 & B2 Distribution Panels]]
: JAF-CALC-Residual Heat Removal-00392, "Calculation for Design Basis/Acceptance Criteriafor
: [[71ACA]] [[2 & 71]]
: ST-2Y," Revision 0
: [[ACB]] [[2," Revision 18]]
: GE-NE-T23-00373-01, "Higher Residual Heat Removal Service Water Temperature Analysis," Revision 1
: [[LO]] [[-]]
: GE-NE-T23-00766-00-01, "Containment Analysis to Support ECCS Pump NPSH Evaluation," Revision 0
OEN-2005-00193, "Response to NRC Information Notice 2005-011, 'Internal
: JPN-91-015, "Updated Response to Generic Letter 89-13 Service Water System Problems Affecting Safety-Related Equipment," dated April 18, 1991
Flooding/Spray-Down of Safety-Related Equipment Due to Unsealed Equipment Hatch Floor
: JPN-93-015, "Updated Response to Generic Letter 89-13 Service Water System Problems Affecting Safety-Related Equipment," dated March 16, 1993
Plugs and/or Blocked Floor Drains'"
: [[LO]] [[-]]
OEN-2005-00555, "Response to NRC Information Notice 2005-030, 'Safe Shutdown
Potentially Challenged by Unanalyzed Internal Flooding Events and Inadequate Design'"
0090-0006-C-003, "JAF NPP Fire Suppression Effects Analysis," Revision 0
DBD-071 Tab 1, "Design Basis Document for the Electrical Distribution Systems 4160V and
600V
: [[AC]] [[Power Systems," Revision 2]]
: [[JAF]] [[-]]
: [[RPT]] [[-MULTI-02107, "IPE Update, Appendix H, Internal Flooding Analysis," Revision 2Section]]
: [[1RO]] [[7: Heat Sink Performance]]
AOP-30, "Loss of Shutdown Cooling," Revision 19AP-19.14, "Eddy Current Testing of Heat Exchanger Tubes," Revision 8
EOP-4, "Primary Containment Control," Revision 7
OP-13, "Residual Heat Removal System," Revision 93
A-3AttachmentST-2Y, "Residual Heat Removal Heat Exchanger Performance Test," Revision 7, datedJanuary 24, 2007
: [[JAF]] [[-]]
CALC-MISC-02634, "Torus Station Blackout Torus Heat-Up for Power Uprate Conditions,"
Revision 0
: [[JAF]] [[-]]
: [[CALC]] [[-Residual Heat Removal-01903, "Instrument Indication Uncertainty for Residual HeatRemoval Heat Exchanger Performance Test," Revision 1]]
: [[JAF]] [[-]]
CALC-Residual Heat Removal-02953, "Residual Heat Removal Heat Exchanger K-value
with Reduced Tube Side Fouling Factor," Revision 0
: [[JAF]] [[-]]
: [[CALC]] [[-Residual Heat Removal-00392, "Calculation for Design Basis/Acceptance Criteriafor ST-2Y," Revision 0]]
: [[GE]] [[-]]
NE-T23-00373-01, "Higher Residual Heat Removal Service Water Temperature Analysis,"
Revision 1
: [[GE]] [[-]]
: [[NE]] [[-T23-00766-00-01, "Containment Analysis to Support]]
: [[ECCS]] [[Pump]]
NPSH Evaluation,"
Revision 0
JPN-91-015, "Updated Response to Generic Letter 89-13 Service Water System Problems
Affecting Safety-Related Equipment," dated April 18, 1991
JPN-93-015, "Updated Response to Generic Letter 89-13 Service Water System Problems
Affecting Safety-Related Equipment," dated March 16, 1993
"Eddy Current Inspection for Residual Heat Removal B Heat Exchanger," dated April 17, 2003
"Eddy Current Inspection for Residual Heat Removal B Heat Exchanger," dated April 17, 2003
Master Lee
: Master Lee NDE Services, "Inspection Summary," dated April 2003
: [[NDE]] [[Services, "Inspection Summary," dated April 2003Section 1R11: Licensed Operator Requalification Program71775-0, Loss of B Reactor Protection System Bus, failure to Automatically Scram, with aScram Discharge Isolation Valve Rupture After the Manual ScramSection 1R12: Maintenance Effectiveness]]
 
: [[JAF]] [[-RPT-ELEC-02302, "MR Basis Document for System 071 DC Electrical Distribution,Revision 3]]
==Section 1R11: Licensed Operator Requalification Program71775-0, Loss of B Reactor Protection System Bus, failure to Automatically Scram, with aScram Discharge Isolation Valve Rupture After the Manual ScramSection 1R12: Maintenance EffectivenessJAF-RPT-ELEC-02302, "MR Basis Document for System 071==
: [[JAF]] [[-]]
: DC Electrical Distribution,Revision 3
RPT-MULTI-02294, "MR Basis Document for Service Water Systems Including System
: JAF-RPT-MULTI-02294, "MR Basis Document for Service Water Systems Including System
010 - Residual Heat Removal Service Water - System 046 -- Emergency Service Water -
: 010 - Residual Heat Removal Service Water - System 046 -- Emergency Service Water -
System 046-000 - Normal Service Water," Revision 6Section 1R15: Operability EvaluationsST-39B, "Type B and C
: System 046-000 - Normal Service Water," Revision 6
: [[LL]] [[]]
 
RT of Containment Penetrations," Revision 31
==Section 1R15: Operability EvaluationsST-39B, "Type B and C==
Section 1R19: Post Maintenance TestingJAF-RPT-MULTI-02294, "MR Basis Document for Service Water Systems," Revision
: LLRT of Containment Penetrations," Revision 31
: [[6AP]] [[-05.07, "Maintenance Testing and Post-Work Testing (]]
 
ISI)," Revision 36
==Section 1R19: Post Maintenance TestingJAF-RPT-MULTI-02294, "MR Basis Document for Service Water Systems," Revision 6AP-05.07, "Maintenance Testing and Post-Work Testing (ISI)," Revision 36==
: [[MP]] [[-059.45, "Piston Check Valves," Revision 10]]
: MP-059.45, "Piston Check Valves," Revision 10
: [[JAF]] [[-]]
: JAF-RPT-MULTI-00406, "Inservice Test Program Basis Document," Revision 0
RPT-MULTI-00406, "Inservice Test Program Basis Document," Revision 0Section 1R22: Surveillance Testing
 
