IR 05000247/2016002: Difference between revisions

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[[Issue date::March 30, 2017]]
[[Issue date::March 30, 2017]]


EA-16-223 Mr. Anthony Site Vice President Entergy Nuclear Operations, Inc.
EA-16-223  


Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249
Mr. Anthony Site Vice President Entergy Nuclear Operations, Inc.
 
Indian Point Energy Center  
 
450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249


SUBJECT: INDIAN POINT NUCLEAR GENERATING - REVISED INTEGRATED INSPECTION REPORT 05000247/2016002 AND 05000286/2016002
SUBJECT: INDIAN POINT NUCLEAR GENERATING - REVISED INTEGRATED INSPECTION REPORT 05000247/2016002 AND 05000286/2016002


==Dear Mr. Vitale:==
==Dear Mr. Vitale:==
In a letter dated September 29, 2016 (ML16281A4861), Entergy Nuclear Operations, Inc. (Entergy), provided a response to U.S. Nuclear Regulatory Commission (NRC) inspection report 05000247/2016002 and 05000286/2016002 (ML16243A245) issued on August 30, 2016. In this letter, Entergy disagreed with non-cited violation (NCV) 05000286/2016002-02, "Failure to Follow Operability Determination Procedure for Unit 3 Baffle-Former Bolts." This was a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR), Appendix B, Criterion V, "Instructions, Procedures, and Drawings," and described a failure by Entergy to follow the operability process as prescribed in Entergy procedures at Indian Point concerning degraded baffle-former bolts in the Unit 2 reactor vessel. Specifically, Entergy determined that a more detailed operability evaluation for the Unit 3 reactor vessel was not required in accordance with procedure EN-OP-104, "Operability Determination Process." In the aforementioned letter, Entergy provided the NRC with an interpretation of procedure EN-OP-104 related to the baffle-former bolt condition that was different from the NRC's initial position as described in the inspection report. In response to the contested NCV, the NRC formed a panel to review the information provided by Entergy. The staff members of the panel were independent of the initial inspection effort.
In a letter dated September 29, 2016 (ML16281A486 1), Entergy Nuclear Operations, Inc.


Based on a thorough review of the issue, the NRC concluded that the violation as written could not be supported. Specifically, the independent review noted different steps within the procedure that could be interpreted to either support or refute Entergy's claim related to the need for direct evidence of a degraded or non-conforming condition. As a result of this potential for differing interpretations of the procedural requirements, the NRC determined it was not appropriate to issue a violation for failure to comply with the procedure. The NRC issued the withdrawal of the NCV in a March 2, 2017 letter (ML17061A112). Therefore, the NRC is   1 The designation in parentheses refers to an Agencywide Documents Access and Management System (ADAMS) Accession Number. Documents referenced in this letter are publicly available using the Accession Number in ADAMS. removing the NCV documented in Section 1R15 of the enclosed inspection report 05000247/2016002 and 05000286/2016002.
(Entergy), provided a response to U.S. Nuclear Regulatory Commission (NRC) inspection report 05000247/2016002 and 05000286/2016002 (ML16243A245) issued on August 30, 2016. In this letter, Entergy disagreed with non-cited violation (NCV) 05000286/2016002-02, "Failure to Follow Operability Determination Procedure for Unit 3 Baffle-Former Bolts." This was a Green
 
NCV of Title 10 of the Code of Federal Regulations (10 CFR), Appendix B, Criterion V, "Instructions, Procedures, and Drawings," and described a failure by Entergy to follow the operability process as prescribed in Entergy procedures at Indian Point concerning degraded baffle-former bolts in the Unit 2 reactor vessel. Specifically, Entergy determined that a more detailed operability evaluation for the Unit 3 reactor vessel was not required in accordance with procedure EN-OP-104, "Operability Determination Process." In the aforementioned letter, Entergy provided the NRC with an interpretation of procedure EN-OP-104 related to the baffle-former bolt condition that was different from the NRC's initial position as described in the inspection report.
 
In response to the contested NCV, the NRC formed a panel to review the information provided by Entergy. The staff members of the panel were independent of the initial inspection effort.
 
Based on a thorough review of the issue, the NRC concluded that the violation as written could not be supported. Specifically, the independent review noted different steps within the procedure that could be interpreted to either support or refute Entergy's claim related to the need for direct evidence of a degraded or non-conforming condition. As a result of this potential for differing interpretations of the procedural requirements, the NRC determined it was not appropriate to issue a violation for failure to comply with the procedure. The NRC issued the withdrawal of the NCV in a March 2, 2017 letter (ML17061A112). Therefore, the NRC is  
 
1 The designation in parentheses refers to an Agencywide Documents Access and Management System (ADAMS) Accession Number. Documents referenced in this letter are publicly available using the Accession Number in ADAMS. removing the NCV documented in Section 1R15 of the enclosed inspection report 05000247/2016002 and 05000286/2016002.


This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding."
This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding."
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This performance deficiency is more than minor because it was associated with the configuration control (shutdown equipment lineup) attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A detailed risk assessment was conducted and determined that the change in core damage frequency was determined to be 7E-9, therefore, this issue represents a Green finding. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Entergy did not consider potential undesired consequences of actions before performing work and implement appropriate error-reduction tools. Specifically, the work crew did not understand the requirements and potential consequences prior to commencing work and the gate monitor did not enforce these requirements to maintain at least one gate locked or pinned closed as required by OAP-007.
This performance deficiency is more than minor because it was associated with the configuration control (shutdown equipment lineup) attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A detailed risk assessment was conducted and determined that the change in core damage frequency was determined to be 7E-9, therefore, this issue represents a Green finding. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Entergy did not consider potential undesired consequences of actions before performing work and implement appropriate error-reduction tools. Specifically, the work crew did not understand the requirements and potential consequences prior to commencing work and the gate monitor did not enforce these requirements to maintain at least one gate locked or pinned closed as required by OAP-007.


[H.12 - Avoid Complacency] (Section 1R20)
[H.12 - Avoid Complacency] (Section 1R20)  


===Cornerstone: Barrier Integrity===
===Cornerstone: Barrier Integrity===
: '''Green.'''
: '''Green.'''
The inspectors identified a Green NCV of 10 CFR 50.65(b)(1) for Entergy's failure to include a function of a safety-related system within the scope of the maintenance rule program. Specifically, Entergy failed to include the feedwater isolation function performed by the main boiler feedwater pumps (MBFPs) discharge valves, MBFPs, and feedwater regulating valves, which are required to remain functional during and following a design basis event to mitigate the consequence of the accident within the scope of the maintenance rule monitoring program. Entergy initiated corrective actions to include the feedwater isolation function performed by the MBFP discharge valves, MBFPs, and feedwater regulating valves within the maintenance rule monitoring program. Entergy entered this issue into the corrective action program (CAP) as CR-IP2-2016-03963.
The inspectors identified a Green NCV of 10 CFR 50.65(b)(1) for Entergy's failure to include a function of a safety-related system within the scope of the maintenance rule program. Specifically, Entergy failed to include the feedwater isolation function performed by the main boiler feedwater pumps (MBF Ps) discharge valves, MBFPs, and feedwater regulating valves, which are required to remain functional during and following a design basis event to mitigate the consequence of the accident within the scope of the maintenance rule monitoring program. Entergy initiated corrective actions to include the feedwater isolation function performed by the MBFP discharge valves, MBFPs, and feedwater regulating valves within the maintenance rule monitoring program. Entergy entered this issue into the corrective action program (CAP) as CR-IP2-2016-03963.
 
This performance deficiency is more than minor because it was associated with barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to properly scope the feedwater isolation function prevented Entergy from identifying that equipment reliability was no longer effectively controlled through preventive maintenance. In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. This finding does not have a cross-cutting aspect since the failure to scope this equipment into the maintenance rule program was not recognized when Entergy combined the maintenance rule basis documents for Units 2 and 3 in 2012 and, as a result, is not indicative of current licensee performance.  (Section 4OA3)


This performance deficiency is more than minor because it was associated with barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to properly scope the feedwater isolation function prevented Entergy from identifying that equipment reliability was no longer effectively controlled through preventive maintenance. In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. This finding does not have a cross-cutting aspect since the failure to scope this equipment into the maintenance rule program was not recognized when Entergy combined the maintenance rule basis documents for Units 2 and 3 in 2012 and, as a result, is not indicative of current licensee performance.  (Section 4OA3)5
5


=REPORT DETAILS=
=REPORT DETAILS=
Summary of Plant Status Unit 2 began the inspection period during refueling outage (RFO) 2R22 which lasted 102 days.
 
===Summary of Plant Status===
 
Unit 2 began the inspection period during refueling outage (RFO) 2R22 which lasted 102 days.


Upon completion of the outage, the operators restarted Unit 2 on June 14, 2016, and increased power slowly to 93 percent for fuel preconditioning. On June 23, 2016, the operators shutdown the reactor to repair a service water leak on the 21 component cooling water (CCW) heat exchanger (Hx) inlet line and replace switchyard breaker 9. Unit 2 returned to 100 percent power on June 29, 2016. Unit 2 remained at or near 100 percent power for the remainder of the inspection period.
Upon completion of the outage, the operators restarted Unit 2 on June 14, 2016, and increased power slowly to 93 percent for fuel preconditioning. On June 23, 2016, the operators shutdown the reactor to repair a service water leak on the 21 component cooling water (CCW) heat exchanger (Hx) inlet line and replace switchyard breaker 9. Unit 2 returned to 100 percent power on June 29, 2016. Unit 2 remained at or near 100 percent power for the remainder of the inspection period.
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==REACTOR SAFETY==
==REACTOR SAFETY==
Cornerstones:  Initiating Events, Mitigating Systems, and Barrier Integrity 1R04 Equipment Alignment Partial System Walkdowns (71111.04Q - 5 samples)
Cornerstones:  Initiating Events, Mitigating Systems, and Barrier Integrity  
 
{{a|1R04}}
==1R04 Equipment Alignment==
 
Partial System Walkdowns (71111.04Q - 5 samples)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed partial walkdowns of the following systems:  
The inspectors performed partial walkdowns of the following systems:  


Unit 2  Spent fuel pool cooling system following core offload on May 19, 2016  Shutdown cooling system following core reload on June 6, 2016  CCW system following maintenance on June 28, 2016 Unit 3  32 emergency diesel generator (EDG) following maintenance on May 9, 2016 (this sample was part of an in-depth review of the EDG system)  Residual heat removal pumps following CCW system testing on May 20, 2016 The inspectors selected these systems based on their risk-significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the updated final safety analysis report (UFSAR), TSs, work orders (WOs), CRs, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance of their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and were operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. The inspectors also reviewed whether Entergy had properly identified equipment issues and entered them into the CAP for resolution with the appropriate significance characterization. Documents reviewed for each section of this inspection report are listed in the Attachment.
Unit 2  Spent fuel pool cooling system following core offload on May 19, 2016  Shutdown cooling system following core reload on June 6, 2016  CCW system following maintenance on June 28, 2016 Unit 3  32 emergency diesel generator (EDG) following maintenance on May 9, 2016 (this sample was part of an in-depth review of the EDG system)  Residual heat removal pumps following CCW system testing on May 20, 2016 The inspectors selected these systems based on their risk-significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the updated final safety analysis report (UFSAR), TSs, work orders (WOs), CRs, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance of their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and were operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. The inspectors also reviewed whether Entergy had properly identified equipment issues and entered them into the CAP for resolution with the appropriate significance characterization. Documents reviewed for each section of this inspection report are listed in the  
.


====b. Findings====
====b. Findings====
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The inspectors conducted tours of the areas listed below to assess the material condition and operational status of fire protection features. The inspectors verified that Entergy controlled combustible materials and ignition sources in accordance with administrative procedures. The inspectors verified that fire protection and suppression equipment were available for use as specified in the area pre-fire plan (PFP) and passive fire barriers were maintained in good material condition. The inspectors also verified that station personnel implemented compensatory measures for out-of-service (OOS), degraded, or inoperable fire protection equipment, as applicable, in accordance with procedures.
The inspectors conducted tours of the areas listed below to assess the material condition and operational status of fire protection features. The inspectors verified that Entergy controlled combustible materials and ignition sources in accordance with administrative procedures. The inspectors verified that fire protection and suppression equipment were available for use as specified in the area pre-fire plan (PFP) and passive fire barriers were maintained in good material condition. The inspectors also verified that station personnel implemented compensatory measures for out-of-service (OOS), degraded, or inoperable fire protection equipment, as applicable, in accordance with procedures.


Unit 2  Containment, 95-foot elevation, during baffle bolt repair activities with hot work in progress (PFP-203 was reviewed) on June 2, 2016  Residual heat removal pump rooms in primary auxiliary building (PAB), 15-foot elevation (PFP-204 was reviewed), on June 6, 2016  CCW pump room, 68-foot elevation (PFP-209 was reviewed), on June 25, 2016  PAB, 80-foot elevation, CCW heat exchanger area with hot work in progress (PFP-211 was reviewed) on June 25, 2016 Unit 3  32 EDG room, 10-foot elevation (PFP-354 was reviewed), on May 9, 2016  480V switchgear room, 15-foot elevation (PFP-351 was reviewed), on June 30, 2016
Unit 2  Containment, 95-foot elevation, during baffle bolt repair activities with hot work in progress (PFP-203 was reviewed) on June 2, 2016  Residual heat removal pump rooms in primary auxiliary building (PAB), 15-foot elevation (PFP-204 was reviewed), on June 6, 2016  CCW pump room, 68-foot elevation (PFP-209 was reviewed), on June 25, 2016  PAB, 80-foot elevation, CCW heat exchanger area with hot work in progress (PFP-211 was reviewed) on June 25, 2016  
 
Unit 3  32 EDG room, 10-foot elevation (PFP-354 was reviewed), on May 9, 2016  480V switchgear room, 15-foot elevation (PFP-351 was reviewed), on June 30, 2016


====b. Findings====
====b. Findings====
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Inspectors from the NRC Region I Office, specializing in materials and inservice examination activities, observed portions of Entergy's activities involving baffle-former bolt examinations and replacements during Unit 2 RFO 2R22. The inspectors reviewed work documentation and examination procedures and results, and discussed these activities with Entergy. The inspectors were on-site from April 27 to April 28, 2016, and on May 23, 2016. The inspectors verified that Entergy completed baffle-former bolt examinations in accordance with their approved procedures which implemented activities described in the Materials Reliability Program (MRP)-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," as they relate to this component. Specifically, the inspectors reviewed the results of the visual and volumetric examinations of the baffle-former bolts, including capabilities, limitations, and acceptance criteria that were performed during the current RFO.
Inspectors from the NRC Region I Office, specializing in materials and inservice examination activities, observed portions of Entergy's activities involving baffle-former bolt examinations and replacements during Unit 2 RFO 2R22. The inspectors reviewed work documentation and examination procedures and results, and discussed these activities with Entergy. The inspectors were on-site from April 27 to April 28, 2016, and on May 23, 2016. The inspectors verified that Entergy completed baffle-former bolt examinations in accordance with their approved procedures which implemented activities described in the Materials Reliability Program (MRP)-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," as they relate to this component. Specifically, the inspectors reviewed the results of the visual and volumetric examinations of the baffle-former bolts, including capabilities, limitations, and acceptance criteria that were performed during the current RFO.


Non-Destructive Examination Activities   The inspectors reviewed the ultrasonic testing (UT) procedure used for the examination of the Unit 2 baffle-former bolts to verify the procedure was in accordance with the applicable guidance in MRP-227-A and MRP-228. The inspectors reviewed the UT data records and the detailed UT channel analysis for a sample of baffle-former bolts to verify the examinations and evaluations were performed in accordance with approved procedures and applicable guidance. The inspectors reviewed video recordings of the visual examinations of the baffle-former bolts during the current RFO. The inspectors also reviewed recorded video of visual examinations performed in 2006 at Unit 2, completed as part of the existing inservice inspection program for the 10-year reactor vessel examinations, to independently assess the past conditions of the baffle-former bolts and assembly.
Non-Destructive Examination Activities The inspectors reviewed the ultrasonic testing (UT) procedure used for the examination of the Unit 2 baffle-former bolts to verify the procedure was in accordance with the applicable guidance in MRP-227-A and MRP-228. The inspectors reviewed the UT data records and the detailed UT channel analysis for a sample of baffle-former bolts to verify the examinations and evaluations were performed in accordance with approved procedures and applicable guidance. The inspectors reviewed video recordings of the visual examinations of the baffle-former bolts during the current RFO. The inspectors also reviewed recorded video of visual examinations performed in 2006 at Unit 2, completed as part of the existing inservice inspection program for the 10-year reactor vessel examinations, to independently assess the past conditions of the baffle-former bolts and assembly.


The inspectors reviewed certifications of the UT technicians performing the ultrasonic examinations to verify the examinations were performed by qualified individuals and to verify the results were reviewed and evaluated by certified level III non-destructive examination personnel.
The inspectors reviewed certifications of the UT technicians performing the ultrasonic examinations to verify the examinations were performed by qualified individuals and to verify the results were reviewed and evaluated by certified level III non-destructive examination personnel.


Baffle-Former Bolt Replacement Activities The inspectors reviewed the baffle-former bolt replacement activities performed as part of a corrective action to resolve the degraded condition identified at Unit 2. The inspectors observed a sample of in-process bolt removal activities, which included lock bar milling and bolt hole machining. The inspectors reviewed the documentation for in-process and completed bolt installation activities and verified that loose parts generated as part of the bolt replacements were properly tracked. The inspectors verified that bolt replacement activities were performed in accordance with approved procedures. The inspectors also reviewed the Engineering Change (EC) package associated with the new baffle-former bolt design. This review is documented in Section 1R18 of this report. After completion of the bolt replacement activities, the inspectors reviewed the video of the final visual examination of the baffle assembly to verify that the baffle-former bolt work was accomplished as planned and that there were no visual indications of deficiencies.
Baffle-Former Bolt Replacement Activities  
 
The inspectors reviewed the baffle-former bolt replacement activities performed as part of a corrective action to resolve the degraded condition identified at Unit 2. The inspectors observed a sample of in-process bolt removal activities, which included lock bar milling and bolt hole machining. The inspectors reviewed the documentation for in-process and completed bolt installation activities and verified that loose parts generated as part of the bolt replacements were properly tracked. The inspectors verified that bolt replacement activities were performed in accordance with approved procedures. The inspectors also reviewed the Engineering Change (EC) package associated with the new baffle-former bolt design. This review is documented in Section 1R18 of this report. After completion of the bolt replacement activities, the inspectors reviewed the video of the final visual examination of the baffle assembly to verify that the baffle-former bolt work was accomplished as planned and that there were no visual indications of deficiencies.


====b. Findings====
====b. Findings====
No findings were identified.
No findings were identified.