A-4AttachmentIEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, andReplacement of Vented Lead-Acid Batteries for Stationary Applications"
==Section 1R22: Surveillance Testing==
MP-057-.06, "Battery Maintenance," Revision 27
: A-4AttachmentIEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, andReplacement of Vented Lead-Acid Batteries for Stationary Applications"
JAF-89-026, "Establish Acceptance Criteria for Low Pressure Coolant Injection Battery Intercell
: MP-057-.06, "Battery Maintenance," Revision 27
Connector Resistances," Revision 0
: JAF-89-026, "Establish Acceptance Criteria for Low Pressure Coolant Injection Battery Intercell Connector Resistances," Revision 0
: [[JAF]] [[-]]
: JAF-CALC-ELEC-00562, "Low Pressure Coolant Injection Battery Testing Duty Cycle," Revision
: [[CALC]] [[-ELEC-00562, "Low Pressure Coolant Injection Battery Testing Duty Cycle," Revision]]
: JAF-CALC-ELEC-01857, "419 V DC Low Pressure Coolant Injection Power System 3A & 3B
: [[JAF]] [[-]]
: Sizing," Revision 0
CALC-ELEC-01857, "419 V DC Low Pressure Coolant Injection Power System 3A & 3B
: JAF-CALC-ELEC-02213, "Low Pressure Coolant Injection UPS System Testing Load Bank Characteristics and Low Pressure Coolant Injection Battery and Inverter on Line Testing Conditions and/or Limitation," Revision 0
Sizing," Revision 0
: JAF-CAL-ELEC-00523, "71/Low Pressure Coolant Injection UPS System Testing Duty Cycle," Revision 3 and Margin Revisions 3A and 0A
: [[JAF]] [[-]]
: JAF Manual No. E356-0048, "Stationary Lead-Acid Batteries"
CALC-ELEC-02213, "Low Pressure Coolant Injection UPS System Testing Load Bank
: JAF-CALC-NMS-00758, "Setpoint Calculation for APRM A through F," Revision 11
Characteristics and Low Pressure Coolant Injection Battery and Inverter on Line Testing
: JTS-93-0877, "Surveillance Test Adequacy Review of APRMs and Control Rod Blocks Findings," Revision 1Section 4OA2: Identification and Resolution of Problems Condition Reports2003-021042003-02269
Conditions and/or Limitation," Revision 0
: 2000-06351
: [[JAF]] [[-]]
: 2003-01787
CAL-ELEC-00523, "71/Low Pressure Coolant Injection UPS System Testing Duty Cycle,"
: 2006-04276
Revision 3 and Margin Revisions 3A and 0A
: 2006-05039
: [[JAF]] [[Manual No. E356-0048, "Stationary Lead-Acid Batteries"]]
: 2006-04873
: [[JAF]] [[-]]
: 2006-05108
: [[CALC]] [[-NMS-00758, "Setpoint Calculation for]]
: 2006-05021
: [[AP]] [[]]
: 2006-04965
: [[RM]] [[A through F," Revision 11]]
: 2006-04645
: [[JTS]] [[-93-0877, "Surveillance Test Adequacy Review of]]
: 2006-05106
APRMs and Control Rod Blocks
: 2007-00151
Findings," Revision 1Section 4OA2: Identification and Resolution of Problems Condition Reports2003-021042003-02269
: 2007-00127
2000-06351
: 2007-003762007-012362007-01019
2003-01787
: 2007-01128
2006-04276
: 2007-00914
2006-05039
: 2007-00884
2006-04873
: 2007-00909
2006-05108
: 2007-00862
2006-05021
: 2007-00809
2006-04965
: 2007-00804
2006-04645
: 2007-00824
2006-05106
: 2007-00647
2007-00151
: 2007-00641
2007-00127
: 2007-00629
2007-003762007-012362007-01019
: 2007-00628
2007-01128
: 2007-006252007-004972007-00531
2007-00914
: 2007-00454
2007-00884
: 2007-00392
2007-00909
: 2007-01045
2007-00862
: 2007-01043
2007-00809
: 2007-00592
2007-00804
: 2007-00719
2007-00824
: 2007-00983
2007-00647
: 2007-00845
2007-00641
: 2007-00241
2007-00629
: 2007-00104
2007-00628
: 2007-00740
2007-006252007-004972007-00531
: 2007-00741
2007-00454
: 2007-01196
2007-00392
 
2007-01045
==Section 4OA5: Other ActivitiesTP-7.03, "Simulator Test Program," Revision 1CR-JAF-2006-02057==
2007-01043
: SDR-9494, ANSI Transient Questions
2007-00592
: SDR-9297, ANSI06 Max Size Unisolable Steam Rupture  
2007-00719
: A-5
2007-00983
==LIST OF ACRONYMS==
2007-00845
ADAMSAgency Documents Access Management SystemALARAas low as is reasonably achievable
2007-00241
2007-00104
2007-00740
2007-00741
2007-01196 Section
: [[4OA]] [[5: Other Activities]]
: [[TP]] [[-7.03, "Simulator Test Program," Revision]]
: [[1CR]] [[-]]
: [[JAF]] [[-2006-02057]]
: [[SDR]] [[-9494,]]
: [[ANSI]] [[Transient Questions]]
: [[SDR]] [[-9297,]]
ANSI06 Max Size Unisolable Steam Rupture
A-5LIST
: [[OF]] [[]]
ACRONYMSADAMSAgency Documents Access Management SystemALARAas low as is reasonably achievable
APadministrative procedure
APadministrative procedure
: [[CDF]] [[core damage frequency]]
CDFcore damage frequency
: [[CF]] [[]]
CFRCode of Federal Regulations
RCode of Federal Regulations
CRcondition report
CRcondition report
DBDdesign basis document
DBDdesign basis document
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IMCinspection manual chapter
IMCinspection manual chapter
ISIinservice inspection
ISIinservice inspection
: [[IST]] [[inservice testing]]
ISTinservice testing
: [[LP]] [[]]
LPCIlow pressure coolant injection
CIlow pressure coolant injection
MPmaintenance procedure
MPmaintenance procedure
: [[NCV]] [[non-cited violation]]
NCVnon-cited violation
: [[NR]] [[]]
NRCNuclear Regulatory Commission
CNuclear Regulatory Commission
OPoperating procedure
: [[OP]] [[operating procedure]]
PARSpublicly available records
: [[PA]] [[]]
RSpublicly available records
psigpounds per square inch gauge
psigpounds per square inch gauge
SDPsignificance determination process
SDPsignificance determination process
SSCstructures, systems or components
SSCstructures, systems or components
STsurveillance test
STsurveillance test
: [[TS]] [[technical specification]]
TStechnical specification
: [[UFSA]] [[]]
UFSARUpdated Final Safety Evaluation Report
RUpdated Final Safety Evaluation Report
: [[URI]] [[unresolved item]]
: [[URI]] [[unresolved item]]
}}
}}

Revision as of 21:45, 23 October 2018

IR 05000333-07-002, on 01/01/2007 - 03/31/2007, Entergy Nuclear Northeast (Entergy); James A. FitzPatrick Nuclear Power Plant; Maintenance Risk Assessment and Emergent Work Control - NRC Integrated Inspection Report
ML071340409
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 05/11/2007
From: Cobey E W
Reactor Projects Branch 2
To: Peter Dietrich
Entergy Nuclear Northeast
Cobey, Eugene W. RI/DRP/PB2/610-337-5171
References
IR-07-002
Download: ML071340409 (26)
v • d • e


Text

May 11, 2007

Mr. Peter T. DietrichSite Vice President Entergy Nuclear Northeast James A. FitzPatrick Nuclear Power PlantPost Office Box 110 Lycoming, NY 13093

SUBJECT: JAMES A. FITZPATRICK NUCLEAR POWER PLANT - NRC INTEGRATEDINSPECTION REPORT 05000333/2007002

Dear Mr. Dietrich:

On March 31, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspectionat your James A. FitzPatrick Nuclear Power Plant. The enclosed integrated inspection report documents the inspection results, which were discussed on March 27, 2007, with you and other members of your staff.The inspection examined activities conducted under your license as they relate to safety andcompliance with the Commission's 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 findings of very low safety significance (Green) wereidentified. These findings were also determined to be violations of NRC requirements.

However, because of their very low safety significance, and because they were entered into your corrective action program, the NRC is treating these findings as non-cited violations (NCVs) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest any NCV in this report, you should provide a written response within 30 days of the date of this inspection report with the basis or your denial, to the 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; and the NRC Senior Resident Inspector at the James A. FitzPatrick Nuclear Power Plant.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in the P. Dietrich2NRC Public Document Room or from the Publicly Available Records (PARS) component of theNRC's document system (ADAMS). ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/Eugene W. Cobey, ChiefProjects Branch 2 Division of Reactor ProjectsDocket No.:50-333License No.: DPR-59

Enclosure:

Inspection Report 05000333/2007002

w/Attachment:

Supplemental Informationcc w/encl:G. J. Taylor, Chief Executive Officer, Entergy Operations M. R. Kansler, President, Entergy Nuclear Operations, Inc.

J. T. Herron, Senior Vice President for Operations M. Balduzzi, Senior Vice President, Northeastern Regional Operations W. Campbell, Senior Vice President of Engineering and Technical Services C. Schwarz, Vice President, Operations Support K. Mulligan, General Manager, Plant Operations O. Limpias, Vice President, Engineering (ENO)

J. McCann, Director, Licensing (ENO)

C. Faison, Manager, Licensing (ENO)

M. Colomb, Director of Oversight (ENO)

W. Rheaume, Director, Nuclear Safety Assurance J. Costedio, Manager, Regulatory Compliance T. McCullough, Assistant General Counsel (ENO)

P. Smith, President, New York State Energy Research and Development Authority P. Eddy, New York State Department of Public Service S. Lyman, Oswego County Administrator Supervisor, Town of Scriba C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law J. Sniezek, PWR SRC Consultant M. Lyster, PWR SRC Consultant S. Lousteau, Treasury Department, Entergy Services J. Spath, Program Director, New York State Energy Research and Development Authority

SUMMARY OF FINDINGS

IR 05000333/2007-002; 01/01/2007 - 03/31/2007; James A. FitzPatrick Nuclear Power Plant;Maintenance Risk Assessment and Emergent Work Control.The report covered a three-month period of inspection by resident inspectors and region-basedinspectors. Two Green findings were identified, both of which were determined to be non-cited violations. 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 NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000. A.

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

A Green, self-revealing, non-cited violation (NCV) of Title 10 of the Codeof Federal Regulations (CFR), Part 50, Appendix B, Criterion V, "Instructions,

Procedures, and Drawings," was identified when Entergy failed to properly implement a torus exhaust valve maintenance procedure. As a result, on February 25, 2007, valve 27AOV-118 did not open on demand to vent the torus and maintain drywell to torus differential pressure. Entergy entered this issue into their corrective action program and performed an extent of condition review.The inspectors determined that this finding more than minor because it wasassociated with the Barrier Performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Failure of the valve to operate remotely from the relay room would have required operators to open the valve locally using the manual operator in accordance with procedure Emergency Procedure 6, "Post-Accident Containment Venting and Gas Control." The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A,

"Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance (Green)because it did not represent an actual open pathway in the physical integrity of reactor containment, or involve an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment. (Section 1R13)*Green. A Green, self-revealing, NCV of 10 CFR Part 50.65(a)(4), "Requirementsfor Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" was identified when Entergy did not perform a risk assessment for planned maintenance activities when a tagout was applied on the 'B' electro-hydraulic control (EHC) pump, in conjunction with a previous emergent failure of torus exhaust outer isolation valve 27AOV-118. Entergy performed a risk assessment and entered the deficiency into their corrective action program.

vThe inspectors determined that this finding affected the initiating eventscornerstone; and it was was more than minor because it was similar to Example 7(f) in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that, the emergent failure of 27AOV-118, in combination with the subsequent removal of the 'B' electro hydraulic control pump availability resulted in the plant being in a higher risk category, which required risk management actions, under Entergy's on-line risk management procedure. The inspectors evaluated this finding using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management SDP," Flowchart 1, "Assessment of Risk Deficit," and determined the finding to be of very low safety significance (Green)because the finding resulted in an increase in the incremental core damage probability deficit of less than 1 x 10

-6 (actual increase was in the high 10

-8range). The inspectors determined that this finding had a cross-cutting aspect in the areaof human performance because Entergy did not incorporate appropriate risk insights into planned work activities. (Section 1R13)

B.Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant StatusThe James A. FitzPatrick Nuclear Power Plant began the inspection period operating at fullpower. On February 17, 2007, the 'B' feedwater pump inboard seal exhibited increased leakage; and, as a result, the licensee elected to downpower to approximately 50 percent power to remove the feedwater pump from service. Following repairs, the plant was returned to full power on February 21, 2007, and continued to operate at or near full power for the remainder of the inspection period.1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R01Adverse Weather Protection (71111.01 - 1 sample)

a. Inspection Scope

In mid-February, Oswego county experienced significant lake effect snowfall exceedingtotals in excess of eight feet in some areas with snowfall rates of several inches per hour. The inspectors reviewed Entergy's preparations and response to these conditions including actions specified in Supplemental Action Procedure 19, "Severe Weather," and Administrative Procedure (AP) 12.04, "Seasonal Weather Preparations." In addition, the inspectors verified that Entergy took action to ensure that adequate operator and onsite staff were available during the storm. The inspectors also verified the operability of offsite and onsite emergency power supplies and that control room operators communicated with the transmission system operators in accordance with AP 12.13, "345/115 kV [kilovolt] Transmission Line Operations and Interface." This inspection satisfied one inspection sample for the onset of adverse weather.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04Q - 3 samples, 71111.04S - 1 sample).1Partial System Walkdown (3 samples)

a. Inspection Scope

The inspectors performed three partial system walkdowns to verify the operability ofredundant or diverse trains and components during periods of system train unavailability or following periods of maintenance. The inspectors referenced the system procedures, the Updated Final Safety Analysis Report (UFSAR), and system drawings in order to verify that the alignment of the available train was proper to support its required safety functions. The inspectors also reviewed applicable condition reports (CR) and work orders to ensure that Entergy had identified and properly addressed equipment discrepancies that could potentially impair the capability of the available train, as required by 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." The 2Enclosuredocuments reviewed are listed in the Attachment. The inspectors performed a partialwalkdown on the following systems which represented three inspection samples:*Train 'A' low pressure coolant injection (LPCI) independent power supply systemduring testing of the LPCI battery and battery charger;*Train 'B' emergency service water and emergency diesel generator systemsduring planned maintenance on the 'A' emergency service water train; and*Train 'B' residual heat removal system when the 'A' residual heat removal systemwas out of service for testing.