Update to URI 05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies This inspection was conducted to follow-up on NRC Unresolved Item (URI)05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies, to determine whether there was a performance deficiency associated with the degraded baffle-former bolt condition discovered at Unit 2. The inspectors plan to review additional technical information from Entergy as it becomes available, including any revisions to the root cause evaluation. The URI remains open until review of this additional information is completed.  (URI 05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies)  
Update to URI 05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies This inspection was conducted to follow-up on NRC Unresolved Item (URI)05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies, to determine whether there was a performance deficiency associated with the degraded baffle-former bolt condition discovered at Unit 2. The inspectors plan to review additional technical information from Entergy as it becomes available, including any revisions to the root cause evaluation. The URI remains open until review of this additional information is  
 
completed.  (URI 05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies)  
 
{{a|1R11}}
{{a|1R11}}
==1R11 Licensed Operator Requalification Program==
==1R11 Licensed Operator Requalification Program==
{{IP sample|IP=IP 71111.11Q|count=5}}
{{IP sample|IP=IP 71111.11Q|count=5}}
Unit 2
Unit 2


===.1 Quarterly Review of Unit 2 Licensed Operator Requalification Testing and Training===
===.1 Quarterly Review of Unit 2 Licensed Operator Requalification Testing and Training===
{{IP sample|IP=IP 71111.11Q|count=1}}
 
(71111.11Q - 1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed Unit 2 licensed operator simulator training on May 24, 2016, which included reactor coolant pump seal failure with loss of normal heat sink requiring implementation of feed and bleed cooling. The inspectors evaluated operator performance during the simulated event and verified completion of risk significant operator actions, including the use of abnormal and emergency operating procedures.
The inspectors observed Unit 2 licensed operator simulator training on May 24, 2016, which included reactor coolant pump seal failure with loss of normal heat sink requiring implementation of feed and bleed cooling. The inspectors evaluated operator performance during the simulated event and verified completion of risk significant operator actions, including the use of abnormal and emergency operating procedures.


The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the control room supervisor. The inspectors verified the accuracy and timeliness of the emergency classification made by the shift manager and the TS action statements entered by the shift technical advisor. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document crew performance problems.
The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the control room supervisor. The inspectors verified the accuracy and timeliness of the emergency classification made by the shift manager and the TS action statements entered by the shift technical advisor. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document  
 
crew performance problems.


====b. Findings====
====b. Findings====
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===.2 Quarterly Review of Unit 3 Licensed Operator Requalification Testing and Training===
===.2 Quarterly Review of Unit 3 Licensed Operator Requalification Testing and Training===
{{IP sample|IP=IP 71111.11Q|count=1}}
 
(71111.11Q - 1 sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed a Unit 3 licensed operator simulator requalification training evaluated scenario on May 24, 2016, which included failure of a pressurizer pressure instrument, charging pump trip, loss of 480V safety bus 5A, a small break loss-of-coolant accident (LOCA), and entry into FR-C.2 core cooling. The inspectors evaluated operator performance during the simulated event and verified completion of risk significant operator actions, including the use of abnormal and emergency operating procedures.
The inspectors observed a Unit 3 licensed operator simulator requalification training evaluated scenario on May 24, 2016, which included failure of a pressurizer pressure instrument, charging pump trip, loss of 480V safety bus 5A, a small break loss-of-coolant accident (LOCA), and entry into FR-C.2 core cooling. The inspectors evaluated operator performance during the simulated event and verified completion of risk significant operator actions, including the use of abnormal and emergency operating procedures.


The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the control room supervisor. The inspectors verified the accuracy and timeliness of the emergency classification made by the shift manager and the TS action statements entered by the shift technical advisor. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document crew performance problems.
The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the control room supervisor. The inspectors verified the accuracy and timeliness of the emergency classification made by the shift manager and the TS action statements entered by the shift technical advisor. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document  
 
crew performance problems.


====b. Findings====
====b. Findings====
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====b. Findings====
====b. Findings====
No findings were identified.  
No findings were identified.
 
{{a|1R12}}
{{a|1R12}}
==1R12 Maintenance Effectiveness==
==1R12 Maintenance Effectiveness==
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the samples listed below to assess the effectiveness of maintenance activities on SSCs performance and reliability. The inspectors reviewed system health reports, CAP documents, maintenance WOs, and maintenance rule basis documents to ensure that Entergy was identifying and properly evaluating performance problems within the scope of the maintenance rule. For each SSC sample selected, the inspectors verified that the SSC was properly scoped into the maintenance rule in accordance with 10 CFR 50.65 and verified that the (a)(2) performance criteria established by Entergy was reasonable. As applicable, for SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that Entergy was identifying and addressing common cause failures that occurred within and across maintenance rule system boundaries.
The inspectors reviewed the samples listed below to assess the effectiveness of maintenance activities on SSCs performance and reliability. The inspectors reviewed system health reports, CAP documents, maintenance WOs, and maintenance rule basis  
 
documents to ensure that Entergy was identifying and properly evaluating performance problems within the scope of the maintenance rule. For each SSC sample selected, the inspectors verified that the SSC was properly scoped into the maintenance rule in accordance with 10 CFR 50.65 and verified that the (a)(2) performance criteria established by Entergy was reasonable. As applicable, for SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that Entergy was identifying and addressing common cause failures that occurred within and across maintenance rule system boundaries.


Unit 2 EDGs  Unit 3 EDGs (this sample was part of an in-depth review of the EDG system)  Units 2 and 3 CVCS
Unit 2 EDGs  Unit 3 EDGs (this sample was part of an in-depth review of the EDG system)  Units 2 and 3 CVCS
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No findings were identified.
No findings were identified.


URI Opened, CVCS Goal Monitoring Under the Maintenance Rule   Introduction The inspectors identified issues of potential concern with Entergy's application of 10 CFR 50.65(a)(1), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants," (the maintenance rule) in regards to the reliability of the Unit 2 CVCS system. These concerns included the establishment of appropriate (a)(1) goals and whether appropriate justification was established that the corrective actions to address identified maintenance weaknesses were effective prior to removal from (a)(1) status.
URI Opened, CVCS Goal Monitoring Under the Maintenance Rule Introduction The inspectors identified issues of potential concern with Entergy's application of 10 CFR 50.65(a)(1), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants," (the maintenance rule) in regards to the reliability of the Unit 2 CVCS system. These concerns included the establishment of appropriate (a)(1) goals and whether appropriate justification was established that the corrective actions to address identified maintenance weaknesses were effective prior to removal from (a)(1) status.


Specifically, Entergy may have established restrictive goals without defensible justification and may not have demonstrated their chosen goal before ending the goal monitoring interval.
Specifically, Entergy may have established restrictive goals without defensible justification and may not have demonstrated their chosen goal before ending the goal monitoring interval.


Description The maintenance rule requires that licensees shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. EN-DC-206, "Maintenance Rule (a)(1) Process," provides the requirements and processes for managing SSCs for which (a)(2) monitoring has not demonstrated effective maintenance. EN-DC-206 specifies that (a)(1) action plans should not be closed until effectiveness of all corrective actions has been demonstrated by meeting performance goals through the monitoring period (or by other means specified in the action plan).
Description  


Since 2013, there have been several repeat functional failures of equipment in the CVCS resulting in a failure to meet the performance criterion for reliability. These failures included:   A failure of the 23 charging pump on August 6, 2013, after the internal oil pump discharge tubing broke causing the pump to trip on low oil pressure and a loss of charging. The 21 charging pump had tripped for the same reason in 2010. A failure of the 22 charging pump on January 14, 2014, due to cracked internal check valves caused by an inadequate fill-and-vent that left air in the pump following maintenance. The 21 charging pump had failed due to the same cause in 2013. A failure of the Unit 2 valve FCV-110A, boric acid flow control valve, to fully open on January 5, 2015. The valve had insufficient insulation; and as a result, boron crystalized above the valve plug and blocked its movement. The Unit 3 FCV-110A had failed in the same way in 2011, with earlier failures of other valves for the same cause going back to 1997.
The maintenance rule requires that licensees shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. EN-DC-206, "Maintenance Rule (a)(1) Process," provides the requirements and processes for managing SSCs for which (a)(2) monitoring has not demonstrated effective maintenance. EN-DC-206 specifies that (a)(1) action plans should not be closed until effectiveness of all corrective actions has been demonstrated by meeting performance goals through the monitoring period (or by other means specified in the action plan).
 
Since 2013, there have been several repeat functional failures of equipment in the CVCS resulting in a failure to meet the performance criterion for reliability. These failures included:
A failure of the 23 charging pump on August 6, 2013, after the internal oil pump discharge tubing broke causing the pump to trip on low oil pressure and a loss of charging. The 21 charging pump had tripped for the same reason in 2010. A failure of the 22 charging pump on January 14, 2014, due to cracked internal check valves caused by an inadequate fill-and-vent that left air in the pump following maintenance. The 21 charging pump had failed due to the same cause in 2013. A failure of the Unit 2 valve FCV-110A, boric acid flow control valve, to fully open on January 5, 2015. The valve had insufficient insulation; and as a result, boron crystalized above the valve plug and blocked its movement. The Unit 3 FCV-110A had failed in the same way in 2011, with earlier failures of other valves for the same cause going back to 1997.
 
In each case, the CVCS for Unit 2 was already (a)(1), so Entergy either updated the existing (a)(1) action plan or created another one to operate in parallel with the existing one. Upon reviewing the associated (a)(1) action plans, the inspectors noted that in each example Entergy's goals may not have been in accordance with EN-DC-206(a)(1) Process. It specifies that monitoring intervals should be at least six months for normally operating SSCs, at least three surveillances for SSCs monitored by surveillance and long enough to detect recurrence of the applicable failure mechanism. It also states that performance goals that provide reasonable assurance that the SSC is capable of performing its intended functions should be monitored throughout the time the SSC is classified (a)(1). EN-DC-206 defines an SSC as any discreet component grouping that has caused a monitoring failure, including any applicable extent of condition. In the
 
examples provided, NRC inspectors challenged whether Entergy either chose a shorter monitoring interval or a goal that did not include the applicable extent of condition. Specifically:


In each case, the CVCS for Unit 2 was already (a)(1), so Entergy either updated the existing (a)(1) action plan or created another one to operate in parallel with the existing one. Upon reviewing the associated (a)(1) action plans, the inspectors noted that in each example Entergy's goals may not have been in accordance with EN-DC-206(a)(1) Process. It specifies that monitoring intervals should be at least six months for normally operating SSCs, at least three surveillances for SSCs monitored by surveillance and long enough to detect recurrence of the applicable failure mechanism. It also states that performance goals that provide reasonable assurance that the SSC is capable of performing its intended functions should be monitored throughout the time the SSC is classified (a)(1). EN-DC-206 defines an SSC as any discreet component grouping that has caused a monitoring failure, including any applicable extent of condition. In the examples provided, NRC inspectors challenged whether Entergy either chose a shorter monitoring interval or a goal that did not include the applicable extent of condition. Specifically:
The (a)(1) action plan for the broken oil tubing had a goal of no noticeable decrease in 23 charging pump's running oil pressure for the next three quarterly surveillances. The chosen monitoring interval met the procedural expectation, but Entergy limited the monitoring to the 23 charging pump without written justification, when the 21 charging pump had failed previously for the same reason and the other pumps were susceptible to the same failure mechanism.
The (a)(1) action plan for the broken oil tubing had a goal of no noticeable decrease in 23 charging pump's running oil pressure for the next three quarterly surveillances. The chosen monitoring interval met the procedural expectation, but Entergy limited the monitoring to the 23 charging pump without written justification, when the 21 charging pump had failed previously for the same reason and the other pumps were susceptible to the same failure mechanism.


During the monitoring interval, the 21 charging pump experienced low oil pressure. When Entergy performed repairs on the 21 charging pump for an unrelated issue, they discovered that the oil tubing had failed in the same way the 23 charging pump oil tubing had failed, although it had not yet caused a pump trip. The (a)(1) action plan for the cracked check valves had a goal of no check valve failure for six months for the next charging pump that underwent maintenance. This happened to be the 22 charging pump. Entergy chose a six-month monitoring interval, even though only one of the three charging pumps is in service at any given time, and the 22 charging pump only ran for four out of the six months it was monitored. Additionally, the action plan did not justify why a single successful fill-and-vent demonstrated adequate corrective actions. On November 19, 2014, during the six month monitoring interval, the 21 charging pump underwent maintenance requiring a fill-and-vent, and experienced check valve failure two weeks later on December 4. Entergy documented this as a maintenance rule functional failure, and discussed the possibility that it could be due to an inadequate fill-and-vent, but did not change the (a)(1) action plan. The (a)(1) action plan for FCV-110A specified a monitoring interval of six months to include the winter because the previous valve failures had all occurred during the winter months. However, the actual monitoring interval documented in the corrective action was from April to October 2015, and therefore did not cover the winter months as intended. In January 2016, Entergy performed maintenance on valve CH-297 on Unit 3, which is a heat-traced boric acid valve, and did not properly restore the insulation. The valve function was not impacted because it does not often contain high concentrations of boric acid.
During the monitoring interval, the 21 charging pump experienced low oil pressure. When Entergy performed repairs on the 21 charging pump for an unrelated issue, they discovered that the oil tubing had failed in the same way the 23 charging pump oil tubing had failed, although it had not yet caused a pump trip. The (a)(1) action plan for the cracked check valves had a goal of no check valve failure for six months for the next charging pump that underwent maintenance. This happened to be the 22 charging pump. Entergy chose a six-month monitoring  
 
interval, even though only one of the three char ging pumps is in service at any given time, and the 22 charging pump only ran for four out of the six months it was monitored. Additionally, the action plan did not justify why a single successful fill-and-vent demonstrated adequate corrective actions. On November 19, 2014, during the six month monitoring interval, the 21 charging pump underwent maintenance  
 
requiring a fill-and-vent, and experienced check valve failure two weeks later on December 4. Entergy documented this as a maintenance rule functional failure, and discussed the possibility that it could be due to an inadequate fill-and-vent, but did not change the (a)(1) action plan. The (a)(1) action plan for FCV-110A specified a monitoring interval of six months to include the winter because the previous valve failures had all occurred during the winter months. However, the actual monitoring interval documented in the corrective action was from April to October 2015, and therefore did not cover the winter months as intended. In January 2016, Entergy performed maintenance on valve CH-297 on Unit 3, which is a heat-traced boric acid valve, and did not properly restore the insulation. The valve function was not impacted because it does not often contain high concentrations of boric acid.


The (a)(1) action plans described above were all reviewed and approved by the maintenance rule expert panel.
The (a)(1) action plans described above were all reviewed and approved by the maintenance rule expert panel.


Further information regarding the performance of these SSCs is required to determine whether these issues of concern represent performance deficiencies and whether they are more than minor.  (URI 05000247/2016002-01, CVCS Goal Monitoring Under the Maintenance Rule)
Further information regarding the performance of these SSCs is required to determine whether these issues of concern represent performance deficiencies and whether they  
 
are more than minor.  (URI 05000247/2016002-01, CVCS Goal Monitoring Under the Maintenance Rule)


===.2 Quality Control===
===.2 Quality Control===
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====b. Findings====
====b. Findings====
No findings were identified.  
No findings were identified.
 
{{a|1R13}}
{{a|1R13}}
==1R13 Maintenance Risk Assessments and Emergent Work Control==
==1R13 Maintenance Risk Assessments and Emergent Work Control==
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The inspectors reviewed station evaluation and management of plant risk for the maintenance and emergent work activities listed below to verify that Entergy performed the appropriate risk assessments prior to removing equipment for work. The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that Entergy performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When Entergy performed emergent work, the inspectors verified that operations personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work and discussed the results of the assessment with the station's probabilistic risk analyst to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the TS requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.
The inspectors reviewed station evaluation and management of plant risk for the maintenance and emergent work activities listed below to verify that Entergy performed the appropriate risk assessments prior to removing equipment for work. The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that Entergy performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When Entergy performed emergent work, the inspectors verified that operations personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work and discussed the results of the assessment with the station's probabilistic risk analyst to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the TS requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.


Unit 2  Temporary loss of spent fuel pool cooling due to 345-kilovolt disturbance on April 3, 2016  Equipment hatch closure plug seal demonstration for outage risk on April 5, 2016  Reduced inventory operations during vessel reassembly on June 7, 2016  21 CCW heat exchanger OOS during mode 4 on June 25, 2016 Unit 3  32 EDG OOS while Bus Tie BT 4-5 was OOS on May 4, 2016 (this sample was part of an in-depth review of the EDG system)  33 EDG OOS while Bus Tie BT 4-5 was OOS on June 2, 2016  31 EDG OOS while Bus Tie BT 4-5 was OOS on June 21, 2016
Unit 2  Temporary loss of spent fuel pool cooling due to 345-kilovolt disturbance on April 3, 2016  Equipment hatch closure plug seal demonstration for outage risk on April 5, 2016  Reduced inventory operations during vessel reassembly on June 7, 2016  21 CCW heat exchanger OOS during mode 4 on June 25, 2016 Unit 3  32 EDG OOS while Bus Tie BT 4-5 was OOS on May 4, 2016 (this sample was part of an in-depth review of the EDG system)  33 EDG OOS while Bus Tie BT 4-5 was OOS on June 2, 2016  31 EDG OOS while Bus Tie BT 4-5 was OOS on June 21, 2016


====b. Findings====
====b. Findings====
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The inspectors reviewed operability determinations for the following degraded or non-conforming conditions:  
The inspectors reviewed operability determinations for the following degraded or non-conforming conditions:  


Unit 2  23 EDG failure to run on March 7, 2016, and subsequent failure to pass the surveillance test on March 10, 2016, as identified in CR-IP2-2016-01260  Operability determination for N33 gamma metrics wide range nuclear instrument channel in CR-IP2-2016-03660 on June 13, 2016  Pressurizer level transmitter LT-461 reads high in CR-IP-2016-3806 on June 14, 2016  Through-wall leak in line 411, service water inlet to the 21 CCW heat exchanger, on June 15, 2016 Unit 3  Immediate operability determination of the degraded condition of the baffle-former bolts identified from Unit 2 operating experience in CR-IP3-2016-01035 on April 1, 2016  Anomalies noted during digital metal impact monitoring system self-test in CR-IP3-2015-03468 on April 1, 2016  Prompt operability determination of the degraded condition of the baffle-former bolts identified from Unit 2 operating experience in CR-IP2-2016-03660 on June 30, 2016 The inspectors selected these issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the operability determinations to assess whether TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TSs and UFSAR to Entergy's evaluations to determine whether the components or systems were operable.
Unit 2  23 EDG failure to run on March 7, 2016, and subsequent failure to pass the surveillance test on March 10, 2016, as identified in CR-IP2-2016-01260  Operability determination for N33 gamma metrics wide range nuclear instrument channel in CR-IP2-2016-03660 on June 13, 2016  Pressurizer level transmitter LT-461 reads high in CR-IP-2016-3806 on June 14, 2016  Through-wall leak in line 411, service water inlet to the 21 CCW heat exchanger, on June 15, 2016  
 
Unit 3  Immediate operability determination of the degraded condition of the baffle-former bolts identified from Unit 2 operating experience in CR-IP3-2016-01035 on April 1, 2016  Anomalies noted during digital metal impact monitoring system self-test in CR-IP3-2015-03468 on April 1, 2016  Prompt operability determination of the degraded condition of the baffle-former bolts identified from Unit 2 operating experience in CR-IP2-2016-03660 on June 30, 2016 The inspectors selected these issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the operability determinations to assess whether TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TSs and UFSAR to Entergy's evaluations to determine whether the components or systems were operable.