b. Findings

No findings of significance were identified..2Complete System Walkdown (1 sample)

a. Inspection Scope

The inspectors performed a complete system alignment inspection of the containmentatmosphere control and dilution system to identify any discrepancies between the existing equipment lineup and the required lineup. During the inspection, system drawings and operating procedures were used to verify proper equipment alignment and operational status. The inspectors reviewed the open maintenance work orders associated with the system for any deficiencies that could affect the ability of the system to perform its function. Documentation associated with unresolved design issues such as temporary modifications, operator workarounds, and items tracked by plant engineering were also reviewed to assess their collective impact on system operation.

In addition, the inspectors reviewed the CR database to verify that equipment problems were being identified and appropriately resolved. The documents reviewed during this inspection are listed in the Attachment. The inspection represented one inspection

sample.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05Q - 9 samples)

a. Inspection Scope

The inspectors conducted a tour of the areas listed below to assess the materialcondition and operational status of fire protection features. The inspectors verified that:

combustibles and ignition sources were controlled in accordance with Entergy's administrative procedures; fire detection and suppression equipment was available for use; passive fire barriers were maintained; and compensatory measures for out-of-service, degraded, or inoperable fire protection equipment were implemented in accordance with Entergy's fire plan. The inspectors used procedure ENN-DC-161, 3Enclosure"Transient Combustible Program," in performing the inspection. The inspectorsevaluated the fire protection program against the requirements of License Condition 2.C.3. The documents reviewed are listed in the Attachment. This inspection represented nine inspection samples for fire protection tours and were conducted in the following areas:: *Fire Area/Zone 1E/TB-1 North, elevation 252 foot;*Fire Area/Zone 1E-TB-1 South, elevation 252 foot;

  • Fire Area/Zone 1E/TB-1 North, elevation 272 foot;
  • Fire Area/Zone IX/SG-1;
  • Fire Area/Zone IX/RB-1A, elevation 272 foot;
  • Fire Area/Zone II/SW-1, elevation 272 foot;
  • Fire Area/Zone IC/SW-1, elevation 272 foot;
  • Fire Area/Zone V/EG-1 South, elevation 272 foot; and
  • Fire Area/Zone VII/CS-1, elevation 272 foot.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures Internal Flooding (71111.06 -1 sample)

a. Inspection Scope

The inspectors reviewed selected risk-important plant design features and Entergy'sprocedures intended to protect the cable spreading and relay rooms and associated safety-related equipment from internal flooding events. The inspectors reviewed flood analysis and design documents, including the Individual Plant Examination and the UFSAR, engineering calculations, and abnormal operating procedures. The documents reviewed are listed in the Attachment. These activities represented one inspection

sample.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance (71111.07A - 1 sample)

a. Inspection Scope

The inspectors reviewed the testing and evaluation of test results for the residual heatremoval system heat exchanger 10E-2B performed in accordance with Entergy's response to NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." Heat removal measurements and heat exchanger capacity calculations were reviewed to verify that cooler performance was consistent with design 4Enclosurecalculations and the UFSAR. The documents reviewed are listed in the Attachment. These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program.1Resident Inspector Quarterly Review (71111.11Q - 1 sample)

a. Inspection Scope

On March 1, 2007, the inspectors observed licensed operator simulator training toassess operator performance during several scenarios to verify that operator performance was adequate and evaluators were identifying and documenting crew performance problems. The inspectors evaluated the performance of risk significant operator actions, including the use of emergency operating procedures. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance of timely control board operation and manipulation, and the oversight and direction provided by the shiftmanager. The inspectors also reviewed simulator fidelity to evaluate the degree of similarity to the actual control room. Licensed operator training was evaluated against the requirements of 10 CFR Part 55, "Operators' Licenses." The documents reviewed are listed in the Attachment. This observation of operator simulator training represented one inspection sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q - 2 samples)

a. Inspection Scope

The inspectors reviewed performance-based problems involving selected in-scopestructures, systems, or components (SSCs) to assess the effectiveness of the maintenance program. The reviews focused on:*Proper Maintenance Rule scoping in accordance with 10 CFR Part 50.65;*Characterization of reliability issues;

  • Changing system and component unavailability;
  • Identifying and addressing common cause failures;
  • Trending of system flow and temperature values;
  • Appropriateness of performance criteria for SSCs classified (a)(2); and
  • Adequacy of goals and corrective actions for SSCs classified (a)(1).

5EnclosureThe inspectors reviewed system health reports, maintenance backlogs, andMaintenance Rule basis documents. The inspectors evaluated the maintenance program against the requirements of 10 CFR Part 50.65. The documents reviewed are listed in the Attachment. The following Maintenance Rule samples were reviewed and represent two inspection samples:*125 volt station and 419 volt low pressure coolant injection batteries, chargers,and inverters; and*Residual heat removal service water system.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 5 samples)

a. Inspection Scope

The inspectors reviewed maintenance activities to verify that the appropriate riskassessments were performed prior to removing equipment for work. 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 documents reviewed are listed in the Attachment. The review of the following activities represented five inspection samples:*Week of January 8, 2007, which included biennial maintenance on the 'A' trainlow pressure coolant injection inverter and full load surveillance testing of the 'A' train emergency diesel generator; *Week of January 22, 2007, which included full load testing of 'B' train emergencydiesel generators, quarterly inservice testing of the high pressure coolant injection system, and service testing of the 'B' train low pressure coolant injection battery and inverter; *Week of February 19, 2007, which included emergent work on the 'B' feedwaterpump due to a degraded inboard seal and 'C' condensate pump packing replacement;*Week of February 26, 2007, which included emergent work on the torus exhaustouter isolation valve, 27-AOV-118, emergency diesel generator surveillance testing, and high pressure coolant injection system instrument surveillance testing; and*Week of March 12, 2007, which included 'B' reactor water cleanup pump sealfailure and system isolation with the 'A' steam packing exhauster and 'C' residual heat removal service water pump out of service.