The inspectors confirmed, where appropriate, compliance with bounding limitations associated with the evaluations. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled by Entergy. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations.
The inspectors confirmed, where appropriate, compliance with bounding limitations associated with the evaluations. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled by Entergy. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations.
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The voltage anomaly issues exhibited during the March 10, 2016, test were documented in CR-IP2-2016-01430 which was closed in CR-IP2-2016-01260 to be included in the causal assessment associated with the tripping of 23 EDG breaker on March 7, 2016. Entergy assigned a vendor to perform laboratory bench testing and failure analysis of the 23 EDG AVR card. The vendor report attributed the cause of the March 10, 2016, loss of voltage control to a degraded solder joint on the AVR card. However, the vendor report explicitly did not attribute the event on March 7, 2016, to the same cause.
The voltage anomaly issues exhibited during the March 10, 2016, test were documented in CR-IP2-2016-01430 which was closed in CR-IP2-2016-01260 to be included in the causal assessment associated with the tripping of 23 EDG breaker on March 7, 2016. Entergy assigned a vendor to perform laboratory bench testing and failure analysis of the 23 EDG AVR card. The vendor report attributed the cause of the March 10, 2016, loss of voltage control to a degraded solder joint on the AVR card. However, the vendor report explicitly did not attribute the event on March 7, 2016, to the same cause.


Entergy assigned a corrective action in CR-IP2-2016-01260 to review the cause of the  23 EDG overcurrent trip on March 7, 2016, in light of the vendor report. The inspectors determined that the issue of concern remains open as a URI until this causal assessment has been completed by Entergy and assessed by NRC. (URI 05000247/2016001-06, 23 Emergency Diesel Generator Automatic Voltage Regulator Failure)    
Entergy assigned a corrective action in CR-IP2-2016-01260 to review the cause of the  23 EDG overcurrent trip on March 7, 2016, in light of the vendor report. The inspectors determined that the issue of concern remains open as a URI until this causal assessment has been completed by Entergy and assessed by NRC.
 
(URI 05000247/2016001-06, 23 Emergency Diesel Generator Automatic Voltage Regulator Failure)


{{a|1R18}}
{{a|1R18}}
==1R18 Plant Modifications==
==1R18 Plant Modifications==
{{IP sample|IP=IP 71111.18|count=2}}
{{IP sample|IP=IP 71111.18|count=2}}
Permanent Modifications
Permanent Modifications


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The inspectors reviewed the post-maintenance tests for the maintenance activities listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedure to verify that the procedure adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure was consistent with the information in the applicable licensing basis and/or design basis documents, and that the test results were properly reviewed and accepted and problems were appropriately documented. The inspectors also walked down the affected job site, observed the pre-job brief and post-job critique where possible, confirmed work site cleanliness was maintained, witnessed the test or reviewed test data to verify quality control hold points were performed and checked, and that results adequately demonstrated restoration of the affected safety functions.
The inspectors reviewed the post-maintenance tests for the maintenance activities listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedure to verify that the procedure adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure was consistent with the information in the applicable licensing basis and/or design basis documents, and that the test results were properly reviewed and accepted and problems were appropriately documented. The inspectors also walked down the affected job site, observed the pre-job brief and post-job critique where possible, confirmed work site cleanliness was maintained, witnessed the test or reviewed test data to verify quality control hold points were performed and checked, and that results adequately demonstrated restoration of the affected safety functions.


Unit 2  21 EDG fuel oil transfer pump after planned maintenance on May 5, 2016  Replacement of pressurizer level transmitters LT-459 and LT-460 on May 25, 2016  21 CCW heat exchanger service water outlet weld repair on June 26, 2016  Flux mapping system drive repairs following motor failures on June 28, 2016 Unit 3  Maintenance on service water components associated with the 32 EDG on May 5, 2016 (this sample was part of an in-depth review of the EDG system)  Modification of the 32 EDG space heaters on May 5, 2016 (this sample was part of an in-depth review of the EDG system)  Maintenance on the 32 EDG air start system on May 6, 2016 (this sample was part of an in-depth review of the EDG system)  Replacement of failed bistable LC-427K, steam generator 32 low level mismatch trip interlock, on May 18, 2016
Unit 2  21 EDG fuel oil transfer pump after planned maintenance on May 5, 2016  Replacement of pressurizer level transmitters LT-459 and LT-460 on May 25, 2016  21 CCW heat exchanger service water outlet weld repair on June 26, 2016  Flux mapping system drive repairs following motor failures on June 28, 2016 Unit 3  Maintenance on service water components associated with the 32 EDG on May 5, 2016 (this sample was part of an in-depth review of the EDG system)  Modification of the 32 EDG space heaters on May 5, 2016 (this sample was part of an in-depth review of the EDG system)  Maintenance on the 32 EDG air start system on May 6, 2016 (this sample was part of an in-depth review of the EDG system)  Replacement of failed bistable LC-427K, steam generator 32 low level mismatch trip interlock, on May 18, 2016


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the station's work schedule and outage risk plan for the Unit 2 maintenance during RFO 2R22, which was conducted from March 7, 2016, to June 16, 2016. The inspectors reviewed Entergy's development and implementation of outage plans and schedules to verify that risk, industry experience, previous site-specific problems, and defense-in-depth were considered. During the outage, the inspectors observed portions of the shutdown and cooldown processes and monitored controls associated with the following outage activities:   Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TSs when taking equipment OOS  Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work or testing  Installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication and instrument error accounting  Status and configuration of electrical systems and switchyard activities to ensure that TSs were met  Monitoring of decay heat removal operations  Impact of outage work on the ability of the operators to operate the spent fuel pool cooling system  Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss  Activities that could affect reactivity  Maintenance of secondary containment as required by TSs  Refueling activities, including fuel handling and fuel receipt inspections  Fatigue management  Tracking of startup prerequisites, walkdown of the primary containment to verify that debris had not been left which could block the ECCS suction strainers, and startup and ascension to full power operation  Foreign Object Search and Retrieval for missing baffle bolts and locking tabs  Identification and resolution of problems related to RFO activities During this outage, Entergy replaced 278 degraded baffle bolts in the Unit 2 reactor vessel baffle assembly. This emergent project resulted in the extension of the outage schedule from 30 days to 102 days.
The inspectors reviewed the station's work schedule and outage risk plan for the Unit 2 maintenance during RFO 2R22, which was conducted from March 7, 2016, to June 16, 2016. The inspectors reviewed Entergy's development and implementation of outage plans and schedules to verify that risk, industry experience, previous site-specific problems, and defense-in-depth were considered. During the outage, the inspectors observed portions of the shutdown and cooldown processes and monitored controls associated with the following outage activities:
Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TSs when taking equipment OOS  Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work or testing  Installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication and instrument error accounting  Status and configuration of electrical systems and switchyard activities to ensure that TSs were met  Monitoring of decay heat removal operations  Impact of outage work on the ability of the operators to operate the spent fuel pool cooling system  Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss  Activities that could affect reactivity  Maintenance of secondary containment as required by TSs  Refueling activities, including fuel handling and fuel receipt inspections  Fatigue management  Tracking of startup prerequisites, walkdown of the primary containment to verify that debris had not been left which could block the ECCS suction strainers, and startup and ascension to full power operation  Foreign Object Search and Retrieval for missing baffle bolts and locking tabs  Identification and resolution of problems related to RFO activities During this outage, Entergy replaced 278 degraded baffle bolts in the Unit 2 reactor vessel baffle assembly. This emergent project resulted in the extension of the outage schedule from 30 days to 102 days.


====b. Findings====
====b. Findings====
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=====Analysis.=====
=====Analysis.=====
The inspectors determined that Energy's failure to maintain either gate 17 or gate 23 closed during passage in accordance with OAP-007 was a performance deficiency. The performance deficiency was more than minor because it is associated with the configuration control (shutdown equipment lineup) attribute and adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated the finding in accordance with IMC 0609, Appendix G, Attachment 1, Exhibit 3, and determined that a detailed risk evaluation was necessary because the finding represented a loss of system safety function. A detailed risk assessment was conducted conservatively assuming complete failure of the recirculation and containment sumps due to the performance deficiency. Given that Unit 2 was in mode 4, in plant operating state 1, with a late time window, the at-power simplified plant analysis risk model for large-break LOCAs was determined to best model the degrade condition and plant response. An exposure time of one day was assumed. No credit was assumed for the decrease in energy that would be anticipated in a release during a LOCA in mode 4, nor the corresponding reduction in debris generation. This was also considered conservative. Utilizing Systems Analysis Program for Hands-On Integrated Reliability Evaluation, version 8.13, with Indian Point Unit 2 Simplified Plant Analysis Risk Model, version 8.19, for the assumed conditions, the change in core damage frequency was determined to be 7E-9. Therefore, this issue represents a Green finding.
The inspectors determined that Energy's failure to maintain either gate 17 or gate 23 closed during passage in accordance with OAP-007 was a performance deficiency. The performance deficiency was more than minor because it is associated with the configuration control (shutdown equipment lineup) attribute and adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated the finding in accordance with IMC 0609, Appendix G, Attachment 1, Exhibit 3, and determined that a detailed risk evaluation was necessary because the finding represented a loss of system safety function. A detailed risk assessm ent was conducted conservatively assuming complete failure of the recirculation and containment sumps due to the performance deficiency. Given that Unit 2 was in mode 4, in plant operating state 1, with a late time window, the at-power simplified plant analysis risk model for large-break LOCAs was determined to best model the degrade condition and plant response. An exposure time of one day was assumed. No credit was assumed for the decrease in energy that would be anticipated in a release during a LOCA in mode 4, nor the corresponding reduction in debris generation. This was also considered conservative. Utilizing Systems Analysis Program for Hands-On Integrated Reliability Evaluation, version 8.13, with Indian Point Unit 2 Simplified Plant Analysis Risk Model, version 8.19, for the assumed conditions, the change in core damage frequency was determined to be 7E-9. Therefore, this issue represents a Green finding.


This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Entergy did not consider potential undesired consequences of actions before performing work and implement appropriate error-reduction tools. Specifically, the work crew did not understand the requirements and potential consequences prior to commencing work and the gate monitor did not enforce these requirements to maintain at least one gate locked or pinned closed as required by OAP-007.  [H.12]  
This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Entergy did not consider potential undesired consequences of actions before performing work and implement appropriate error-reduction tools. Specifically, the work crew did not understand the requirements and potential consequences prior to commencing work and the gate monitor did not enforce these requirements to maintain at least one gate locked or pinned closed as required by OAP-007.  [H.12]  
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====a. Inspection Scope====
====a. Inspection Scope====
Unit 2 conducted a forced outage from June 24 to 27, 2016, in order to complete weld repairs on a through-wall leak on the service water inlet line to the 21 CCW heat exchanger. These repairs required shutting down to mode 4 in order to meet the TS 3.7.7, "Component Cooling Water (CCW) System," limiting condition for operations for CCW operability. While these repairs were being completed, the grid operator completed repairs to breaker 9 in the offsite switchyard. During the outage, the inspectors observed portions of the shutdown and cooldown processes and monitored controls associated with the following outage activities:   Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TSs when taking equipment OOS  Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work or testing  Status and configuration of electrical systems and switchyard activities to ensure that TSs were met  Monitoring of decay heat removal operations  Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss  Activities that could affect reactivity  Tracking of startup prerequisites   Identification and resolution of problems related to RFO activities When all repairs had been completed, Entergy restarted Unit 2 on June 27, 2016.
Unit 2 conducted a forced outage from June 24 to 27, 2016, in order to complete weld repairs on a through-wall leak on the service water inlet line to the 21 CCW heat exchanger. These repairs required shutting down to mode 4 in order to meet the TS 3.7.7, "Component Cooling Water (CCW) System," limiting condition for operations for CCW operability. While these repairs were being completed, the grid operator completed repairs to breaker 9 in the offsite switchyard. During the outage, the inspectors observed portions of the shutdown and cooldown processes and monitored controls associated with the following outage activities:
Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TSs when taking equipment OOS  Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work or testing  Status and configuration of electrical systems and switchyard activities to ensure that TSs were met  Monitoring of decay heat removal operations  Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss  Activities that could affect reactivity  Tracking of startup prerequisites Identification and resolution of problems related to RFO activities When all repairs had been completed, Entergy restarted Unit 2 on June 27, 2016.


====b. Findings====
====b. Findings====
No findings were identified.
No findings were identified.
{{a|1R22}}
==1R22 Surveillance Testing (71111.22==


{{a|1R22}}
- 6 samples)
==1R22 Surveillance Testing==
{{IP sample|IP=IP 71111.22|count=6}}


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed performance of surveillance tests and/or reviewed test data of selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR, and Entergy's procedure requirements. The inspectors verified that test acceptance criteria were clear, tests demonstrated operational readiness and were consistent with design documentation, test instrumentation had current calibrations and the range and accuracy for the application, tests were performed as written, and applicable test prerequisites were satisfied. Upon test completion, the inspectors considered whether the test results supported that equipment was capable of performing the required safety functions. The inspectors reviewed the following surveillance tests:
The inspectors observed performance of surveillance tests and/or reviewed test data of selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR, and Entergy's procedure requirements. The inspectors verified that test acceptance criteria were clear, tests demonstrated operational readiness and were consistent with design documentation, test instrumentation had current calibrations and the range and accuracy for the application, tests were performed as written, and applicable test prerequisites were satisfied. Upon test completion, the inspectors considered whether the test results supported that equipment was capable of performing the required safety functions. The inspectors reviewed the following surveillance tests:  
 
Unit 2  WO 446385, 21 EDG AVR card inspection, on May 24, 2016  2-PT-Q013 for containment isolation valve 851B (22 safety injection (SI) pump tie to 23 SI pump discharge) on June 6, 2016  2-PT-Q029B quarterly in-service surveillance test for the 22 SI pump on June 6, 2016 Unit 3  3-PT-M079B 32 EDG monthly surveillance on May 30, 2016 (this sample was part of an in-depth review of the EDG system)  34 steam generator pressure instrument channel check on June 21, 2016  0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation and Leak Identification, beginning on June 28, 2016
Unit 2  WO 446385, 21 EDG AVR card inspection, on May 24, 2016  2-PT-Q013 for containment isolation valve 851B (22 safety injection (SI) pump tie to 23 SI pump discharge) on June 6, 2016  2-PT-Q029B quarterly in-service surveillance test for the 22 SI pump on June 6, 2016 Unit 3  3-PT-M079B 32 EDG monthly surveillance on May 30, 2016 (this sample was part of an in-depth review of the EDG system)  34 steam generator pressure instrument channel check on June 21, 2016  0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation and Leak Identification, beginning on June 28, 2016


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===Cornerstone:===
===Cornerstone:===
Emergency Preparedness  
Emergency Preparedness  
{{a|1EP6}}
{{a|1EP6}}
==1EP6 Drill Evaluation==
==1EP6 Drill Evaluation==
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During May 10-12 and June 13-17, 2016, the inspectors reviewed Entergy's performance in assessing the radiological hazards and exposure control in the workplace. The inspectors used the requirements in 10 CFR 20, TSs, applicable industry standards, and procedures required by TSs as criteria for determining compliance.
During May 10-12 and June 13-17, 2016, the inspectors reviewed Entergy's performance in assessing the radiological hazards and exposure control in the workplace. The inspectors used the requirements in 10 CFR 20, TSs, applicable industry standards, and procedures required by TSs as criteria for determining compliance.


Radiological Hazards Control and Work Coverage The inspectors reviewed:
Radiological Hazards Control and Work Coverage The inspectors reviewed:  
 
Ambient radiological conditions during tours of the radiological controlled area, posted surveys, radiation work permits, adequacy of radiological controls, radiation protection job coverage, and contamination controls  Controls for highly activated or contaminated materials stored within spent fuel pools  Posting and physical controls for high radiation areas and very high radiation areas
Ambient radiological conditions during tours of the radiological controlled area, posted surveys, radiation work permits, adequacy of radiological controls, radiation protection job coverage, and contamination controls  Controls for highly activated or contaminated materials stored within spent fuel pools  Posting and physical controls for high radiation areas and very high radiation areas


====b. Findings====
====b. Findings====
No findings were identified.
No findings were identified.
{{a|2RS2}}
{{a|2RS2}}
==2RS2 Occupational As Low As Is Reasonably Achievable (ALARA) Planning and Controls==
==2RS2 Occupational As Low As Is Reasonably Achievable (ALARA) Planning and Controls==
{{IP sample|IP=IP 71124.02}}
 
(71124.02)


====a. Inspection Scope====
====a. Inspection Scope====
During May 10-12 and June 13-17, 2016, the inspectors assessed performance with respect to maintaining occupational individual and collective radiation exposures ALARA. The inspectors used the requirements in 10 CFR 20, TSs, applicable industry standards, and procedures required by TSs as criteria for determining compliance.
During May 10-12 and June 13-17, 2016, the inspectors assessed performance with respect to maintaining occupational individual and collective radiation exposures ALARA. The inspectors used the requirements in 10 CFR 20, TSs, applicable industry standards, and procedures required by TSs as criteria for determining compliance.


Radiological Work Planning The inspectors reviewed:   ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements  ALARA work planning, use of dose mitigation features and dose goals  Work planning and the integration of ALARA requirements
Radiological Work Planning The inspectors reviewed:
ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements  ALARA work planning, use of dose mitigation features and dose goals  Work planning and the integration of ALARA requirements


====b. Findings====
====b. Findings====
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The inspectors reviewed Entergy's 2014 and 2015 annual radiological environmental and effluent monitoring reports, REMP program audits, ODCM changes, land use census, the UFSAR, and inter-laboratory comparison program results.
The inspectors reviewed Entergy's 2014 and 2015 annual radiological environmental and effluent monitoring reports, REMP program audits, ODCM changes, land use census, the UFSAR, and inter-laboratory comparison program results.


Site Inspection   The inspectors walked down various thermoluminescent dosimeter and air and water sampling locations and reviewed associated calibration and maintenance records. The inspectors observed the sampling of various environmental media as specified in the ODCM and reviewed any anomalous environmental sampling events including assessment of any positive radioactivity results. The inspectors reviewed any changes to the ODCM. The inspectors verified the operability and calibration of the meteorological tower instruments and meteorological data readouts. The inspectors reviewed environmental sample laboratory analysis results, laboratory instrument measurement detection sensitivities, laboratory quality control program audit results, and the inter- and intra-laboratory comparison program results. The inspectors reviewed the groundwater monitoring program as it applies to selected potential leaking SSCs. GPI Implementation  The inspectors reviewed groundwater monitoring results, changes to the GPI program since the last inspection, anomalous results or missed groundwater samples, leakage or spill events including entries made into the decommissioning files (10 CFR 50.75(g)), evaluations of surface water discharges, and Entergy's evaluation of any positive groundwater sample results including appropriate stakeholder notifications and effluent reporting requirements.
Site Inspection The inspectors walked down various thermoluminescent dosimeter and air and water sampling locations and reviewed associated calibration and maintenance records. The inspectors observed the sampling of various environmental media as specified in the ODCM and reviewed any anomalous environmental sampling events including assessment of any positive radioactivity results. The inspectors reviewed any changes to the ODCM. The inspectors verified the operability and calibration of the meteorological tower instruments and meteorological data readouts. The inspectors reviewed environmental sample laboratory analysis results, laboratory instrument measurement detection sensitivities, laboratory quality control program audit results, and the inter- and intra-laboratory comparison program results. The inspectors reviewed the groundwater monitoring program as it applies to selected potential leaking SSCs.
 
GPI Implementation  The inspectors reviewed groundwater monitoring results, changes to the GPI program since the last inspection, anomalous results or missed groundwater samples, leakage or spill events including entries made into the decommissioning files (10 CFR 50.75(g)), evaluations of surface water discharges, and Entergy's evaluation of any positive groundwater sample results including appropriate stakeholder notifications and effluent reporting requirements.
 
Identification and Resolution of Problems The inspectors evaluated whether problems associated with the REMP were identified at


Identification and Resolution of Problems  The inspectors evaluated whether problems associated with the REMP were identified at an appropriate threshold and properly addressed in Entergy's CAP.
an appropriate threshold and properly addressed in Entergy's CAP.


====b. Findings====
====b. Findings====
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==4OA1 Performance Indicator Verification==
==4OA1 Performance Indicator Verification==
{{IP sample|IP=IP 71151|count=6}}
{{IP sample|IP=IP 71151|count=6}}
Initiating Events Performance Indicators
Initiating Events Performance Indicators


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The inspectors reviewed Entergy's submittals for the following Initiating Events cornerstone performance indicators for the period April 1, 2015, to March 31, 2016:  
The inspectors reviewed Entergy's submittals for the following Initiating Events cornerstone performance indicators for the period April 1, 2015, to March 31, 2016:  


Unit 2  Unplanned scrams per 7000 critical hours (IE01)  Unplanned power changes per 7000 critical hours (IE03)  Unplanned scrams with complications (IE04) Unit 3  Unplanned scrams (IE01)  Unplanned power changes (IE03)  Unplanned scrams with complications (IE04) To determine the accuracy of the performance indicator data reported during those periods, inspectors used definitions and guidance contained in Nuclear Energy Institute 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7.
Unit 2  Unplanned scrams per 7000 critical hours (IE01)  Unplanned power changes per 7000 critical hours (IE03)  Unplanned scrams with complications (IE04)
Unit 3  Unplanned scrams (IE01)  Unplanned power changes (IE03)  Unplanned scrams with complications (IE04)
To determine the accuracy of the performance indicator data reported during those periods, inspectors used definitions and guidance contained in Nuclear Energy Institute 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7.


The inspectors reviewed Entergy's operator narrative logs, maintenance planning schedules, CRs, event reports, and NRC integrated inspection reports to validate the accuracy of the submittals. There were no unplanned power changes or scrams with complications during the review period.
The inspectors reviewed Entergy's operator narrative logs, maintenance planning schedules, CRs, event reports, and NRC integrated inspection reports to validate the accuracy of the submittals. There were no unplanned power changes or scrams with complications during the review period.
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====a. Inspection Scope====
====a. Inspection Scope====
As required by Inspection Procedure 71152, "Problem Identification and Resolution," the inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that Entergy entered issues into the CAP at an appropriate threshold, gave adequate attention to timely corrective actions, and identified and addressed adverse trends. In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow up, the inspectors performed a daily screening of items entered into the CAP and periodically attended CR screening meetings. The inspectors also confirmed, on a sampling basis, that, as applicable, for identified defects and non-conformances, Entergy performed an evaluation in accordance with 10 CFR 21.
As required by Inspection Procedure 71152, "Problem Identification and Resolution," the  
 
inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that Entergy entered issues into the CAP at an appropriate threshold, gave adequate attention to timely corrective actions, and identified and addressed adverse trends. In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow up, the inspectors performed a daily screening of items entered into the CAP and periodically attended CR screening meetings. The inspectors also confirmed, on a sampling basis, that, as applicable, for identified defects and non-conformances, Entergy performed an evaluation in accordance with 10 CFR 21.


====b. Findings====
====b. Findings====
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====b. Findings and Observations====
====b. Findings and Observations====
No findings were identified. The inspectors identified a trend in work being performed that was contrary to written work instructions and procedures, and work packages had been closed out without documenting the deviation from the work order. While reviewing completed work order WO 447966 Task 10, Internal Coating Repair for 21 CCW HX, the inspectors found a note in the work order stating that the internal coating repair to the pipe had not been done in accordance with the engineering change. The engineering change had been written when the coating repair was expected to be small, but the actual area that was recoated was much larger. A larger area of coating increases the impact on the heat exchanger if the coating were to flake off and block the flow of service water. The work package was closed and no condition report was written. This performance deficiency is minor because the coating was applied with procedurally directed quality controls and the likelihood that it would flake off is very small; and is the same as the original smaller area specified in the work package. However, the work package was closed without documenting the deviation and no CR was written.
No findings were identified.
 
The inspectors identified a trend in work being performed that was contrary to written work instructions and procedures, and work packages had been closed out without  
 
documenting the deviation from the work order. While reviewing completed work order WO 447966 Task 10, Internal Coating Repair for 21 CCW HX, the inspectors found a note in the work order stating that the internal coating repair to the pipe had not been done in accordance with the engineering change. The engineering change had been written when the coating repair was expected to be small, but the actual area that was recoated was much larger. A larger area of coating increases the impact on the heat exchanger if the coating were to flake off and block the flow of service water. The work package was closed and no condition report was written. This performance deficiency is minor because the coating was applied with procedurally directed quality controls and the likelihood that it would flake off is very small; and is the same as the original smaller area specified in the work package. However, the work package was closed without documenting the deviation and no CR was written.


In another example, the inspectors noted that WO 412920 Task 15 to perform a surge test on 11 centrifugal air compressor (CENTAC) after overhaul was completed on December 22, 2015. However, the completion notes and documentation for the task showed that the test was unable to be performed due to a test equipment problem. The work package was closed and no CR was written. Subsequently, after being returned to service, the compressor failed in service due to multiple surging events on January 7, 2016. Troubleshooting under WO 433939 revealed that the motor high load limit had not been adjusted to account for the increased load due to reduced compressor clearances introduced by the overhaul. This performance deficiency is screened to minor because the 11 CENTAC is not a safety-related piece of equipment and would not affect the MC 0609 cornerstone thresholds or other generic criteria. Unit 2 and Unit 3 have dedicated instrument air compressors that are credited in the FSAR to respond to a loss of instrument air event. If the 11 CENTAC (located in Unit 1) were to fail, the unit-specific IACs would automatically start to prevent a loss of instrument air at both Units 2 and 3.
In another example, the inspectors noted that WO 412920 Task 15 to perform a surge test on 11 centrifugal air compressor (CENTAC) after overhaul was completed on December 22, 2015. However, the completion notes and documentation for the task showed that the test was unable to be performed due to a test equipment problem. The work package was closed and no CR was written. Subsequently, after being returned to service, the compressor failed in service due to multiple surging events on January 7, 2016. Troubleshooting under WO 433939 revealed that the motor high load limit had not been adjusted to account for the increased load due to reduced compressor clearances introduced by the overhaul. This performance deficiency is screened to minor because the 11 CENTAC is not a safety-related piece of equipment and would not affect the MC 0609 cornerstone thresholds or other generic criteria. Unit 2 and Unit 3 have dedicated instrument air compressors that are credited in the FSAR to respond to a loss of instrument air event. If the 11 CENTAC (located in Unit 1) were to fail, the unit-specific IACs would automatically start to prevent a loss of instrument air at both Units 2 and 3.
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The inspectors performed an in-depth review of Entergy's corrective actions associated with self-assessment LO-IP3LO-2015-72, "Maintenance Rule (a)(3) Assessment."  The self-assessment was performed to satisfy both the self-assessment criteria in EN-LI-104, "Self-Assessment and Benchmark Process," and the maintenance rule periodic assessment criteria in EN-DC-207.
The inspectors performed an in-depth review of Entergy's corrective actions associated with self-assessment LO-IP3LO-2015-72, "Maintenance Rule (a)(3) Assessment."  The self-assessment was performed to satisfy both the self-assessment criteria in EN-LI-104, "Self-Assessment and Benchmark Process," and the maintenance rule periodic assessment criteria in EN-DC-207.


The inspectors assessed Entergy's problem identification threshold, extent of condition reviews, and the prioritization and timeliness of Entergy corrective actions to determine whether Entergy was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of Entergy's CAP and 10 CFR 50, Appendix B. In addition, the inspectors interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.
The inspectors assessed Entergy's problem identification threshold, extent of condition reviews, and the prioritization and timeliness of Entergy corrective actions to determine whether Entergy was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of Entergy's CAP and 10 CFR 50, Appendix B. In addition, the inspectors interviewed engineering personnel to assess the effe ctiveness of the im plemented corrective actions.


====b. Findings and Observations====
====b. Findings and Observations====
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed an in-depth review of Entergy's root cause evaluation and corrective actions associated with CR-IP2-2016-02348 for baffle-former ("baffle") bolts found with indications of degradation during the Indian Point Unit 2 RFO 2R22. Entergy performed ultrasonic examinations of the baffle bolts in accordance with their procedures as part of a planned activity. After an unexpected number of degraded baffle bolts were discovered, Entergy staff reported the issue to the NRC as Event Notification No. 51829 on March 29, 2016, because the as-found number and location of degraded bolts represented an unanalyzed condition. Entergy staff completed corrective actions to replace all of the potentially degraded baffle bolts on Unit 2. Entergy staff further replaced a population of additional bolts that exhibited no indications of degradation and performed an evaluation to determine the potential for baffle bolt failures at Unit 3. The baffle-former bolts help secure vertical plates (also referred to as baffle plates)inside the reactor vessel, which then forms a structure surrounding the reactor fuel assemblies to orient the fuel and to direct coolant flow through the core. A sufficient number of baffle bolts are required to remain intact to secure the baffle plates in place so as to not affect control rod insertion or impede emergency core cooling flow during postulated accident conditions. Bolt heads that separate and are no longer held in place by bolt lock-tabs can also become a loose parts concern.
The inspectors performed an in-depth review of Entergy's root cause evaluation and corrective actions associated with CR-IP2-2016-02348 for baffle-former ("baffle") bolts found with indications of degradation during the Indian Point Unit 2 RFO 2R22. Entergy performed ultrasonic examinations of the baffle bolts in accordance with their procedures as part of a planned activity. After an unexpected number of degraded baffle bolts were discovered, Entergy staff reported the issue to the NRC as Event Notification No. 51829 on March 29, 2016, because the as-found number and location of degraded bolts represented an unanalyzed condition. Entergy staff completed corrective actions to replace all of the potentially degraded baffle bolts on Unit 2. Entergy staff further replaced a population of additional bolts that exhibited no indications of degradation and performed an evaluation to determine the potential for baffle bolt failures at Unit 3.
 
The baffle-former bolts help secure vertical plates (also referred to as baffle plates)inside the reactor vessel, which then forms a structure surrounding the reactor fuel assemblies to orient the fuel and to direct coolant flow through the core. A sufficient number of baffle bolts are required to remain intact to secure the baffle plates in place so as to not affect control rod insertion or impede emergency core cooling flow during postulated accident conditions. Bolt heads that separate and are no longer held in place by bolt lock-tabs can also become a loose parts concern.


The inspectors determined whether Entergy's acceptable baffle bolt pattern analysis for Unit 2 was completed in accordance with the NRC-approved methodology and provided appropriate structural margin for the next cycle of operation to ensure the Unit 2 baffle plates will remain in place during both normal operation and limiting postulated accident conditions. The inspectors further determined whether Entergy's evaluations of the baffle bolts installed in Indian Point Unit 3 were technically sufficient to conclude the Unit 3 baffle assembly will perform as intended until the next planned RFO, at which time Entergy plans to examine the bolts. The inspectors reviewed Entergy's procedures for determining the functionality and operability of degraded SSC as they relate to Unit 3. The inspectors further interviewed Entergy engineering personnel and contractor staff to discuss the results of Entergy's technical evaluations and to assess the effectiveness of the implemented and planned corrective actions.
The inspectors determined whether Entergy's acceptable baffle bolt pattern analysis for Unit 2 was completed in accordance with the NRC-approved methodology and provided appropriate structural margin for the next cycle of operation to ensure the Unit 2 baffle plates will remain in place during both normal operation and limiting postulated accident conditions. The inspectors further determined whether Entergy's evaluations of the baffle bolts installed in Indian Point Unit 3 were technically sufficient to conclude the Unit 3 baffle assembly will perform as intended until the next planned RFO, at which time Entergy plans to examine the bolts. The inspectors reviewed Entergy's procedures for determining the functionality and operability of degraded SSC as they relate to Unit 3. The inspectors further interviewed Entergy engineering personnel and contractor staff to discuss the results of Entergy's technical evaluations and to assess the effectiveness of the implemented and planned corrective actions.
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====b. Findings and Observations====
====b. Findings and Observations====
The NRC responded to the initial discovery of an unexpected number of baffle bolts found degraded at Indian Point Unit 2 by implementing a comprehensive inspection plan consisting of various baseline inspection samples to assess the extent of the issue and to determine the necessary NRC actions. A follow-up inservice inspection sample (Refer to Section 1R08) was conducted to review the capability of the non-destructive examination techniques, evaluate the UT results, and observe a portion of bolt replacement activities on-site. A permanent modification sample (Refer to Section 1R18) was conducted to review the design change package and evaluations associated with the new, replacement baffle bolts. The NRC resident inspectors reviewed Entergy's foreign material controls and loose parts analysis (Refer to Section 1R20) to address the potential for missing bolt heads and concluded it would not impact safe operation of the plant.
The NRC responded to the initial discovery of an unexpected number of baffle bolts found degraded at Indian Point Unit 2 by implementing a comprehensive inspection plan consisting of various baseline inspection samples to assess the extent of the issue and to determine the necessary NRC actions. A follow-up inservice inspection sample (Refer to Section 1R08) was conducted to review the capability of the non-destructive examination techniques, evaluate the UT results, and observe a portion of bolt replacement activities on-site. A permanent modification sample (Refer to Section 1R18) was conducted to review the design change package and evaluations associated  
 
with the new, replacement baffle bolts. The NRC resident inspectors reviewed Entergy's foreign material controls and loose parts analysis (Refer to Section 1R20) to address the potential for missing bolt heads and concluded it would not impact safe operation of the plant.


NRC Region I based inspectors accompanied by an expert from the NRC Office of Nuclear Reactor Regulation completed an annual problem identification and resolution inspection, documented in this section of the report, to verify that Entergy's evaluations and corrective actions to replace Unit 2 baffle bolts were completed in accordance with an NRC approved methodology to support a conclusion that the Unit 2 baffle assembly meets the plant design basis. The inspectors also determined the adequacy of Entergy's evaluations completed to determine there is a reasonable expectation that the Unit 3 baffle assembly will perform as intended during the current operating cycle.
NRC Region I based inspectors accompanied by an expert from the NRC Office of Nuclear Reactor Regulation completed an annual problem identification and resolution inspection, documented in this section of the report, to verify that Entergy's evaluations and corrective actions to replace Unit 2 baffle bolts were completed in accordance with an NRC approved methodology to support a conclusion that the Unit 2 baffle assembly meets the plant design basis. The inspectors also determined the adequacy of Entergy's evaluations completed to determine there is a reasonable expectation that the Unit 3 baffle assembly will perform as intended during the current operating cycle.
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As a corrective action, Entergy staff performed an operability evaluation and demonstrated that the Unit 3 baffle former assembly remained operable. The inspectors concluded that this supplemental evaluation provided appropriate technical justification for the continued operation of Unit 3 until the next RFO in spring 2017, at which time Entergy plans to examine the baffle bolts. Entergy also implemented a corrective action as part of an enhancement to plant operations to monitor the RCS for any signs of fuel leakage, which could be an indicator of baffle bolt failures.
As a corrective action, Entergy staff performed an operability evaluation and demonstrated that the Unit 3 baffle former assembly remained operable. The inspectors concluded that this supplemental evaluation provided appropriate technical justification for the continued operation of Unit 3 until the next RFO in spring 2017, at which time Entergy plans to examine the baffle bolts. Entergy also implemented a corrective action as part of an enhancement to plant operations to monitor the RCS for any signs of fuel leakage, which could be an indicator of baffle bolt failures.


The inspectors reviewed Westinghouse Nuclear Safety Advisory Letter NSAL-16-1, which discussed the results of recent baffle-former bolt inspections and provided Westinghouse's recommendations on this issue. The letter described the plants as most susceptible (i.e. Tier 1a) to this degradation as Westinghouse 4-loop reactors limited to those with a down-flow configuration and using Type 347 stainless steel bolts. The inspectors noted the recommendation was to complete UT volumetric examination of the baffle bolts at the next scheduled RFO, and that Entergy had already planned this action for Unit 3. Entergy also planned a long-term corrective action to convert Units 2 and 3 from a "down-flow" baffle configuration to an "up-flow" configuration, which would significantly reduce the load on baffle-former bolts and provide for increased structural margin of the baffle-former assembly. The inspectors determined Entergy's overall response to the issue was commensurate with the safety significance, was timely, and included appropriate compensatory actions. The inspectors concluded that the actions completed and planned were reasonable to address the ongoing aging management of baffle bolts.
The inspectors reviewed Westinghouse Nuclear Safety Advisory Letter NSAL-16-1, which discussed the results of recent baffle-former bolt inspections and provided Westinghouse's recommendations on this issue. The letter described the plants as most susceptible (i.e. Tier 1a) to this degradation as Westinghouse 4-loop reactors limited to those with a down-flow configuration and using Type 347 stainless steel bolts. The inspectors noted the recommendation was to complete UT volumetric examination of the baffle bolts at the next scheduled RFO, and that Entergy had already planned this action for Unit 3. Entergy also planned a long-term corrective action to convert Units 2 and 3 from a "down-flow" baffle configuration to an "up-flow" configuration, which would  
 
significantly reduce the load on baffle-former bolts and provide for increased structural margin of the baffle-former assembly. The inspectors determined Entergy's overall response to the issue was commensurate with the safety significance, was timely, and included appropriate compensatory actions. The inspectors concluded that the actions completed and planned were reasonable to address the ongoing aging management of baffle bolts.


{{a|4OA3}}
{{a|4OA3}}
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===.2 (Closed) Licensee Event Report (LER) 05000247/2015-003-00:===
===.2 (Closed) Licensee Event Report (LER) 05000247/2015-003-00:===
Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure The inspector's reviewed Entergy's actions and reportability criteria associated with LER 05000247/2015-003-00, which was submitted to the NRC on February 3, 2016. On December 5, 2015, control room operators initiated a manual reactor trip after observing indications consistent with multiple dropped control rods following an alarm for the trip of MCC-24/24A. No control rod indication was available due to MCC-24 being faulted and de-energized. The direct cause of the event was the loss of MCC-24 due to an internal fault at the line sides leads at cubicle 2H where they connect to the bucket stab assemblies. The apparent cause was an unanticipated loss of power to the control rod system due to the degradation of the primary control rod PS1 which failed to function when the operating PS2 was lost. The inspectors determined that both the unexpected failure of PS2 and the internal fault in PS1 was not within Entergy's ability to foresee and prevent and was not a performance deficiency. The inspectors reviewed the LER, the associated apparent cause evaluation analysis, and interviewed Entergy staff. This LER is closed.
Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure  
 
The inspector's reviewed Entergy's actions and reportability criteria associated with LER 05000247/2015-003-00, which was submitted to the NRC on February 3, 2016. On December 5, 2015, control room operators initiated a manual reactor trip after observing indications consistent with multiple dropped control rods following an alarm for the trip of MCC-24/24A. No control rod indication was available due to MCC-24 being faulted and de-energized. The direct cause of the event was the loss of MCC-24 due to an internal fault at the line sides leads at cubicle 2H where they connect to the bucket stab assemblies. The apparent cause was an unanticipated loss of power to the control rod system due to the degradation of the primary control rod PS1 which failed to function when the operating PS2 was lost. The inspectors determined that both the unexpected failure of PS2 and the internal fault in PS1 was not within Entergy's ability to foresee and prevent and was not a performance deficiency. The inspectors reviewed the LER, the associated apparent cause evaluation analysis, and interviewed Entergy staff. This LER is closed.