b. Findings

.1Introduction: A Green, self-revealing, non-cited violation (NCV) of 10 CFR 50,Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified when Entergy failed to properly implement a torus exhaust valve maintenance. As a result, on February 25, 2007, valve 27AOV-118 did not open on demand to vent the torus to maintain drywell-to-torus differential pressure.Description: Valve 27AOV-118 is a normally closed 20-inch air-operated containmentisolation valve in the containment atmosphere control system. During normal operation the valve is remotely operated from a relay room control panel to vent the torus via the standby gas treatment system to maintain drywell-to-torus differential pressure. During loss of normal decay heat removal events, the valve is used for containment pressure control and reactor decay heat removal. The system is manually initiated from the relay room in accordance with emergency operating procedures when drywell pressure reaches 44 psig. The valve can also be operated using a separately mounted manual operator.On February 25, 2007, while attempting to vent the torus, Entergy identified that valve27AOV-118 could not be opened from the relay room. Investigation revealed that the key connecting the air actuator coupling hub to the valve coupling hub had sheared and that the gap between the hubs appeared to be excessive. A similar key failure occurred in 1996 and had been attributed to an excessive coupling hub gap which overstressed the key. Based on the results of calculation JAF-CALC-CAD-02766, "Containment Purge Actuator to Valve Coupling Gap for Valves 27AOV-101A/B and 111 Through 118," an allowable range of 0.0625 to 0.125 inches was established and incorporated on December 19, 2002, into the valve actuator Maintenance Procedure (MP) 060.02, "GH Bettis Pneumatic Valve Actuator Maintenance," Revision 5. Following the February 25, 2007 key failure, Entergy found the coupling hub gap to be 0.625 inches.

Analysis:

The inspectors determined that the performance deficiency was that Entergyfailed to implement the actuator-to-valve coupling gap limit specified in the maintenance procedure. This was reasonably within Entergy's ability to foresee and prevent.

Traditional enforcement does not apply because the issue did not have an actual safety consequence or a potential for impacting the NRC's regulatory function, and it was not the result of any willful violation of NRC requirements.The inspectors determined that this finding was more than minor because it wasassociated with the Barrier Performance attribute of the Barrier Integrity cornerstone; and it impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radio nuclide releases caused by accidents or events. Failure of the valve to operate remotely from the relay room would have required operators to open the valve locally using the manual operator in accordance with procedure Emergency Procedure 6, "Post-Accident Containment Venting and Gas Control." The inspectors evaluated this finding using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance (Green)7Enclosurebecause it did not represent an actual open pathway in the physical integrity of reactorcontainment, or involve an actual reduction in defense-in-depth for the atmospheric pressure control or hydrogen control functions of the reactor containment.Enforcement: 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, andDrawings," requires that activities affecting quality shall be accomplished in accordance with instructions, procedures, or drawings. Entergy Maintenance Procedure (MP)060.02, "GH Bettis Pneumatic Valve Actuator Maintenance," Revision 5, requires adjustment of the valve stem coupling hub to hub gap be 0.0625 to 0.125 inches on both the manual and air operator side of valve 27AOV-118 as measured at the key interface locations. Contrary to the above, in October 2004, Entergy did not maintain the valve stem coupling hub to hub gap on valve 27AOV-118 in accordance with maintenance procedure MP-060.02 resulting in failure of the actuator-to-valve coupling hub key on February 25, 2007. Because the issue was of very low safety significance (Green) and was entered into Entergy's corrective action program as condition report CR-JAF-2007-00752, this violation is being treated as an NCV consistent with Section VI.A.1 of theNRC Enforcement Policy: (NCV 05000333/2007002-01, Inadequate Maintenance onContainment Atmosphere Control Valve.).

Introduction:

A Green, self-revealing NCV of 10 CFR Part 50.65 (a)(4), "Requirementsfor Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" was identified when Entergy did not perform a risk assessment for planned maintenance activities when a tagout was applied on the 'B' electro-hydraulic control (EHC) pump, in conjunction with a previous emergent failure of torus exhaust outer isolation valve 27AOV-118.

Description:

On February 25, 2007, at 11:10 p.m., 27AOV-118 would not open remotelyto vent the torus. 27AOV-118 is a primary containment isolation valve, and, as required by TS 3.6.1.3, operators isolated the affected containment penetration flow path by closing and deactivating the inboard isolation valve. Subsequently, a tagout was applied on the 'B' EHC pump for scheduled maintenance. A risk assessment that considered the emergent failure of 27AOV-118 and the impact on containment venting prior to application of the tagout on the 'B' EHC pump was not performed. Specifically, Entergy's risk assessment failed to consider risk significant systems, structures, andcomponents, as well as support systems that were unavailable during the maintenance.

Administrative Procedure (AP) 10.10, "On-Line Risk Assessment," assigns a risk category color in risk significant order from Green, Yellow, Orange or Red based on core damage frequency. Risk management actions are implemented depending on the risk category color. With 27AOV-118 inoperable, combined with the impact of the tagout on the EHC system, the color increased from Yellow to Orange. On February 26, at approximately 6:46 a.m., the plant staff recognized the condition and the 'B' EHC pump was returned to standby.Analysis: The inspectors determined that the finding was a performance deficiencybecause Entergy did not perform a risk assessment following the emergent failure of 27AOV-118, and subsequently continued scheduled work activities including tagout of 8Enclosurethe 'B' EHC pump. The inspectors determined that this finding impacted the InitiatingEvents cornerstone due to the increased likelihood of a plant transient with the loss of the normal heat sink. It was reasonable that Entergy should have identified the condition and updated the risk assessment, because Entergy procedure AP 10.10, "On-Line Risk Assessment," specifies that plant risk must be reassessed when plant conditions change. Traditional enforcement does not apply since there were no actual safety consequences or potential for impacting the NRC's regulatory function, and the finding was not the result of any willful violation of NRC requirements or Entergy's procedures. The inspectors determined that this finding was more than minor because it was similarto Example 7(f) in Inspection Manual Chapter 0612, Appendix E, "Examples of Minor Issues," in that, the emergent failure of 27AOV-118, in combination with the subsequent removal of the 'B' electro hydraulic control pump availability resulted in the plant being in a higher risk category, which required risk management actions, under Entergy's on-line risk management procedure. The ability to vent the containment from the torus is a risk important action to prevent containment failure and core damage for situations that involve the inability to remove core decay heat from the torus water. The tagout of the

'B' electro hydraulic control pump increased the chance of a plant transient with the loss of the normal heat sink.Using IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk ManagementSDP," Flowchart 1, "Assessment of Risk Deficit," the inspectors determined the incremental core damage probability deficit from Entergy's core damage frequency as a result of the actual duration of the 27AOV-118 maintenance combined with the time the

'B' EHC system was not available due to the tagout (two hours). The inspectors calculated the incremental core damage probability deficit to be in the high 10

-8 range. Because the calculated risk deficit was not greater than 1 x 10

-6 incremental coredamage probability deficit, the inspectors determined that this finding was of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area ofhuman performance because Entergy did not incorporate appropriate risk insights into planned work activities.Enforcement: 10 CFR 50.65 (a)(4), requires, in part, that before performingmaintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities.

Contrary to the above, on February 26, 2007, Entergy did not assess and manage the increase in risk for planned maintenance activities when a tagout was applied on the 'B' EHC pump, following an emergent failure of torus exhaust outer isolation valve 27AOV-118. Because this finding was of very low safety significance and was entered into Entergy's corrective action program as condition report CR-JAF-2007-00755, this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000333/2007002-02, Failure to Perform a Risk Assessment When Required by 10 CFR 50.65(a)(4).