===.3 (Closed) LER 05000247/2016-003-00:===
===.3 (Closed) LER 05000247/2016-003-00:===
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=====Analysis.=====
=====Analysis.=====
The failure to appropriately scope the safety-related feedwater isolation function within the maintenance rule program was a performance deficiency. This finding is more than minor because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to properly scope the feedwater isolation function prevented Entergy from identifying that equipment reliability was no longer effectively controlled through preventative maintenance. Additionally, this issue is similar to example 7.d described in IMC 0612, Appendix E, "Examples of Minor Issues," dated August 11, 2009. In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. There are redundant methods of feedwater isolation. They include tripping the MBFPs and closing the MBFP discharge valves, closing the main feedwater regulating valves and low flow bypass valves, and closing the main feedwater isolation valves. On both December 5, 2015, and March 7, 2016, the main feedwater regulating valves and isolation valves were functional; so there was no loss of the ability to isolate feedwater to mitigate accident and transient conditions.
The failure to appropriately scope the safety-related feedwater isolation function within the maintenance rule program was a performance deficiency. This finding is more than minor because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by acci dents or events. Specifically, the failure to properly scope the feedwater isolation function prevented Entergy from identifying that equipment reliability was no longer effectively controlled through preventative maintenance. Additionally, this issue is similar to example 7.d described in IMC 0612, Appendix E, "Examples of Minor Issues," dated August 11, 2009. In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. There are redundant methods of feedwater isolation. They include tripping the MBFPs and closing the MBFP discharge valves, closing the main feedwater regulating valves and low flow bypass valves, and closing the main feedwater isolation valves. On both December 5, 2015, and March 7, 2016, the main feedwater regulating valves and isolation valves were functional; so there was no loss of the ability to isolate feedwater to mitigate accident and transient conditions.


This finding does not have a cross-cutting aspect, since the failure to scope this equipment into the maintenance rule program was not recognized when Entergy combined the maintenance rule basis documents for Units 2 and 3 in 2012 and as a result, is not indicative of current licensee performance.
This finding does not have a cross-cutting aspect, since the failure to scope this equipment into the maintenance rule program was not recognized when Entergy combined the maintenance rule basis documents for Units 2 and 3 in 2012 and as a result, is not indicative of current licensee performance.


=====Enforcement.=====
=====Enforcement.=====
10 CFR 50.65(b)(1) requires, in part, that the holders of an operating license shall include within the scope of the monitoring program, specified in 10 CFR 50.65(a)(1), SSCs that are relied upon to remain functional during and following design basis events. Contrary to the above, since the combined maintenance rule scoping for Units 2 and 3 in 2012, Entergy failed to include within the scope of the monitoring program specified in 10 CFR 50.65(a)(1), the safety-related MBFP discharge valves. These SSCs are relied upon during and after design basis events to mitigate the consequences of a feedwater line break accident inside containment. Entergy's corrective action included entering this issue into the corrective action program.
10 CFR 50.65(b)(1) requires, in part, that the holders of an operating license shall include within the scope of the monitoring program, specified in 10 CFR 50.65(a)(1), SSCs that are relied upon to remain functional during and following  
 
design basis events. Contrary to the above, since the combined maintenance rule scoping for Units 2 and 3 in 2012, Entergy failed to include within the scope of the monitoring program specified in 10 CFR 50.65(a)(1), the safety-related MBFP discharge valves. These SSCs are relied upon during and after design basis events to mitigate the consequences of a feedwater line break accident inside containment. Entergy's corrective action included entering this issue into the corrective action program.
 
Because the violation was of very low safety significance (Green) and Entergy entered this issue into their CAP as CR-IP2-2016-03963, this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy.  (NCV 05000247/2016002-04, Failure to Scope Safety-Related Main Boiler Feedwater Pump Discharge Valves into the Maintenance Rule Program)


Because the violation was of very low safety significance (Green) and Entergy entered this issue into their CAP as CR-IP2-2016-03963, this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy.  (NCV 05000247/2016002-04, Failure to Scope Safety-Related Main Boiler Feedwater Pump Discharge Valves into the Maintenance Rule Program)
{{a|4OA5}}
{{a|4OA5}}
==4OA5 Other Activities==
==4OA5 Other Activities==
Line 544: Line 634:


====b. Findings and Observations====
====b. Findings and Observations====
No findings were identified. Update to URI 05000247/2016001-07, January 2016 Groundwater Contamination  Entergy continues to conduct weekly, biweekly, and monthly groundwater sampling of MWs at the initial site of groundwater contamination and at downstream wells towards the Hudson River. For the initial three MWs (MW-30, MW-31, and MW-32), the general trend in tritium activity has been downward, with periodic increases seen in some weekly samples. The downstream MWs located in the Unit 2 switchyard (especially MW-55) showed an initial increase in activity up to 117,000 pCi/l, but the activity at that location has plateaued at the end of the reporting period. Entergy documented its investigation of this event as root cause evaluation for CR-IP2-2016-00564. The inspectors reviewed Entergy's root cause evaluation for this event. Entergy concluded that the source of the groundwater contamination was from the reject water of a temporary reverse osmosis unit used to process water from the refueling water storage tank at Unit 2 in preparation for RFO 2R22. Although this analysis documents a number of issues identified during the operation of the contractor reverse osmosis unit, which is believed to be the source of the groundwater contamination, one of two leakage paths to groundwater have still not been established.
No findings were identified.


The established pathway involves leakage from two cut drain lines located above the floor on the 35-foot elevation of the PAB. Further investigation by Entergy following the conclusion of the Unit 2 RFO 2R22 must be conducted to verify the second pathway to groundwater via the floor of the fuel storage building truck bay. Entergy's long-term corrective action for reducing tritium levels in the groundwater is the same as previously identified for the March 2014 tritium spike (CR-IP2-2015-03806), the start-up and operation of recovery well (RW)-1. Following installation of equipment and system testing, full operation of the RW system is expected later this year. This system will allow for the collection of tritiated groundwater in the vicinity of Unit 2 to be returned inside the Unit 2 PAB for processing. The NRC will be conducting an inspection in August 2016 to review the testing plan and results of the RW-1 tests. This inspection will include a specialist region-based inspector, and a staff hydrogeologist. The NRC's continuing assessment of the safety significance of this event focused on validating the safety impact of dose to the public from the release of tritium to the site groundwater, and ultimately to the Hudson River. The NRC verified that Entergy's bounding public dose calculations on the groundwater contamination leak was sufficiently conservative and a maximum worst case scenario would result in a dose of 0.000112 millirem per year, which represents a very small fraction of the allowable dose (liquid effluent dose objective of 3 millirem per year). This low value is due to groundwater at Indian Point not being a source of any drinking water. There are no drinking water wells on the Indian Point site, groundwater flow from the site is to the Hudson River and not to any near site drinking water wells, and the Hudson River has no downstream drinking water intakes as it is brackish water. Pathways to the public are therefore limited to the consumption of fish and river invertebrates. The inspection determined that there is no safety impact to the public as a result of this groundwater contamination event.  (URI 05000247/2016001-07, January 2016 Groundwater Contamination)
Update to URI 05000247/2016001-07, January 2016 Groundwater Contamination Entergy continues to conduct weekly, biweekly, and monthly groundwater sampling of MWs at the initial site of groundwater contamination and at downstream wells towards the Hudson River. For the initial three MWs (MW-30, MW-31, and MW-32), the general trend in tritium activity has been downward, with periodic increases seen in some weekly samples. The downstream MWs located in the Unit 2 switchyard (especially MW-55) showed an initial increase in activity up to 117,000 pCi/l, but the activity at that location has plateaued at the end of the reporting period.
 
Entergy documented its investigation of this event as root cause evaluation for CR-IP2-2016-00564. The inspectors reviewed Entergy's root cause evaluation for this event. Entergy concluded that the source of the groundwater contamination was from the reject water of a temporary reverse osmosis unit used to process water from the refueling water storage tank at Unit 2 in preparation for RFO 2R22. Although this analysis documents a number of issues identified during the operation of the contractor reverse osmosis unit, which is believed to be the source of the groundwater contamination, one of two leakage paths to groundwater have still not been established.
 
The established pathway involves leakage from two cut drain lines located above the floor on the 35-foot elevation of the PAB. Further investigation by Entergy following the conclusion of the Unit 2 RFO 2R22 must be conducted to verify the second pathway to groundwater via the floor of the fuel storage building truck bay.
 
Entergy's long-term corrective action for reducing tritium levels in the groundwater is the same as previously identified for the March 2014 tritium spike (CR-IP2-2015-03806), the start-up and operation of recovery well (RW)-1. Following installation of equipment and system testing, full operation of the RW system is expected later this year. This system will allow for the collection of tritiated groundwater in the vicinity of Unit 2 to be returned inside the Unit 2 PAB for processing. The NRC will be conducting an inspection in  
 
August 2016 to review the testing plan and results of the RW-1 tests. This inspection will include a specialist region-based inspector, and a staff hydrogeologist.
 
The NRC's continuing assessment of the safety significance of this event focused on validating the safety impact of dose to the public from the release of tritium to the site groundwater, and ultimately to the Hudson River. The NRC verified that Entergy's bounding public dose calculations on the groundwater contamination leak was sufficiently conservative and a maximum worst case scenario would result in a dose of 0.000112 millirem per year, which represents a very small fraction of the allowable dose (liquid effluent dose objective of 3 millirem per year). This low value is due to groundwater at Indian Point not being a source of any drinking water. There are no drinking water wells on the Indian Point site, groundwater flow from the site is to the Hudson River and not to any near site drinking water wells, and the Hudson River has no downstream drinking water intakes as it is brackish water. Pathways to the public are therefore limited to the consumption of fish and river invertebrates. The inspection determined that there is no safety impact to the public as a result of this groundwater  
 
contamination event.  (URI 05000247/2016001-07, January 2016 Groundwater Contamination)


===.2 Institute of Nuclear Power Operations (INPO) Report Review===
===.2 Institute of Nuclear Power Operations (INPO) Report Review===
Line 558: Line 660:
{{a|4OA6}}
{{a|4OA6}}
==4OA6 Meetings, Including Exit==
==4OA6 Meetings, Including Exit==
On August 4, 2016, the inspectors presented the inspection results to Mr. Larry Coyle, Site Vice President, and other members of Entergy. Based on additional information provided, the inspectors conducted an updated exit meeting on August 30, 2016 with John Kirkpatrick, Plant Operations General Manager and other members of Entergy.
On August 4, 2016, the inspectors presented the inspection results to Mr. Larry Coyle, Site Vice President, and other members of Entergy. Based on additional information provided, the inspectors conducted an updated exit meeting on August 30, 2016 with John Kirkpatrick, Plant Operations General Manager and other members of Entergy.


Line 567: Line 670:


==KEY POINTS OF CONTACT==
==KEY POINTS OF CONTACT==
Entergy Personnel  
Entergy Personnel  
: [[contact::A. Vitale]], Site Vice President  
: [[contact::A. Vitale]], Site Vice President  
Line 617: Line 721:
===Opened/Closed===
===Opened/Closed===
: 05000247/2016002-03 NCV  Failure to Maintain Flow Channeling Gates Closed in Accordance with the Containment Procedure  
: 05000247/2016002-03 NCV  Failure to Maintain Flow Channeling Gates Closed in Accordance with the Containment Procedure  
(Section 1R20)
 
: 05000247/2016002-04 NCV  Failure to Scope Safety-Related Main Boiler      Feedwater Pump Discharge Valves into the Maintenance Rule Program (Section 4OA3)
(Section 1R20)  
: 05000247/2016002-04 NCV  Failure to Scope Safety-Related Main Boiler      Feedwater Pump Discharge Valves into the
Maintenance Rule Program (Section 4OA3)  


===Closed===
===Closed===
: [[Closes LER::05000247/LER-2015-003]]-00 LER  Manual Reactor Trip due to Indications of Multiple      Dropped Control Rods Caused by Loss of Control  
: [[Closes LER::05000247/LER-2015-003]]-00 LER  Manual Reactor Trip due to Indications of Multiple      Dropped Control Rods Caused by Loss of Control  
: Rod Power Due to a Power Supply Failure (Section 4OA3)
: Rod Power Due to a Power Supply Failure  
: [[Closes LER::05000247/LER-2016-003]]-00 LER  Technical Specification Prohibited Condition Due to an Inoperable 21 Main Boiler Feedwater Pump Discharge Valve for Greater Than the TS  
(Section 4OA3)
: Allowed Outage Time (Section 4OA3)
: [[Closes LER::05000247/LER-2016-003]]-00 LER  Technical Specification Prohibited Condition Due to an Inoperable 21 Main Boiler Feedwater  
: Pump Discharge Valve for Greater Than the TS  
: Allowed Outage Time (Section 4OA3)  


===Discussed===
===Discussed===
: 05000247/2016001-01 URI Baffle-Former Bolts with Identified Anomalies  
: 05000247/2016001-01 URI Baffle-Former Bolts with Identified Anomalies  
   (Section 1R08)  
   (Section 1R08)  
: 05000247/2016001-06 URI Emergency Diesel Generator Automatic Voltage Regulator Failure (Section 1R15)
: 05000247/2016001-06 URI Emergency Diesel Generator Automatic Voltage
Regulator Failure (Section 1R15)  
: 05000247/2016001-07 URI  January 2016 Groundwater Contamination
: 05000247/2016001-07 URI  January 2016 Groundwater Contamination
Section (Section 4OA5)  
Section (Section 4OA5)  


==LIST OF DOCUMENTS REVIEWED==
==LIST OF DOCUMENTS REVIEWED==
Line 644: Line 753:
: 3-COL-EL-005, Diesel Generators, Revision 37
: 3-COL-EL-005, Diesel Generators, Revision 37
: OAP-019, Component Verification and System Status Control, Revision 7
: OAP-019, Component Verification and System Status Control, Revision 7
: OAP-044, Plant Labeling Program, Revision 3  
: OAP-044, Plant Labeling Program, Revision 3
===Condition Reports===
===Condition Reports===
(CR-IP2) 2016-01311 2016-01505 2016-01761 2016-02330 2016-02428 2016-02470   
(CR-IP2)  
: 2016-01311 2016-01505 2016-01761 2016-02330 2016-02428 2016-02470   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2016-01382 2016-01810   
(CR-IP3)  
: 2016-01382 2016-01810   
===Drawings===
===Drawings===
: 209762, Flow Diagram Service Water System Nuclear Steam Supply Plant, Revision 75  
: 209762, Flow Diagram Service Water System Nuclear Steam Supply Plant, Revision 75  
: 27781, Flow Diagram Auxiliary Coolant System, Revision 22 9321-2720, Auxiliary Coolant System, Sheet 2, Revision 22  
: 27781, Flow Diagram Auxiliary Coolant System, Revision 22 9321-2720, Auxiliary Coolant System, Sheet 2, Revision 22
===Miscellaneous===
===Miscellaneous===
: IP3-DBD-308, CCW System, Revision 3
: IP3-DBD-308, CCW System, Revision 3
Line 660: Line 771:
: EN-MA-133, Control of Scaffolding, Revision 12   
: EN-MA-133, Control of Scaffolding, Revision 12   
===Condition Reports===
===Condition Reports===
(CR-IP2) 2016-04148   
(CR-IP2)  
: 2016-04148   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2016-01272   
(CR-IP3)  
: 2016-01272   
===Miscellaneous===
===Miscellaneous===
: PFP-203, Containment Building, 95-Foot Elevation (Fire Zone 86A), Revision 15
: PFP-203, Containment Building, 95-Foot Elevation (Fire Zone 86A), Revision 15
Line 675: Line 788:
: 0-HTX-405-EDG, EDG Lube Oil and Jacket Water Heat Exchanger Maintenance, Revision 4   
: 0-HTX-405-EDG, EDG Lube Oil and Jacket Water Heat Exchanger Maintenance, Revision 4   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2010-02900 2011-03594 2011-03596 2011-03961 2012-02071 2012-03912  
(CR-IP3)  
: 2010-02900 2011-03594 2011-03596 2011-03961 2012-02071 2012-03912  
: 2013-02338 2013-02695 2013-03009 2014-00957 2014-01239 2014-03158  
: 2013-02338 2013-02695 2013-03009 2014-00957 2014-01239 2014-03158  
: 2014-03175 2015-00031 2015-00599 2015-02848 2015-05209 2015-05526  
: 2014-03175 2015-00031 2015-00599 2015-02848 2015-05209 2015-05526  
Line 681: Line 795:
: Maintenance Orders/Work Orders
: Maintenance Orders/Work Orders
: WO 52489888
: WO 52489888
: WO 52626563  
: WO 52626563
===Miscellaneous===
===Miscellaneous===
: SEP-SW-IPC-001, Indian Point Energy Center NRC Generic Letter 89-13 Service Water Program, Revision 0
: SEP-SW-IPC-001, Indian Point Energy Center NRC Generic Letter 89-13 Service Water Program, Revision 0
Line 691: Line 805:
: GBRA-175-115, Field Service Procedure for Baffle Bolt Replacement, Revision 3
: GBRA-175-115, Field Service Procedure for Baffle Bolt Replacement, Revision 3
: WDI-STD-088, Underwater Remote Visual Examination of Reactor Vessel Internals, Revision 13
: WDI-STD-088, Underwater Remote Visual Examination of Reactor Vessel Internals, Revision 13
: WDI-STD-1073, Ultrasonic Test Procedure for the Inspection of Internal Hex Head Baffle-Former Bolts with Welded Lock Bars, Revision 4  
: WDI-STD-1073, Ultrasonic Test Procedure for the Inspection of Internal Hex Head Baffle-Former Bolts with Welded Lock Bars, Revision 4
===Condition Reports===
===Condition Reports===
(CR-IP2) 2016-02081  
(CR-IP2)  
: Maintenance Orders/Work Orders
: 2016-02081  
: 442412-13  
: Maintenance Orders/Work Orders  
: 2412-13
===Miscellaneous===
===Miscellaneous===
: Indian Point Unit 2 Baffle Bolt Ultrasonic Examination Expanded Analysis Report, dated April 28, 2016 IP2 Reactor Vessel Visual Examination Report, dated May 2006 Loose Parts Inventory Log for Baffle Bolt Replacements, dated May 24, 2016
: Indian Point Unit 2 Baffle Bolt Ultrasonic Examination Expanded Analysis Report, dated April 28, 2016 IP2 Reactor Vessel Visual Examination Report, dated May 2006 Loose Parts Inventory Log for Baffle Bolt Replacements, dated May 24, 2016
Line 712: Line 827:


===Procedures===
===Procedures===
: 2-AOP-480V-1, Loss of Normal Power to Any 480V Vital Bus, Revision 8 2-AOP-RCP-1, Reactor Coolant Pump Malfunction, Revision 14
: 2-AOP-480V-1, Loss of Normal Power to Any 480V Vital Bus, Revision 8  
: 2-AOP-RCP-1, Reactor Coolant Pump Malfunction, Revision 14
: 2-AOP-TURB-1, Main Turbine Trip without a Reactor Trip, Revision 5  
: 2-AOP-TURB-1, Main Turbine Trip without a Reactor Trip, Revision 5  
: 2-E-0, Reactor Trip or Safety Injection, Revision 7  
: 2-E-0, Reactor Trip or Safety Injection, Revision 7  
Line 718: Line 834:
: 2-POP-1.2, Reactor Startup, Revision 59 2-SOP-26.4, Turbine Generator Startup, Synchronizing, Voltage Control and Shutdown, Revision 62 3-AOP-480V-1, Loss of Normal Power to Any Safeguards Bus, Revision 7  
: 2-POP-1.2, Reactor Startup, Revision 59 2-SOP-26.4, Turbine Generator Startup, Synchronizing, Voltage Control and Shutdown, Revision 62 3-AOP-480V-1, Loss of Normal Power to Any Safeguards Bus, Revision 7  
: 3-AOP-CVCS-1, Chemical and Volume Control System Malfunction, Revision 8  
: 3-AOP-CVCS-1, Chemical and Volume Control System Malfunction, Revision 8  
: 3-AOP-FW-1, Loss of Feedwater, Revision 8 3-AOP-INST-1, Instrument/Controller Failures, Revision 11 3-E-0, Reactor Trip or Safety Injection, Revision 6  
: 3-AOP-FW-1, Loss of Feedwater, Revision 8 3-AOP-INST-1, Instrument/Controller Failures, Revision 11  
: 3-E-0, Reactor Trip or Safety Injection, Revision 6  
: 3-E-1, Loss of Reactor or Secondary Coolant, Revision 4  
: 3-E-1, Loss of Reactor or Secondary Coolant, Revision 4  
: 3-FR-C.2, Response to Degraded Core Cooling, Revision 3   
: 3-FR-C.2, Response to Degraded Core Cooling, Revision 3   
===Condition Reports===
===Condition Reports===
(CR-IP2) 2016-03946 2016-04162 2016-04164 2016-04165 2016-04169 2016-04178   
(CR-IP2)  
: 2016-03946 2016-04162 2016-04164 2016-04165 2016-04169 2016-04178   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2016-01087 2016-01092 2016-01098 2016-01336   
(CR-IP3)  
: 2016-01087 2016-01092 2016-01098 2016-01336   
===Miscellaneous===
===Miscellaneous===
: 13SX-LOR-SES026, Licensed Operator Requalification Program Scenario Emergency Action Level Table, Revision 15.2
: 13SX-LOR-SES026, Licensed Operator Requalification Program Scenario Emergency Action Level Table, Revision 15.2
Line 734: Line 853:
: CEP-NDE-0640, Non-Section XI Liquid Penetrant Examination, Revision 9
: CEP-NDE-0640, Non-Section XI Liquid Penetrant Examination, Revision 9
: CEP-WP-WIIR-1, Attachment 5.1, Inprocess Inspections for Installation and Replacement Welds Located Inside the ASME Section XI Boundary, Revision 3
: CEP-WP-WIIR-1, Attachment 5.1, Inprocess Inspections for Installation and Replacement Welds Located Inside the ASME Section XI Boundary, Revision 3
: EN-DC-206, Maintenance Rule (a)(1) Process, Revision 3  
: EN-DC-206, Maintenance Rule (a)(1) Process, Revision 3
===Condition Reports===
===Condition Reports===
(CR-IP2) 2010-00864 2013-03130 2014-00162 2014-00185 2014-01144 2014-02184   
(CR-IP2) 2010-00864 2013-03130 2014-00162 2014-00185 2014-01144 2014-02184   
: 2015-00278 2016-01260 2016-01430 2016-01500   
: 2015-00278 2016-01260 2016-01430 2016-01500   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2012-03836 2013-04758 2015-01396 2015-03404 2015-03653 2015-04053 2015-04162 2015-04184 2015-04539 2015-05316 2015-05384 2015-05729  
(CR-IP3)  
: 2016-00098 2016-00653 2016-00723 2016-01189 2016-01227 2016-01274 2016-01313 2016-01531 2016-01536 2016-01543 2016-02432   
: 2012-03836 2013-04758 2015-01396 2015-03404 2015-03653 2015-04053  
: 2015-04162 2015-04184 2015-04539 2015-05316 2015-05384 2015-05729  
: 2016-00098 2016-00653 2016-00723 2016-01189 2016-01227 2016-01274  
: 2016-01313 2016-01531 2016-01536 2016-01543 2016-02432   
: Maintenance Orders/Work Orders
: Maintenance Orders/Work Orders
: WO 00397793
: WO 00397793
Line 758: Line 880:
: EC 65389, 21 CCW HX Leak Repair: Service Water Inlet Nozzle - Elbow Weld Configuration Change IPEC Maintenance Rule Basis Document - Chemical and Volume Control System, Revision 0
: EC 65389, 21 CCW HX Leak Repair: Service Water Inlet Nozzle - Elbow Weld Configuration Change IPEC Maintenance Rule Basis Document - Chemical and Volume Control System, Revision 0
: PQR 913, 134 F42
: PQR 913, 134 F42
: MN-GTAW ASME IX Welding Procedure Qualification Record, Revision 0 System Health Report, Unit 3, EDG, Q1-2016 Weld Map Number 447966-20-01, Revision 0
: MN-GTAW ASME IX Welding Procedure Qualification Record, Revision 0 System Health Report, Unit 3, EDG, Q1-2016  
: Weld Map Number 447966-20-01, Revision 0
: WPS 134 F42,
: WPS 134 F42,
: MN-GTAW, ASME Section IX Welding Procedure Specification, Revision 0
: MN-GTAW, ASME Section IX Welding Procedure Specification, Revision 0
Line 766: Line 889:
===Procedures===
===Procedures===
: EN-OP-119, Protected Equipment, Revision 8
: EN-OP-119, Protected Equipment, Revision 8
: IP-SMM-OU-104, Attachment 9.1, Shiftly Outage Shutdown Safety Assessments, Revision 15
: IP-SMM-OU-104, Attachment 9.1, Shiftly Ou tage Shutdown Safety Assessments, Revision 15
: IP-SMM-OU-104, Attachment 9.2, Shiftly Outage Shutdown Safety Assessment Guidelines, Revision 15  
: IP-SMM-OU-104, Attachment 9.2, Shiftly Ou tage Shutdown Safety Assessment Guidelines, Revision 15
===Condition Reports===
===Condition Reports===
(CR-IP2) 2016-04141   
(CR-IP2)  
: 2016-04141   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2016-01545   
(CR-IP3)  
: 2016-01545   
===Miscellaneous===
===Miscellaneous===
: EOOS Risk Assessment Software Tool
: EOOS Risk Assessment Software Tool
Line 786: Line 911:
: 2016-3636 2016-3784 2016-3806 2016-3818 2016-4085   
: 2016-3636 2016-3784 2016-3806 2016-3818 2016-4085   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2014-01670 2015-03468  
(CR-IP3)  
: 2014-01670 2015-03468  
: Maintenance Orders/Work Orders
: Maintenance Orders/Work Orders
: WO 00327574
: WO 00327574
: WO 00425980
: WO 00425980
: WO 52571030  
: WO 52571030
===Miscellaneous===
===Miscellaneous===
: EN-LI-100, Attachment 9.1, Change Channel Check Comparison Criteria/2-PT-M100, 2-PT-D001, Revision 0
: EN-LI-100, Attachment 9.1, Change Channel Check Comparison Criteria/2-PT-M100,  
: 2-PT-D001, Revision 0


==Section 1R18: Plant Modifications==
==Section 1R18: Plant Modifications==


===Drawings===
===Drawings===
: 10073E87-001, Indian Point Unit 2 Baffle Bolt Replacement - Core Barrel and Baffle Assembly Elevation, Revision 0 10111D06, Indian Point Unit 2 Baffle Bolt Replacement - Replacement Baffle Bolt .625 and .750, Revision 0  
: 10073E87-001, Indian Point Unit 2 Baffle Bolt Replacement - Core Barrel and Baffle Assembly Elevation, Revision 0 10111D06, Indian Point Unit 2 Baffle Bolt Replacement - Replacement Baffle Bolt .625
and .750, Revision 0
===Miscellaneous===
===Miscellaneous===
: EC 64308, IP2 Reactor Vessel Equivalent Replacement Baffle-to-Former Bolt, Revision 0  
: EC 64308, IP2 Reactor Vessel Equivalent Replacement Baffle-to-Former Bolt, Revision 0  
Line 807: Line 935:


===Procedures===
===Procedures===
: 3-PT-M079B, 32 EDG Functional Test, Revision 52 2-PC-Q109-4, Recalibration of NIS and OT/OP Delta-T Parameters - Channel IV, Revision 44  
: 3-PT-M079B, 32 EDG Functional Test, Revision 52 2-PC-Q109-4, Recalibration of NIS and OT/OP Delta-T Parameters - Channel IV, Revision 44
===Condition Reports===
===Condition Reports===
(CR-IP2) 2016-03961 2016-04266   
(CR-IP2)  
: 2016-03961 2016-04266   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2016-01189 2016-01199 2016-01218
(CR-IP3)  
: 2016-01189 2016-01199 2016-01218  
: Maintenance Orders/Work Orders
: Maintenance Orders/Work Orders
: WO 00414886
: WO 00414886
Line 827: Line 957:
: WO 52571030   
: WO 52571030   
===Drawings===
===Drawings===
: 5651D72, Logic Diagrams Steam Generator Trip Signals, Revision 7  
: 5651D72, Logic Diagrams Steam Generator Trip Signals, Revision 7
===Miscellaneous===
===Miscellaneous===
: EC 64545, Emergency Temporary Modification to Disconnect 32 EDG Generator Space Heater Adjacent to End Plate on Outboard End of Generator FIX00091, Pressurizer Level Uncertainty - Indication, Trip Setpoints, and Annunciation Setpoints, Revision 1 E-mail from J. Michetti to G. Newman, dated July 19, 2016, Subject:
: EC 64545, Emergency Temporary Modification to Disconnect 32 EDG Generator Space Heater Adjacent to End Plate on Outboard End of Generator FIX00091, Pressurizer Level Uncertainty - Indication, Trip Setpoints, and Annunciation Setpoints, Revision 1 E-mail from J. Michetti to G. Newman, dated July 19, 2016, Subject:
Line 843: Line 973:
: 2-POP-4.3, Operation without Fuel in the Reactor, Revision 1   
: 2-POP-4.3, Operation without Fuel in the Reactor, Revision 1   
===Condition Reports===
===Condition Reports===
(CR-IP2-) 2016-04118 2016-04119 2016-04123 2016-03124 2016-04126 2016-04129  
(CR-IP2-)  
: 2016-04118 2016-04119 2016-04123 2016-03124 2016-04126 2016-04129  
: 2016-04130 2016-04131 2016-04132 2016-04139 2016-04141* 2016-04142*  
: 2016-04130 2016-04131 2016-04132 2016-04139 2016-04141* 2016-04142*  
: 2016-04144 2016-04145 2016-04146 2016-04148* 2016-04151 2016-04152  
: 2016-04144 2016-04145 2016-04146 2016-04148* 2016-04151 2016-04152  
: 2016-04155 2016-04161 2016-04162 2016-04165 2016-04169  
: 2016-04155 2016-04161 2016-04162 2016-04165 2016-04169  
  *NRC identified Maintenance Orders/Work Orders
 
: 52681465  
  *NRC identified  
: Maintenance Orders/Work Orders
: 2681465  
===Miscellaneous===
===Miscellaneous===
: 2R22 Performance Indicator Curves Daily from March 7 to June 14, 2016  
: 2R22 Performance Indicator Curves Daily from March 7 to June 14, 2016  
Line 862: Line 995:
: 3-PT-M079B, 32 EDG Functional Test, Revision 52   
: 3-PT-M079B, 32 EDG Functional Test, Revision 52   
===Condition Reports===
===Condition Reports===
(CR-IP2) 2016-03360 2016-03363  
(CR-IP2)  
: 2016-03360 2016-03363
===Condition Reports===
===Condition Reports===
(CR-IP3) 2016-01716 2016-01752  
(CR-IP3)  
: 2016-01716 2016-01752  
: Maintenance Orders/Work Orders
: Maintenance Orders/Work Orders
: WO 00443040
: WO 00443040
Line 887: Line 1,022:
===Procedures===
===Procedures===
: 0-CY-1920, REMP Land Use Census, Revision 1  
: 0-CY-1920, REMP Land Use Census, Revision 1  
: 0-CY-1980, Preparation, Placement and Collection of Site Environmental Thermoluminescent Dosimeters, Revision 2  
: 0-CY-1980, Preparation, Placement and Collection of Site Environmental Thermoluminescent Dosimeters, Revision 2
===Condition Reports===
===Condition Reports===
(CR-IP2) 2014-05319 2015-00948 2015-01300 2015-02687 2015-02800 2015-02987  
(CR-IP2)  
: 2014-05319 2015-00948 2015-01300 2015-02687 2015-02800 2015-02987  
: 2015-03271 2015-03396 2016-02313   
: 2015-03271 2015-03396 2016-02313   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2016-00514   
(CR-IP3)  
: 2016-00514   
===Miscellaneous===
===Miscellaneous===
: 2014 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3  
: 2014 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3  
: 2015 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3 Environmental Dosimetry Company, Annual Quality Assurance Status Report, January to December 2015 Indian Point Energy Center ODCM, Revision 4  
: 2015 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3 Environmental Dosimetry Company, Annual Quality Assurance Status Report, January to December 2015 Indian Point Energy Center ODCM, Revision 4  
: June 2015 to May 2016 Meteorological Data Recovery Met One Instruments, Inc. Certificates of Calibration for Temperature, Wind Direction, and Wind Speed Teledyne Brown Engineering Environmental Services Annual 2015 Quality Assurance Report Exelon PowerLabs Certificates of Calibration for Gas Meters
: June 2015 to May 2016 Meteorological Data Recovery Met One Instruments, Inc. Certificates of Calibration for Temperature, Wind Direction, and Wind Speed Teledyne Brown Engineering Environmental Serv ices Annual 2015 Quality Assurance Report  
: Exelon PowerLabs Certificates of Calibration for Gas Meters
: 3471875
: 3471875
: 3482909
: 3482909
Line 914: Line 1,052:
: 3471869
: 3471869
: 3471880
: 3471880
: 3482908
: 3482908  
: Quality Assurance Quality Assurance Audit Report
: Quality Assurance Quality Assurance Audit Report
: QA-2-6-2015-IP-1, Chemistry, Effluents, and Environmental Monitoring Snapshot Self-Assessment,
: QA-2-6-2015-IP-1, Chemistry, Effluents, and Environmental Monitoring Snapshot Self-Assessment,
Line 931: Line 1,069:
: EN-OP-104, Operability Determination Process, Revision 10   
: EN-OP-104, Operability Determination Process, Revision 10   
===Condition Reports===
===Condition Reports===
(CR-IP2) 2010-07013 2015-04574 2015-05458 2015-05460 2015-05461 2015-05464  
(CR-IP2)  
: 2010-07013 2015-04574 2015-05458 2015-05460 2015-05461 2015-05464  
: 2015-05466 2015-05467 2016-01374 2016-02348   
: 2015-05466 2015-05467 2016-01374 2016-02348   
===Condition Reports===
===Condition Reports===
(CR-IP3) 2015-3628 2016-01035 2016-01961
(CR-IP3) 2015-3628 2016-01035 2016-01961  
: Maintenance Orders/Work Orders
: Maintenance Orders/Work Orders
: WO 00442412  
: WO 00442412  
Line 947: Line 1,086:
: CN-RIDA-15-43, Indian Point Units 2 and 3 Acceptable Baffle-Former Bolting LOCA and Seismic Analysis, Revision 2 Engineering Change 63938,
: CN-RIDA-15-43, Indian Point Units 2 and 3 Acceptable Baffle-Former Bolting LOCA and Seismic Analysis, Revision 2 Engineering Change 63938,
: As-left condition of the baffle-former plate assembly following the replacement of degraded bolts, Revision 0 EPRI
: As-left condition of the baffle-former plate assembly following the replacement of degraded bolts, Revision 0 EPRI
: TR-112209, Analysis of Baffle Former Bolt Cracking in EDF CPO Plants (PWRMRP-03), dated June 1999 Indian Point Entergy Center (IPEC) Unit 3, Updated Final Safety Analysis Report, dated May 2013
: TR-112209, Analysis of Baffle Former Bolt Cracking in EDF CPO Plants (PWRMRP-03), dated June 1999 Indian Point Entergy Center (IPEC) Unit 3, Updated Final Safety Analysis Report, dated May  
: 2013
: IP-RPT-16-00025, Evaluation of Indian Point Unit 3 Reactor Core Baffle Bolting Following
: IP-RPT-16-00025, Evaluation of Indian Point Unit 3 Reactor Core Baffle Bolting Following
: MRP-227-A Inspection Findings at Indian Point Unit 2 during 2R22, Revision 0
: MRP-227-A Inspection Findings at Indian Point Unit 2 during 2R22, Revision 0
Line 957: Line 1,097:
: MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (ML120170453)   
: MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (ML120170453)   
: Operation Decision Making Issue Action Plan for IP3 Baffle Bolt Monitoring, dated May 19, 2016
: Operation Decision Making Issue Action Plan for IP3 Baffle Bolt Monitoring, dated May 19, 2016
: WCAP-15029-NP-A, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions, Revision 0 (ML15222A882)
: WCAP-15029-NP-A, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions, Revision 0  
(ML15222A882)
: WCAP-17949-P, Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for Indian Point Unit 2 Reactor Vessel Internals
: WCAP-17949-P, Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for Indian Point Unit 2 Reactor Vessel Internals
: MRP-227-A Primary and Expansion Components, Revision 1
: MRP-227-A Primary and Expansion Components, Revision 1
: WCAP-18048-P, Determination of Acceptable Baffle-Former Bolting for Indian Point Units 2 and 3, Revision 0
: WCAP-18048-P, Determination of Acceptable Baffle-Former Bolting for Indian Point Units 2 and  
: 3, Revision 0


==Section 4OA5: Other Activities==
==Section 4OA5: Other Activities==


===Miscellaneous===
===Miscellaneous===
: INPO Letter, INPO Equipment Reliability Scram Review Visit, May 31, 2016
: INPO Letter, INPO Equipment Reliabilit
y Scram Review Visit, May 31, 2016
: Root Cause Evaluation for CR-IP2-2016-00564
: Root Cause Evaluation for CR-IP2-2016-00564
==LIST OF ACRONYMS==
==LIST OF ACRONYMS==
: [[10 CFR]] [[Title 10 of the Code of Federal Regulations]]
 
CFR Title 10 of the
Code of Federal Regulations
: [[ALARA]] [[as low as is reasonably achievable]]
: [[ALARA]] [[as low as is reasonably achievable]]
: [[AVR]] [[automatic voltage regulator]]
: [[AVR]] [[automatic voltage regulator]]
Line 990: Line 1,135:
: [[MW]] [[monitoring well]]
: [[MW]] [[monitoring well]]
: [[NCV]] [[non-cited violation]]
: [[NCV]] [[non-cited violation]]
: [[NRC]] [[Nuclear Regulatory Commission, U.S.]]
: [[NRC]] [[Nuclear Regulatory Commission,]]
: [[U.S.]] [[]]
: [[ODCM]] [[offsite dose calculation manual]]
: [[ODCM]] [[offsite dose calculation manual]]
: [[OOS]] [[out of service]]
: [[OOS]] [[out of service]]

Revision as of 20:53, 29 June 2018

Indian Point Nuclear Generating - Revised Integrated Inspection Report 05000247/2016002 and 05000286/2016002
ML17089A245
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 03/30/2017
From: Setzer T C
Reactor Projects Branch 2
To: Vitale A J
Entergy Nuclear Operations
Setzer T C
References
EA-16-223 IR 2016002
Download: ML17089A245 (51)
v • d • e


Text

March 30, 2017

EA-16-223

Mr. Anthony Site Vice President Entergy Nuclear Operations, Inc.

Indian Point Energy Center

450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT: INDIAN POINT NUCLEAR GENERATING - REVISED INTEGRATED INSPECTION REPORT 05000247/2016002 AND 05000286/2016002

Dear Mr. Vitale:

In a letter dated September 29, 2016 (ML16281A486 1), Entergy Nuclear Operations, Inc.

(Entergy), provided a response to U.S. Nuclear Regulatory Commission (NRC) inspection report 05000247/2016002 and 05000286/2016002 (ML16243A245) issued on August 30, 2016. In this letter, Entergy disagreed with non-cited violation (NCV)05000286/2016002-02, "Failure to Follow Operability Determination Procedure for Unit 3 Baffle-Former Bolts." This was a Green

NCV of Title 10 of the Code of Federal Regulations (10 CFR), Appendix B, Criterion V, "Instructions, Procedures, and Drawings," and described a failure by Entergy to follow the operability process as prescribed in Entergy procedures at Indian Point concerning degraded baffle-former bolts in the Unit 2 reactor vessel. Specifically, Entergy determined that a more detailed operability evaluation for the Unit 3 reactor vessel was not required in accordance with procedure EN-OP-104, "Operability Determination Process." In the aforementioned letter, Entergy provided the NRC with an interpretation of procedure EN-OP-104 related to the baffle-former bolt condition that was different from the NRC's initial position as described in the inspection report.

In response to the contested NCV, the NRC formed a panel to review the information provided by Entergy. The staff members of the panel were independent of the initial inspection effort.

Based on a thorough review of the issue, the NRC concluded that the violation as written could not be supported. Specifically, the independent review noted different steps within the procedure that could be interpreted to either support or refute Entergy's claim related to the need for direct evidence of a degraded or non-conforming condition. As a result of this potential for differing interpretations of the procedural requirements, the NRC determined it was not appropriate to issue a violation for failure to comply with the procedure. The NRC issued the withdrawal of the NCV in a March 2, 2017 letter (ML17061A112). Therefore, the NRC is

1 The designation in parentheses refers to an Agencywide Documents Access and Management System (ADAMS) Accession Number. Documents referenced in this letter are publicly available using the Accession Number in ADAMS. removing the NCV documented in Section 1R15 of the enclosed inspection report 05000247/2016002 and 05000286/2016002.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding."

Sincerely,/RA/ Thomas C. Setzer, Acting Chief Reactor Projects Branch 2 Division of Reactor Projects

Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

Enclosure:

Inspection Report 05000247/2016002 and 05000286/2016002

w/Attachment:

Supplementary Information cc w/encl: Distribution via ListServ

SUMMARY

Inspection Report 05000247/2016002 and 05000286/2016002; 04/01/2016 - 06/30/2016; Indian Point Nuclear Generating (Indian Point), Units 2 and 3; Refueling and Other Outage Activities, and Follow Up of Events and Notices of Enforcement Discretion.

This report covered a three-month period of inspection by resident inspectors and announced inspections performed by regional inspectors. The inspectors identified two findings of very low safety significance (Green), which were non-cited violations (NCVs). The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310, "Aspects within the Cross-Cutting Areas," dated December 4, 2014. All violations of U.S. Nuclear Regulatory Commission (NRC) requirements are dispositioned in accordance with the NRC's Enforcement Policy, dated February 4, 2015. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 6.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1, "Procedures," for Entergy's failure to implement procedure OAP-007, "Containment Entry and Egress." Specifically, workers transiting the inner and outer crane wall sections of containment failed to maintain at least one (of two) flow channeling gate closed to ensure availability of the containment sumps to provide suction for the emergency core cooling system (ECCS). Entergy immediately coached the gate monitor and restored the gates to an acceptable position. Entergy generated condition report (CR)-IP2-2016-04036 to address this issue.

This performance deficiency is more than minor because it was associated with the configuration control (shutdown equipment lineup) attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A detailed risk assessment was conducted and determined that the change in core damage frequency was determined to be 7E-9, therefore, this issue represents a Green finding. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Entergy did not consider potential undesired consequences of actions before performing work and implement appropriate error-reduction tools. Specifically, the work crew did not understand the requirements and potential consequences prior to commencing work and the gate monitor did not enforce these requirements to maintain at least one gate locked or pinned closed as required by OAP-007.

[H.12 - Avoid Complacency] (Section 1R20)

Cornerstone: Barrier Integrity

Green.

The inspectors identified a Green NCV of 10 CFR 50.65(b)(1) for Entergy's failure to include a function of a safety-related system within the scope of the maintenance rule program. Specifically, Entergy failed to include the feedwater isolation function performed by the main boiler feedwater pumps (MBF Ps) discharge valves, MBFPs, and feedwater regulating valves, which are required to remain functional during and following a design basis event to mitigate the consequence of the accident within the scope of the maintenance rule monitoring program. Entergy initiated corrective actions to include the feedwater isolation function performed by the MBFP discharge valves, MBFPs, and feedwater regulating valves within the maintenance rule monitoring program. Entergy entered this issue into the corrective action program (CAP) as CR-IP2-2016-03963.

This performance deficiency is more than minor because it was associated with barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to properly scope the feedwater isolation function prevented Entergy from identifying that equipment reliability was no longer effectively controlled through preventive maintenance. In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. This finding does not have a cross-cutting aspect since the failure to scope this equipment into the maintenance rule program was not recognized when Entergy combined the maintenance rule basis documents for Units 2 and 3 in 2012 and, as a result, is not indicative of current licensee performance. (Section 4OA3)

5

REPORT DETAILS

Summary of Plant Status

Unit 2 began the inspection period during refueling outage (RFO) 2R22 which lasted 102 days.

Upon completion of the outage, the operators restarted Unit 2 on June 14, 2016, and increased power slowly to 93 percent for fuel preconditioning. On June 23, 2016, the operators shutdown the reactor to repair a service water leak on the 21 component cooling water (CCW) heat exchanger (Hx) inlet line and replace switchyard breaker 9. Unit 2 returned to 100 percent power on June 29, 2016. Unit 2 remained at or near 100 percent power for the remainder of the inspection period.

Unit 3 began the inspection period at 100 percent power. On April 26, 2016, a failed controller caused both heater drain pumps to trip; and the operators reduced power rapidly, stabilizing the unit at 48 percent power. Operators returned Unit 3 to 100 percent power on April 27, 2016, and remained at or near 100 percent power for the remainder of the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R04 Equipment Alignment

Partial System Walkdowns (71111.04Q - 5 samples)

a. Inspection Scope

The inspectors performed partial walkdowns of the following systems:

Unit 2 Spent fuel pool cooling system following core offload on May 19, 2016 Shutdown cooling system following core reload on June 6, 2016 CCW system following maintenance on June 28, 2016 Unit 3 32 emergency diesel generator (EDG) following maintenance on May 9, 2016 (this sample was part of an in-depth review of the EDG system) Residual heat removal pumps following CCW system testing on May 20, 2016 The inspectors selected these systems based on their risk-significance relative to the reactor safety cornerstones at the time they were inspected. The inspectors reviewed applicable operating procedures, system diagrams, the updated final safety analysis report (UFSAR), TSs, work orders (WOs), CRs, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have impacted system performance of their intended safety functions. The inspectors also performed field walkdowns of accessible portions of the systems to verify system components and support equipment were aligned correctly and were operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no deficiencies. The inspectors also reviewed whether Entergy had properly identified equipment issues and entered them into the CAP for resolution with the appropriate significance characterization. Documents reviewed for each section of this inspection report are listed in the

.

b. Findings

No findings were identified.

1R05 Fire Protection

Resident Inspector Quarterly Walkdowns (71111.05Q - 6 samples)

a. Inspection Scope

The inspectors conducted tours of the areas listed below to assess the material condition and operational status of fire protection features. The inspectors verified that Entergy controlled combustible materials and ignition sources in accordance with administrative procedures. The inspectors verified that fire protection and suppression equipment were available for use as specified in the area pre-fire plan (PFP) and passive fire barriers were maintained in good material condition. The inspectors also verified that station personnel implemented compensatory measures for out-of-service (OOS), degraded, or inoperable fire protection equipment, as applicable, in accordance with procedures.

Unit 2 Containment, 95-foot elevation, during baffle bolt repair activities with hot work in progress (PFP-203 was reviewed) on June 2, 2016 Residual heat removal pump rooms in primary auxiliary building (PAB), 15-foot elevation (PFP-204 was reviewed), on June 6, 2016 CCW pump room, 68-foot elevation (PFP-209 was reviewed), on June 25, 2016 PAB, 80-foot elevation, CCW heat exchanger area with hot work in progress (PFP-211 was reviewed) on June 25, 2016

Unit 3 32 EDG room, 10-foot elevation (PFP-354 was reviewed), on May 9, 2016 480V switchgear room, 15-foot elevation (PFP-351 was reviewed), on June 30, 2016

b. Findings

No findings were identified.

1R07 Heat Sink Performance

Main article: IP 71111.07A

a. Inspection Scope

The inspectors reviewed the 32 EDG jacket water and lube oil heat exchanger to determine its readiness and availability to perform its safety functions. The inspectors reviewed the design basis for the component and verified Entergy's commitments to NRC Generic Letter 89-13, "Service Water System Requirements Affecting Safety-Related Equipment." The inspectors observed the annual cleaning and inspection of the heat exchangers and reviewed the results of previous inspections of the Unit 3 EDG heat exchangers. The inspectors discussed the results of the most recent inspection with engineering staff. The inspectors verified that Entergy initiated appropriate corrective actions for identified deficiencies. The inspectors also verified that the number of tubes plugged within the heat exchanger did not exceed the maximum amount allowed.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

Main article: IP 71111.08P

a. Inspection Scope

Inspectors from the NRC Region I Office, specializing in materials and inservice examination activities, observed portions of Entergy's activities involving baffle-former bolt examinations and replacements during Unit 2 RFO 2R22. The inspectors reviewed work documentation and examination procedures and results, and discussed these activities with Entergy. The inspectors were on-site from April 27 to April 28, 2016, and on May 23, 2016. The inspectors verified that Entergy completed baffle-former bolt examinations in accordance with their approved procedures which implemented activities described in the Materials Reliability Program (MRP)-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," as they relate to this component. Specifically, the inspectors reviewed the results of the visual and volumetric examinations of the baffle-former bolts, including capabilities, limitations, and acceptance criteria that were performed during the current RFO.

Non-Destructive Examination Activities The inspectors reviewed the ultrasonic testing (UT) procedure used for the examination of the Unit 2 baffle-former bolts to verify the procedure was in accordance with the applicable guidance in MRP-227-A and MRP-228. The inspectors reviewed the UT data records and the detailed UT channel analysis for a sample of baffle-former bolts to verify the examinations and evaluations were performed in accordance with approved procedures and applicable guidance. The inspectors reviewed video recordings of the visual examinations of the baffle-former bolts during the current RFO. The inspectors also reviewed recorded video of visual examinations performed in 2006 at Unit 2, completed as part of the existing inservice inspection program for the 10-year reactor vessel examinations, to independently assess the past conditions of the baffle-former bolts and assembly.

The inspectors reviewed certifications of the UT technicians performing the ultrasonic examinations to verify the examinations were performed by qualified individuals and to verify the results were reviewed and evaluated by certified level III non-destructive examination personnel.

Baffle-Former Bolt Replacement Activities

The inspectors reviewed the baffle-former bolt replacement activities performed as part of a corrective action to resolve the degraded condition identified at Unit 2. The inspectors observed a sample of in-process bolt removal activities, which included lock bar milling and bolt hole machining. The inspectors reviewed the documentation for in-process and completed bolt installation activities and verified that loose parts generated as part of the bolt replacements were properly tracked. The inspectors verified that bolt replacement activities were performed in accordance with approved procedures. The inspectors also reviewed the Engineering Change (EC) package associated with the new baffle-former bolt design. This review is documented in Section 1R18 of this report. After completion of the bolt replacement activities, the inspectors reviewed the video of the final visual examination of the baffle assembly to verify that the baffle-former bolt work was accomplished as planned and that there were no visual indications of deficiencies.

b. Findings

No findings were identified.

Update to URI 05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies This inspection was conducted to follow-up on NRC Unresolved Item (URI)05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies, to determine whether there was a performance deficiency associated with the degraded baffle-former bolt condition discovered at Unit 2. The inspectors plan to review additional technical information from Entergy as it becomes available, including any revisions to the root cause evaluation. The URI remains open until review of this additional information is

completed. (URI 05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies)

1R11 Licensed Operator Requalification Program

Main article: IP 71111.11Q

Unit 2

.1 Quarterly Review of Unit 2 Licensed Operator Requalification Testing and Training

(71111.11Q - 1 sample)

a. Inspection Scope

The inspectors observed Unit 2 licensed operator simulator training on May 24, 2016, which included reactor coolant pump seal failure with loss of normal heat sink requiring implementation of feed and bleed cooling. The inspectors evaluated operator performance during the simulated event and verified completion of risk significant operator actions, including the use of abnormal and emergency operating procedures.

The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the control room supervisor. The inspectors verified the accuracy and timeliness of the emergency classification made by the shift manager and the TS action statements entered by the shift technical advisor. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document

crew performance problems.

b. Findings

No findings were identified.

.2 Quarterly Review of Unit 3 Licensed Operator Requalification Testing and Training

(71111.11Q - 1 sample)

a. Inspection Scope

The inspectors observed a Unit 3 licensed operator simulator requalification training evaluated scenario on May 24, 2016, which included failure of a pressurizer pressure instrument, charging pump trip, loss of 480V safety bus 5A, a small break loss-of-coolant accident (LOCA), and entry into FR-C.2 core cooling. The inspectors evaluated operator performance during the simulated event and verified completion of risk significant operator actions, including the use of abnormal and emergency operating procedures.

The inspectors assessed the clarity and effectiveness of communications, implementation of actions in response to alarms and degrading plant conditions, and the oversight and direction provided by the control room supervisor. The inspectors verified the accuracy and timeliness of the emergency classification made by the shift manager and the TS action statements entered by the shift technical advisor. Additionally, the inspectors assessed the ability of the crew and training staff to identify and document

crew performance problems.

b. Findings

No findings were identified.

.3 Quarterly Review of Licensed Operator Performance

Main article: IP 71111.11Q

a. Inspection Scope

The inspectors conducted a focused observation of operator performance in the main control room. The inspectors observed pre-job briefings and control room communications to verify they met the criteria specified in Entergy's administrative procedure EN-OP-115, "Conduct of Operations." Additionally, the inspectors observed restoration activities to verify that procedure use, crew communications, and coordination of activities between work groups similarly met established expectations and standards.

Unit 2 Plant startup from RFO 2R22 on June 16, 2016 including response to a turbine trip without a reactor trip and the subsequent turbine-generator synchronization and transfer of plant electrical loads from offsite power to the unit auxiliary transformer. Reactor startup and grid synchronization conducted on June 27, 2016.

Unit 3 Operator response to the feedwater transient which occurred on April 26, 2016

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

Main article: IP 71111.12Q

.1 Routine Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed the samples listed below to assess the effectiveness of maintenance activities on SSCs performance and reliability. The inspectors reviewed system health reports, CAP documents, maintenance WOs, and maintenance rule basis

documents to ensure that Entergy was identifying and properly evaluating performance problems within the scope of the maintenance rule. For each SSC sample selected, the inspectors verified that the SSC was properly scoped into the maintenance rule in accordance with 10 CFR 50.65 and verified that the (a)(2) performance criteria established by Entergy was reasonable. As applicable, for SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that Entergy was identifying and addressing common cause failures that occurred within and across maintenance rule system boundaries.

Unit 2 EDGs Unit 3 EDGs (this sample was part of an in-depth review of the EDG system) Units 2 and 3 CVCS

b. Findings

No findings were identified.

URI Opened, CVCS Goal Monitoring Under the Maintenance Rule Introduction The inspectors identified issues of potential concern with Entergy's application of 10 CFR 50.65(a)(1), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Plants," (the maintenance rule) in regards to the reliability of the Unit 2 CVCS system. These concerns included the establishment of appropriate (a)(1) goals and whether appropriate justification was established that the corrective actions to address identified maintenance weaknesses were effective prior to removal from (a)(1) status.

Specifically, Entergy may have established restrictive goals without defensible justification and may not have demonstrated their chosen goal before ending the goal monitoring interval.

Description

The maintenance rule requires that licensees shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. EN-DC-206, "Maintenance Rule (a)(1) Process," provides the requirements and processes for managing SSCs for which (a)(2) monitoring has not demonstrated effective maintenance. EN-DC-206 specifies that (a)(1) action plans should not be closed until effectiveness of all corrective actions has been demonstrated by meeting performance goals through the monitoring period (or by other means specified in the action plan).

Since 2013, there have been several repeat functional failures of equipment in the CVCS resulting in a failure to meet the performance criterion for reliability. These failures included:

A failure of the 23 charging pump on August 6, 2013, after the internal oil pump discharge tubing broke causing the pump to trip on low oil pressure and a loss of charging. The 21 charging pump had tripped for the same reason in 2010. A failure of the 22 charging pump on January 14, 2014, due to cracked internal check valves caused by an inadequate fill-and-vent that left air in the pump following maintenance. The 21 charging pump had failed due to the same cause in 2013. A failure of the Unit 2 valve FCV-110A, boric acid flow control valve, to fully open on January 5, 2015. The valve had insufficient insulation; and as a result, boron crystalized above the valve plug and blocked its movement. The Unit 3 FCV-110A had failed in the same way in 2011, with earlier failures of other valves for the same cause going back to 1997.

In each case, the CVCS for Unit 2 was already (a)(1), so Entergy either updated the existing (a)(1) action plan or created another one to operate in parallel with the existing one. Upon reviewing the associated (a)(1) action plans, the inspectors noted that in each example Entergy's goals may not have been in accordance with EN-DC-206(a)(1) Process. It specifies that monitoring intervals should be at least six months for normally operating SSCs, at least three surveillances for SSCs monitored by surveillance and long enough to detect recurrence of the applicable failure mechanism. It also states that performance goals that provide reasonable assurance that the SSC is capable of performing its intended functions should be monitored throughout the time the SSC is classified (a)(1). EN-DC-206 defines an SSC as any discreet component grouping that has caused a monitoring failure, including any applicable extent of condition. In the

examples provided, NRC inspectors challenged whether Entergy either chose a shorter monitoring interval or a goal that did not include the applicable extent of condition. Specifically:

The (a)(1) action plan for the broken oil tubing had a goal of no noticeable decrease in 23 charging pump's running oil pressure for the next three quarterly surveillances. The chosen monitoring interval met the procedural expectation, but Entergy limited the monitoring to the 23 charging pump without written justification, when the 21 charging pump had failed previously for the same reason and the other pumps were susceptible to the same failure mechanism.