9Enclosure1R15Operability Evaluations (71111.15 - 6 samples)

a. Inspection Scope

The inspectors reviewed operability determinations to assess the acceptability of theevaluations; when needed, the use and control of compensatory measures; and compliance with TS. The inspectors' review included a verification that the operability determinations were made as specified by ENN-OP-104, "Operability Determinations."

The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR, and associated design basis documents. The documents reviewed are listed in the Attachment. The following evaluations were reviewed and represented six inspection samples:*CR-2007-00151 concerning a commercial grade gate timing circuit card installedin safety-related low pressure coolant injection inverter 71INV-3A;*CR-2007-00293 concerning installation of nonsafety-related contacts insafety-related circuit breaker 71MCC-163-OG-1 for containment isolation valve 10MOV-34B; *CR-2007-00392 concerning an out of calibration condition of a local leak ratemonitor panel that affected the test results of 12 containment penetrations;*CR-2007-00647 concerning inaccurate valve position indication of vacuumbreaker 27VB-2 while performing surveillance test ST-15J, "Torus to Drywell Vacuum Breakers Quarterly Test IST;"*CR-2007-00281 concerning 'D' safety relief valve tailpipe temperature increase; and*CR-2007-00752 concerning extent of condition reviews following failure ofcontainment atmosphere control isolation valve 27AOV-118.

b. Findings

No findings of significance were identified.

1R19 Post Maintenance Testing (71111.19 - 6 samples)

a. Inspection Scope

The inspectors reviewed six post-maintenance test procedures and associated testingactivities for selected risk significant mitigating systems to assess whether the effect of maintenance on plant systems was adequately addressed by control room and engineering personnel. The inspectors verified: test acceptance criteria were clear, demonstrated operational readiness and were consistent with Design Basis Documents; test instrumentation had current calibrations and adequate range and accuracy for the application; and tests were performed, as written, with applicable prerequisites satisfied.

Upon completion, the inspectors verified that equipment was returned to the proper alignment necessary to perform its safety function. Post-maintenance testing was evaluated against the requirements of 10 CFR 50, Appendix B, Criterion XI, "Test 10EnclosureControl." The documents reviewed are listed in the Attachment. The following post-maintenance test activities were reviewed and represented six inspection samples:*Work request JAF-03-07188-01, involving repair of a water leak in service waterpump 46P-1B motor oil reservoir during the week of June 16, 2007;*Work request [[::JAF-05-17157|JAF-05-17157]], involving inspection and repair of a residual heatremoval service water keep-full check valve;*Work request [[::JAF-05-22322|JAF-05-22322]], involving cleaning and replacement of two-inchemergency service water piping to the west electric bay and east cable tunnel coolers;*Work request [[::JAF-07-15650|JAF-07-15650]], involving repair of the air operator coupling of torusexhaust outboard containment isolation valve 27AOV-118;*Work request 51104288, involving replacement of residual heat removal servicewater pump 10P-1C due to high vibration; and*Work request 51178018, involving maintenance and repair of stator watercooling pump 94P-15B due to high vibration.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22 - 7 samples)

a. Inspection Scope

The inspectors witnessed performance of surveillance tests and/or reviewed test data ofselected risk-significant SSCs to assess whether the SSCs satisfied TS, UFSAR, Technical Requirements Manual, and Entergy procedure requirements. The inspectors verified: test acceptance criteria were clear, demonstrated operational readiness, and were consistent with Design Basis Documents; test instrumentation had current calibrations and adequate range and accuracy for the application; and tests were performed, as written, with applicable prerequisites satisfied. Upon surveillance test (ST) completion, the inspectors verified that equipment was returned to the status specified to perform its safety function. The inspectors evaluated the tests against the requirements in TS. The following STs were reviewed and represented seven inspection program samples:*ST-2AM, "Residual Heat Removal Loop B Quarterly Operability Test;"*ST-4N, "High Pressure Coolant Injection Quick Start, Inservice, and TransientMonitoring Test;" *MST-071.30, "Low Pressure Coolant Injection Charger-Inverter Performance andLow Pressure Coolant Injection Battery Service Surveillance Test;" *ST-7F, "Standby Gas Treatment Fan B and Valve Exercising Test IST;"

  • ST-9BB, "Emergency Diesel Generator 'B' and 'D' Full Load Test and EmergencyService Water Pump Operability Test;" and*ST-5BB, "APRM System 'B' Channel Functional Test."

b. Findings

No findings of significance were identified.Cornerstone: Emergency Preparedness1EP6Drill Evaluation (71114.06 - 1 sample)

a. Inspection Scope

The inspectors observed emergency response organization activities during the fullparticipation drill that was conducted on January 11, 2007. The inspectors verified that emergency classification declarations, notifications, and protective action recommendations were properly completed. The inspectors evaluated the drill against the requirements of 10 CFR Part 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities." This observation constituted one inspection program sample.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone:

Occupational Radiation Safety 2OS2ALARA Planning and Controls (71121.02 - 4 samples)

a. Inspection Scope

The inspector conducted the following activities to verify that Entergy was properlymaintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). Implementation of the ALARA program was reviewed against the criteria contained in 10 CFR 20.1101(b) and Entergy's procedures.(1)The following five highest exposure work activities from the Fall 2006 refuelingoutage were selected for review:*Reactor disassembly/reassembly;*In-Service inspection/ erosion-corrosion/intergranular stress corrosioncracking;*Reactor vessel visual inspection and defueling activities;

  • Control rod drive replacement; and
  • Preventive maintenance on motor-operated valves.

12Enclosure(2)With respect to the work activities listed above, the ALARA evaluations,exposure estimates, and applicable exposure mitigation requirements were reviewed. This included a review of exposure mitigation procedures and engineering and work controls to achieve exposures that are ALARA. These work activities were also reviewed to determine if they were reasonably grouped into work activities, based on historical precedence or industry standard groupings.(3)The actual results achieved were compared with the intended dose establishedin the ALARA planning for the work activities. ALARA post-job reviews were reviewed and interviews were conducted to evaluate the adequacy of ALARA controls as implemented, and to identify any significant performance deficiencies that may have resulted in unintended dose consequences.(4)The methodology for adjusting work activity exposure estimates were evaluatedwith respect to the work activities. The reasons for the exposure estimate adjustments were determined and evaluated with respect to sound radiation protection and ALARA principles and to ensure the revised exposure estimates provided an effective ALARA performance measure.

b. Findings

No findings of significance were identified.4.OTHER ACTIVITIES4OA2Identification and Resolution of Problems.1Annual Sample: Operator Workaround Program (71152 - 1 sample)

a. Inspection Scope

The inspectors reviewed the cumulative effects of operator workarounds on thereliability, availability, and potential for mis-operation of a system and on the operator's ability to implement abnormal or emergency operating procedures. The inspectors reviewed the results of Entergy Surveillance Test ST-99H, "Operations Cumulative Impact Assessment," and the resolution of items identified in the assessment. The inspectors reviewed Entergy's program for identifying operator workarounds at an appropriate threshold and for entering them into the corrective action program. In addition, inspectors reviewed operation department records including standing orders for operational decision-making issues and operability evaluations.

b.Assessment and ObservationsNo findings of significance were identified. The corrective action program waseffectively used to identify and resolve operator workarounds. The resolution of operator workaround items has been appropriately prioritized.

13Enclosure.2Review of Items Entered into the Corrective Action Program

a. Inspection Scope

As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of all items entered into Entergy's corrective action program. The review was accomplished by accessing Entergy's computerized database for CRs and attending CR screening meetings.In accordance with the baseline inspection modules, the inspectors selected correctiveaction program items across the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for additional follow-up and review. Additionally, an NRC specialist inspector reviewed six CRs associated with the Occupational Radiation Safety Cornerstone that were initiated between October 2006 and January 2007. The inspectors assessed Entergy's threshold for problem identification, the adequacy of the cause analyses, extent of condition review, operability determinations, and the timeliness of the specified corrective actions. The CRs reviewed are listed in the

. b.Assessment and Observations No findings of significance were identified.

4OA3Event Followup (71153 - 1 sample).1(Closed) LER 05000333/2006002-00, High Pressure Coolant Injection System DeclaredInoperable Due to Turbine Speed Oscillations.On November 4, 2006, with the plant operating in Mode 1, Entergy identified that thehigh pressure coolant injection system was inoperable due to turbine speed oscillations.

The condition was discovered during post-maintenance testing following a refueling outage, and was caused by connecting two turbine hydraulic actuator oil lines to the incorrect oil ports. The enforcement aspects of this violation of maintenance procedures were documented in section

4OA7 of NRC Inspection Report 05000333/2006005.

Entergy entered the event into its corrective action program as CR-2006-04754. This LER is closed. 4OA5Other Activities.1Closed: URI 2006006-03, "Simulator Transient Testing"This URI was opened because the inspectors determined that the facility process forreviewing simulator transient testing was not adequate to comply with 10 CFR Part 55.46(d)(1) and Regulatory Guide 1.149, "Nuclear Power Plant Simulation Facilities for use in Operator Training and License Examinations," Revision 1. The inspectors 14Enclosureconsidered this issue unresolved pending an appropriate review of test results todetermine if significant discrepancies existed in simulator performance.Entergy put together an "expert panel" of six individuals from operations, engineering,and training. This use of an expert panel satisfies regulatory requirements for test review. These individuals reviewed the most recent simulator transient tests and identified discrepancies. Two discrepancies were identified by the expert panel review which resulted in simulatormodel changes. One instance involved solid plant pressure response at high pressure, the other involved indications of recirculation flow under certain conditions with allrecirculation pumps tripped.The inspectors evaluated the impact of these discrepancies against the criteriadescribed in NRC Inspection Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance significance determination process," and determined that these discrepancies were minor. Accordingly, this URI is closed.4OA6Meetings, Including ExitOn March 27, 2007, the inspectors presented the inspection results toMr. Peter T. Dietrich and other members of his staff. The inspectors asked Entergy whether any of the material examined during the inspection should be considered proprietary. No proprietary information was identified.ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

P. Dietrich, Site Vice President
C. Adner, Manager, Operations
N. Avrakotos, Manager, Emergency Preparedness
S. Bono, Director Engineering
J. Costedio, Manager, Regulatory Compliance
M. Durr, Manager, System Engineering
B. Finn, Director, Nuclear Safety Assurance
D. Johnson, Manager, Training
J. LaPlante, Manager, Security
K. Mulligan, General Manager, Plant Operations
J. Pechacek, Manager, Programs and Components Engineering
W. Rheaume, Manager, CA&A
J. Solowski, Radiation Protection

John Pircsuk, Senior Operations Instructor

Darren Deritz, Regulatory Compliance

Greg Pitts, Training

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and

Closed

05000333/LER-2006-002-00LERHigh Pressure Coolant Injection SystemDeclared Inoperable Due to Turbine Speed
Oscillations (Section 4OA3)
05000333/FIN-2007002-01NCVInadequate Maintenance on ContainmentAtmosphere Control Valve (Section 1R13)
05000333/FIN-2007002-02NCVFailure to Perform a Risk AssessmentWhen Required by 10 CFR 50.65(a)(4)

(Section 1R13)

Closed

05000333/LER-2006-002-00LERHigh Pressure Coolant Injection SystemDeclared Inoperable Due to Turbine Speed
Oscillations (Section 4OA3)
05000333/FIN-2007002-01NCVInadequate Maintenance on ContainmentAtmosphere Control Valve (Section 1R13)
05000333/FIN-2007002-02NCVFailure to Perform a Risk AssessmentWhen Required by 10 CFR 50.65(a)(4)

(Section 1R13)

LIST OF DOCUMENTS REVIEWED

Section 1R04: Equipment AlignmentOP-43C, "Low Pressure Coolant Injection Independent Power Supply System," Revision 17MST-071.11, "Low Pressure Coolant Injection Battery Quarterly Surveillance Test," Revision 19

A-2AttachmentDBD-027, "Design Basis Document for The Air Treatment Systems," Revision 10OP-37, "Containment Atmosphere Dilution System," Revision 73
FM-18A, "Flow Diagram - Drywell Inerting C.A.D. and Purge," Revision 57
FM-18B, "Flow Diagram - Drywell Inerting C.A.D. Purge and Containment Differential Pressurization," Revision 37
14620.9011-US(N)-001, "Torque Calculation for Suppression Chamber (20") and Drywell (24")
Vent and Purge Butterfly Valves," Revision 2
JAF-CALC-CAD-02782, "Nitrogen Flows and Volumes for Normal and Accident Conditions," Revision 0

Section 1R05: Fire ProtectionPFP-PWR42, Fire Area/Zone 1E/TB-1PFP-PWR43, Fire Area/Zone 1E/TB-1

PFP-PWR45, Fire Area/Zone 1E/TB-1
PFP-PWR22, Fire Area/Zone IX/SG-1
PFP-PWR20, Fire Area/Zone IX/RB-1A
PFP-PWR29, Fire Area/Zone II/SW-1
PFP-PWR30, Fire Area/Zone IC/SW-1
PFP-PWR31, Fire Area/Zone V/EG-1
PFP-PWR11, Fire Area/Zone VII/CS-1
ST-76J13, "Heat Detector Functional Test - SGT Filter B," Revision 16

Section 1R06: Flood Protection MeasuresAOP-43, "Plant Shutdown from Outside the Control Room," Revision 32AOP-51, "Unexpected Fire Pump Start," Revision 5

AP-16.14, "Hazard Barrier Controls," Revision 3
11825-FE-1AS, "120V AC One Line Diagram, Emergency Bus A2 & B2 Distribution Panels
71ACA2 & 71ACB2," Revision 18
LO-OEN-2005-00193, "Response to NRC Information Notice 2005-011, 'Internal Flooding/Spray-Down of Safety-Related Equipment Due to Unsealed Equipment Hatch Floor Plugs and/or Blocked Floor Drains'"
LO-OEN-2005-00555, "Response to NRC Information Notice 2005-030, 'Safe Shutdown Potentially Challenged by Unanalyzed Internal Flooding Events and Inadequate Design'"
0090-0006-C-003, "JAF NPP Fire Suppression Effects Analysis," Revision 0
DBD-071 Tab 1, "Design Basis Document for the Electrical Distribution Systems 4160V and
600V AC Power Systems," Revision 2
JAF-RPT-MULTI-02107, "IPE Update, Appendix H, Internal Flooding Analysis," Revision 2

Section 1RO7: Heat Sink PerformanceAOP-30, "Loss of Shutdown Cooling," Revision 19AP-19.14, "Eddy Current Testing of Heat Exchanger Tubes," Revision 8

EOP-4, "Primary Containment Control," Revision 7
OP-13, "Residual Heat Removal System," Revision 93
A-3AttachmentST-2Y, "Residual Heat Removal Heat Exchanger Performance Test," Revision 7, datedJanuary 24, 2007
JAF-CALC-MISC-02634, "Torus Station Blackout Torus Heat-Up for Power Uprate Conditions," Revision 0
[[::JAF-CALC-Re|JAF-CALC-Re]]sidual Heat Removal-01903, "Instrument Indication Uncertainty for Residual HeatRemoval Heat Exchanger Performance Test," Revision 1
[[::JAF-CALC-Re|JAF-CALC-Re]]sidual Heat Removal-02953, "Residual Heat Removal Heat Exchanger K-value with Reduced Tube Side Fouling Factor," Revision 0
[[::JAF-CALC-Re|JAF-CALC-Re]]sidual Heat Removal-00392, "Calculation for Design Basis/Acceptance Criteriafor
ST-2Y," Revision 0
GE-NE-T23-00373-01, "Higher Residual Heat Removal Service Water Temperature Analysis," Revision 1
GE-NE-T23-00766-00-01, "Containment Analysis to Support ECCS Pump NPSH Evaluation," Revision 0
JPN-91-015, "Updated Response to Generic Letter 89-13 Service Water System Problems Affecting Safety-Related Equipment," dated April 18, 1991
JPN-93-015, "Updated Response to Generic Letter 89-13 Service Water System Problems Affecting Safety-Related Equipment," dated March 16, 1993

"Eddy Current Inspection for Residual Heat Removal B Heat Exchanger," dated April 17, 2003

Master Lee NDE Services, "Inspection Summary," dated April 2003

Section 1R11: Licensed Operator Requalification Program71775-0, Loss of B Reactor Protection System Bus, failure to Automatically Scram, with aScram Discharge Isolation Valve Rupture After the Manual ScramSection 1R12: Maintenance EffectivenessJAF-RPT-ELEC-02302, "MR Basis Document for System 071

DC Electrical Distribution,Revision 3
JAF-RPT-MULTI-02294, "MR Basis Document for Service Water Systems Including System
010 - Residual Heat Removal Service Water - System 046 -- Emergency Service Water -
System 046-000 - Normal Service Water," Revision 6

Section 1R15: Operability EvaluationsST-39B, "Type B and C

LLRT of Containment Penetrations," Revision 31

Section 1R19: Post Maintenance TestingJAF-RPT-MULTI-02294, "MR Basis Document for Service Water Systems," Revision 6AP-05.07, "Maintenance Testing and Post-Work Testing (ISI)," Revision 36

MP-059.45, "Piston Check Valves," Revision 10
JAF-RPT-MULTI-00406, "Inservice Test Program Basis Document," Revision 0

Section 1R22: Surveillance Testing

A-4AttachmentIEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, andReplacement of Vented Lead-Acid Batteries for Stationary Applications"
MP-057-.06, "Battery Maintenance," Revision 27
[[::JAF-89-026|JAF-89-026]], "Establish Acceptance Criteria for Low Pressure Coolant Injection Battery Intercell Connector Resistances," Revision 0
JAF-CALC-ELEC-00562, "Low Pressure Coolant Injection Battery Testing Duty Cycle," Revision
JAF-CALC-ELEC-01857, "419 V DC Low Pressure Coolant Injection Power System 3A & 3B
Sizing," Revision 0
JAF-CALC-ELEC-02213, "Low Pressure Coolant Injection UPS System Testing Load Bank Characteristics and Low Pressure Coolant Injection Battery and Inverter on Line Testing Conditions and/or Limitation," Revision 0
JAF-CAL-ELEC-00523, "71/Low Pressure Coolant Injection UPS System Testing Duty Cycle," Revision 3 and Margin Revisions 3A and 0A
JAF Manual No. E356-0048, "Stationary Lead-Acid Batteries"
JAF-CALC-NMS-00758, "Setpoint Calculation for APRM A through F," Revision 11
JTS-93-0877, "Surveillance Test Adequacy Review of APRMs and Control Rod Blocks Findings," Revision 1Section 4OA2: Identification and Resolution of Problems Condition Reports2003-021042003-02269
2000-06351
2003-01787
2006-04276
2006-05039
2006-04873
2006-05108
2006-05021
2006-04965
2006-04645
2006-05106
2007-00151
2007-00127
2007-003762007-012362007-01019
2007-01128
2007-00914
2007-00884
2007-00909
2007-00862
2007-00809
2007-00804
2007-00824
2007-00647
2007-00641
2007-00629
2007-00628
2007-006252007-004972007-00531
2007-00454
2007-00392
2007-01045
2007-01043
2007-00592
2007-00719
2007-00983
2007-00845
2007-00241
2007-00104
2007-00740
2007-00741
2007-01196

Section 4OA5: Other ActivitiesTP-7.03, "Simulator Test Program," Revision 1CR-JAF-2006-02057

SDR-9494, ANSI Transient Questions
SDR-9297, ANSI06 Max Size Unisolable Steam Rupture
A-5

LIST OF ACRONYMS

ADAMSAgency Documents Access Management SystemALARAas low as is reasonably achievable

APadministrative procedure

CDFcore damage frequency

CFRCode of Federal Regulations

CRcondition report

DBDdesign basis document

EHCelectro-hydraulic control

EOPemergency operating procedure

EROemergency response organization

kVkilovolt

IMCinspection manual chapter

ISIinservice inspection

ISTinservice testing

LPCIlow pressure coolant injection

MPmaintenance procedure

NCVnon-cited violation

NRCNuclear Regulatory Commission

OPoperating procedure

PARSpublicly available records

psigpounds per square inch gauge

SDPsignificance determination process

SSCstructures, systems or components

STsurveillance test

TStechnical specification

UFSARUpdated Final Safety Evaluation Report

URI unresolved item
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