During the monitoring interval, the 21 charging pump experienced low oil pressure. When Entergy performed repairs on the 21 charging pump for an unrelated issue, they discovered that the oil tubing had failed in the same way the 23 charging pump oil tubing had failed, although it had not yet caused a pump trip. The (a)(1) action plan for the cracked check valves had a goal of no check valve failure for six months for the next charging pump that underwent maintenance. This happened to be the 22 charging pump. Entergy chose a six-month monitoring

interval, even though only one of the three char ging pumps is in service at any given time, and the 22 charging pump only ran for four out of the six months it was monitored. Additionally, the action plan did not justify why a single successful fill-and-vent demonstrated adequate corrective actions. On November 19, 2014, during the six month monitoring interval, the 21 charging pump underwent maintenance

requiring a fill-and-vent, and experienced check valve failure two weeks later on December 4. Entergy documented this as a maintenance rule functional failure, and discussed the possibility that it could be due to an inadequate fill-and-vent, but did not change the (a)(1) action plan. The (a)(1) action plan for FCV-110A specified a monitoring interval of six months to include the winter because the previous valve failures had all occurred during the winter months. However, the actual monitoring interval documented in the corrective action was from April to October 2015, and therefore did not cover the winter months as intended. In January 2016, Entergy performed maintenance on valve CH-297 on Unit 3, which is a heat-traced boric acid valve, and did not properly restore the insulation. The valve function was not impacted because it does not often contain high concentrations of boric acid.

The (a)(1) action plans described above were all reviewed and approved by the maintenance rule expert panel.

Further information regarding the performance of these SSCs is required to determine whether these issues of concern represent performance deficiencies and whether they

are more than minor. (URI 05000247/2016002-01, CVCS Goal Monitoring Under the Maintenance Rule)

.2 Quality Control

a. Inspection Scope

The inspectors reviewed the weld repair performed on the 21 CCW heat exchanger service water inlet nozzle for Unit 2 to verify Entergy was properly applying quality controls specified in their quality assurance program. The inspectors reviewed CAP documents, maintenance WOs, ECs, and engineering procedures associated with the weld repair. The inspectors verified Entergy specified quality control hold points in accordance with their procedures, properly controlled the quality of materials used during the repair, and adequately justified deviations from the existing design.

Additionally, the inspectors reviewed the welding procedure specification qualification by the vendor to ensure it was in accordance with American Society of Mechanical Engineers code.

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

Main article: IP 71111.13

a. Inspection Scope

The inspectors reviewed station evaluation and management of plant risk for the maintenance and emergent work activities listed below to verify that Entergy performed the appropriate risk assessments prior to removing equipment for work. The inspectors selected these activities based on potential risk significance relative to the reactor safety cornerstones. As applicable for each activity, the inspectors verified that Entergy performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When Entergy performed emergent work, the inspectors verified that operations personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work and discussed the results of the assessment with the station's probabilistic risk analyst to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the TS requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.

Unit 2 Temporary loss of spent fuel pool cooling due to 345-kilovolt disturbance on April 3, 2016 Equipment hatch closure plug seal demonstration for outage risk on April 5, 2016 Reduced inventory operations during vessel reassembly on June 7, 2016 21 CCW heat exchanger OOS during mode 4 on June 25, 2016 Unit 3 32 EDG OOS while Bus Tie BT 4-5 was OOS on May 4, 2016 (this sample was part of an in-depth review of the EDG system) 33 EDG OOS while Bus Tie BT 4-5 was OOS on June 2, 2016 31 EDG OOS while Bus Tie BT 4-5 was OOS on June 21, 2016

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

Main article: IP 71111.15

a. Inspection Scope

The inspectors reviewed operability determinations for the following degraded or non-conforming conditions:

Unit 2 23 EDG failure to run on March 7, 2016, and subsequent failure to pass the surveillance test on March 10, 2016, as identified in CR-IP2-2016-01260 Operability determination for N33 gamma metrics wide range nuclear instrument channel in CR-IP2-2016-03660 on June 13, 2016 Pressurizer level transmitter LT-461 reads high in CR-IP-2016-3806 on June 14, 2016 Through-wall leak in line 411, service water inlet to the 21 CCW heat exchanger, on June 15, 2016

Unit 3 Immediate operability determination of the degraded condition of the baffle-former bolts identified from Unit 2 operating experience in CR-IP3-2016-01035 on April 1, 2016 Anomalies noted during digital metal impact monitoring system self-test in CR-IP3-2015-03468 on April 1, 2016 Prompt operability determination of the degraded condition of the baffle-former bolts identified from Unit 2 operating experience in CR-IP2-2016-03660 on June 30, 2016 The inspectors selected these issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the operability determinations to assess whether TS operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TSs and UFSAR to Entergy's evaluations to determine whether the components or systems were operable.

The inspectors confirmed, where appropriate, compliance with bounding limitations associated with the evaluations. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled by Entergy. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations.

b. Findings

No findings were identified.

Update to URI 05000247/2016001-06, 23 Emergency Diesel Generator Automatic Voltage Regulator Failure

Introduction.

The NRC opened a URI in Inspection Report 05000247/2016001 related to two failures of the 23 EDG to run and maintain bus voltage on March 7, 2016, and to provide adequate control of bus voltage on March 10, 2016. This report provides an update of the status of this URI.

Description.

On March 7, 2016, approximately one hour after the trip of the 3A normal feed breaker, the 23 EDG tripped on overcurrent while powering the 6A 480V safety bus.

The 6A bus remained de-energized for approximately one hour until the crew restored the 6A bus via off-site power. The 23 EDG was declared inoperable. All four 480V safety buses were restored to off-site power. Entergy replaced the overcurrent relays and retested the 23 EDG satisfactorily on March 8, 2016. However, bench testing of the overcurrent relays demonstrated that they were accurately calibrated.

Subsequently, on March 10, 2016, during performance of PT-R14, "Automatic Safety Injection System Electrical Load and Blackout Test," the 23 EDG exhibited anomalous behavior during the train 'B' load sequencing. During this test, the voltage on safety bus 6A dropped to approximately 200V when the 23 auxiliary feedwater pump was sequenced onto the bus (CR-IP2-2016-01430) and the sequencer failed to complete the first two sequences. The 23 EDG was again declared inoperable and the period of inoperability was backdated to March 7, 2016, when it originally tripped. Further troubleshooting and additional failure modes analysis by Entergy initially determined that the cause of both events may have been a degraded resistor (R25) on the 23 EDG automatic voltage regulator (AVR) card.

The 23 EDG AVR card was replaced, and the 23 EDG was again tested satisfactorily.

The voltage anomaly issues exhibited during the March 10, 2016, test were documented in CR-IP2-2016-01430 which was closed in CR-IP2-2016-01260 to be included in the causal assessment associated with the tripping of 23 EDG breaker on March 7, 2016. Entergy assigned a vendor to perform laboratory bench testing and failure analysis of the 23 EDG AVR card. The vendor report attributed the cause of the March 10, 2016, loss of voltage control to a degraded solder joint on the AVR card. However, the vendor report explicitly did not attribute the event on March 7, 2016, to the same cause.

Entergy assigned a corrective action in CR-IP2-2016-01260 to review the cause of the 23 EDG overcurrent trip on March 7, 2016, in light of the vendor report. The inspectors determined that the issue of concern remains open as a URI until this causal assessment has been completed by Entergy and assessed by NRC.

(URI 05000247/2016001-06, 23 Emergency Diesel Generator Automatic Voltage Regulator Failure)

1R18 Plant Modifications

Main article: IP 71111.18

Permanent Modifications

.1 Control Rod Guide Tube Repairs in Location E-9

a. Inspection Scope

The inspectors evaluated a modification to the reactor vessel upper internals to swap damaged control rod guide tube in location E-9 with abandoned guide tube in location D-10. The inspectors verified that the design bases, licensing bases, and performance capability of the affected systems were not degraded by the modification. In addition, the inspectors reviewed modification documents associated with the design change, including evaluation of equivalency and core flow changes, and post-modification testing. The inspectors also reviewed revisions to the affected drawings and interviewed refueling and engineering personnel.

b. Findings

No findings were identified.

.2 Core Baffle-Former Bolt EC 64038

a. Inspection Scope

The inspectors reviewed EC 64038, "IP2 Reactor Vessel Equivalent Replacement Baffle-to-Former Bolt." This modification was completed during RFO 2R22 and involved the replacement of 278 baffle-former bolts out of a total of 832 located in the Unit 2 reactor vessel. Entergy replaced all of the bolts that were potentially degraded as observed by visual indications of a protruding bolt head or lock bar problem, bolts that did not pass UT, and bolts inaccessible for UT. Entergy staff also replaced 51 additional bolts that passed ultrasonic and visual examinations to increase the structural margin of the baffle-former assembly for future operating cycles.

The inspectors reviewed the equivalency evaluation completed by Entergy staff to install baffle-former bolts of a different material and configuration than the original bolts. The inspectors reviewed the associated EC package to determine whether the replacement bolts' form, fit, and function were maintained compared to the original bolts and whether the change conformed to the design and licensing bases of the baffle-former assembly.

Specifically, this change involved replacing the original baffle-former bolts made of type 347 stainless steel with bolts made of type 316 stainless steel. The baffle-former bolt head configuration was also changed from an original internal hex and slot design (secured with a welded lock bar) to an external hex configuration with an integral locking cup design. The design change document further evaluated a more gradual fillet geometry between the bolt head and shank intended to reduce the stress concentration at that transition and provide for improved fatigue resistance.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing

Main article: IP 71111.19

a. Inspection Scope

The inspectors reviewed the post-maintenance tests for the maintenance activities listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedure to verify that the procedure adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure was consistent with the information in the applicable licensing basis and/or design basis documents, and that the test results were properly reviewed and accepted and problems were appropriately documented. The inspectors also walked down the affected job site, observed the pre-job brief and post-job critique where possible, confirmed work site cleanliness was maintained, witnessed the test or reviewed test data to verify quality control hold points were performed and checked, and that results adequately demonstrated restoration of the affected safety functions.

Unit 2 21 EDG fuel oil transfer pump after planned maintenance on May 5, 2016 Replacement of pressurizer level transmitters LT-459 and LT-460 on May 25, 2016 21 CCW heat exchanger service water outlet weld repair on June 26, 2016 Flux mapping system drive repairs following motor failures on June 28, 2016 Unit 3 Maintenance on service water components associated with the 32 EDG on May 5, 2016 (this sample was part of an in-depth review of the EDG system) Modification of the 32 EDG space heaters on May 5, 2016 (this sample was part of an in-depth review of the EDG system) Maintenance on the 32 EDG air start system on May 6, 2016 (this sample was part of an in-depth review of the EDG system) Replacement of failed bistable LC-427K, steam generator 32 low level mismatch trip interlock, on May 18, 2016

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

Main article: IP 71111.20

.1 Unit 2 RFO 2R22

a. Inspection Scope

The inspectors reviewed the station's work schedule and outage risk plan for the Unit 2 maintenance during RFO 2R22, which was conducted from March 7, 2016, to June 16, 2016. The inspectors reviewed Entergy's development and implementation of outage plans and schedules to verify that risk, industry experience, previous site-specific problems, and defense-in-depth were considered. During the outage, the inspectors observed portions of the shutdown and cooldown processes and monitored controls associated with the following outage activities:

Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TSs when taking equipment OOS Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work or testing Installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication and instrument error accounting Status and configuration of electrical systems and switchyard activities to ensure that TSs were met Monitoring of decay heat removal operations Impact of outage work on the ability of the operators to operate the spent fuel pool cooling system Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss Activities that could affect reactivity Maintenance of secondary containment as required by TSs Refueling activities, including fuel handling and fuel receipt inspections Fatigue management Tracking of startup prerequisites, walkdown of the primary containment to verify that debris had not been left which could block the ECCS suction strainers, and startup and ascension to full power operation Foreign Object Search and Retrieval for missing baffle bolts and locking tabs Identification and resolution of problems related to RFO activities During this outage, Entergy replaced 278 degraded baffle bolts in the Unit 2 reactor vessel baffle assembly. This emergent project resulted in the extension of the outage schedule from 30 days to 102 days.

b. Findings

Introduction.

The inspectors identified a Green NCV of TS 5.4.1 for Entergy's failure to implement procedure OAP-007, "Containment Entry and Egress." Specifically, workers transiting the inner and outer crane wall sections of containment on June 11, 2016, failed to maintain at least one (of two) flow channeling gate closed to ensure availability of the containment sumps to provide suction for the ECCS.

Description.

On June 11, 2016, in mode 4, in preparation for reactor startup, Entergy was performing maintenance in containment required prior to mode 3, such as reactor coolant pump motor balancing and steam flow transmitter troubleshooting. These activities required scaffolds to be temporarily erected for workers to safely perform maintenance. While transiting from the inner to outer section of containment, the inspectors noted that both flow channeling gates were maintained open simultaneously as workers carried scaffold poles and hardware out of the area.

In the event of a postulated LOCA, Unit 2 relies on two sumps to provide a suction source for the internal recirculation pumps and residual heat removal pumps, respectively, after the injection phase of the accident. The sumps have cylindrical screens with large surface area and small holes to filter small debris and maintain adequate net positive suction head for the associated pumps. The reactor cavity sump and large intervening barriers prevent large debris generated from the accident, such as insulation, from reaching and blocking the recirculation and containment sump screens.

Entergy procedure OAP-007, "Containment Entry and Egress," precaution and limitation step 2.30.2, states, "In mode 1, 2, 3, or 4, entry inside the crane wall shall use the double gate entry point via gates 17 and 23. One gate shall remain shut and secured at all times to maintain flow channeling and sump operability. Securing gates requires a padlock or nut and bolt closure from the outside. This will require posting a gate monitor to allow exit." The inspectors noted, while a gate monitor was posted, both gates were maintained open during passage and not secured with a padlock or nut and bolt closure.

Upon questioning by the inspectors, Entergy immediately coached the gate monitor and restored the gates to an acceptable position. Entergy generated CR-IP2-2016-04036 to address this issue.

Analysis.

The inspectors determined that Energy's failure to maintain either gate 17 or gate 23 closed during passage in accordance with OAP-007 was a performance deficiency. The performance deficiency was more than minor because it is associated with the configuration control (shutdown equipment lineup) attribute and adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated the finding in accordance with IMC 0609, Appendix G, Attachment 1, Exhibit 3, and determined that a detailed risk evaluation was necessary because the finding represented a loss of system safety function. A detailed risk assessm ent was conducted conservatively assuming complete failure of the recirculation and containment sumps due to the performance deficiency. Given that Unit 2 was in mode 4, in plant operating state 1, with a late time window, the at-power simplified plant analysis risk model for large-break LOCAs was determined to best model the degrade condition and plant response. An exposure time of one day was assumed. No credit was assumed for the decrease in energy that would be anticipated in a release during a LOCA in mode 4, nor the corresponding reduction in debris generation. This was also considered conservative. Utilizing Systems Analysis Program for Hands-On Integrated Reliability Evaluation, version 8.13, with Indian Point Unit 2 Simplified Plant Analysis Risk Model, version 8.19, for the assumed conditions, the change in core damage frequency was determined to be 7E-9. Therefore, this issue represents a Green finding.

This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Entergy did not consider potential undesired consequences of actions before performing work and implement appropriate error-reduction tools. Specifically, the work crew did not understand the requirements and potential consequences prior to commencing work and the gate monitor did not enforce these requirements to maintain at least one gate locked or pinned closed as required by OAP-007. [H.12]

Enforcement.

Unit 2 TS 5.4.1.a requires that the procedures listed in Attachment A to Regulatory Guide 1.33, "Quality Assurance Program Requirements," Revision 2, be established and implemented. Attachment A states that instructions should be prepared, as appropriate, for access to containment and changing modes of operation of the ECCS. Entergy procedure OAP-007, "Containment Entry and Egress," Step 2.30.2, states, "In mode 1, 2, 3, or 4, entry inside the crane wall shall use the double gate entry point via gates 17 and 23. One gate shall remain shut and secured at all times to maintain flow channeling and sump operability. Securing gates requires a padlock or nut and bolt closure from the outside." Contrary to the above, on June 11, 2016, Entergy did not maintain one gate secured at all times with a padlock or nut and bolt closure.

Entergy entered this issue into the CAP as CR-IP2-2016-04036. Because this violation was of very low safety significance (Green), and Entergy entered this performance deficiency into the CAP, the NRC is treating this as a NCV in accordance with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000247/2016002-03, Failure to Maintain Flow Channeling Gates Closed in Accordance with the Containment Procedure)

.2 Unit 2 Forced Outage

a. Inspection Scope

Unit 2 conducted a forced outage from June 24 to 27, 2016, in order to complete weld repairs on a through-wall leak on the service water inlet line to the 21 CCW heat exchanger. These repairs required shutting down to mode 4 in order to meet the TS 3.7.7, "Component Cooling Water (CCW) System," limiting condition for operations for CCW operability. While these repairs were being completed, the grid operator completed repairs to breaker 9 in the offsite switchyard. During the outage, the inspectors observed portions of the shutdown and cooldown processes and monitored controls associated with the following outage activities:

Configuration management, including maintenance of defense-in-depth, commensurate with the outage plan for the key safety functions and compliance with the applicable TSs when taking equipment OOS Implementation of clearance activities and confirmation that tags were properly hung and that equipment was appropriately configured to safely support the associated work or testing Status and configuration of electrical systems and switchyard activities to ensure that TSs were met Monitoring of decay heat removal operations Reactor water inventory controls, including flow paths, configurations, alternative means for inventory additions, and controls to prevent inventory loss Activities that could affect reactivity Tracking of startup prerequisites Identification and resolution of problems related to RFO activities When all repairs had been completed, Entergy restarted Unit 2 on June 27, 2016.

b. Findings

No findings were identified.

1R22 Surveillance Testing (71111.22

- 6 samples)

a. Inspection Scope

The inspectors observed performance of surveillance tests and/or reviewed test data of selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR, and Entergy's procedure requirements. The inspectors verified that test acceptance criteria were clear, tests demonstrated operational readiness and were consistent with design documentation, test instrumentation had current calibrations and the range and accuracy for the application, tests were performed as written, and applicable test prerequisites were satisfied. Upon test completion, the inspectors considered whether the test results supported that equipment was capable of performing the required safety functions. The inspectors reviewed the following surveillance tests:

Unit 2 WO 446385, 21 EDG AVR card inspection, on May 24, 2016 2-PT-Q013 for containment isolation valve 851B (22 safety injection (SI) pump tie to 23 SI pump discharge) on June 6, 2016 2-PT-Q029B quarterly in-service surveillance test for the 22 SI pump on June 6, 2016 Unit 3 3-PT-M079B 32 EDG monthly surveillance on May 30, 2016 (this sample was part of an in-depth review of the EDG system) 34 steam generator pressure instrument channel check on June 21, 2016 0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation and Leak Identification, beginning on June 28, 2016

b. Findings

No findings were identified.

Cornerstone:

Emergency Preparedness

1EP6 Drill Evaluation

Main article: IP 71114.06

Training Observations

a. Inspection Scope

The inspectors evaluated the conduct of Entergy's ingestion pathway emergency preparedness drill on April 19, 2016, to identify any weaknesses and deficiencies in the classification, notification, and protective action recommendation development activities.

The inspectors observed emergency response operations in the emergency operations facility to determine whether the event classification, notifications, and protective action recommendations were performed in accordance with procedures. The inspectors also attended the facility drill critique to compare inspector observations with those identified by Entergy staff in order to evaluate Entergy's critique and to verify whether the staff was properly identifying weaknesses and entering them into the CAP.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstone:

Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls

Main article: IP 71124.01

a. Inspection Scope

During May 10-12 and June 13-17, 2016, the inspectors reviewed Entergy's performance in assessing the radiological hazards and exposure control in the workplace. The inspectors used the requirements in 10 CFR 20, TSs, applicable industry standards, and procedures required by TSs as criteria for determining compliance.

Radiological Hazards Control and Work Coverage The inspectors reviewed:

Ambient radiological conditions during tours of the radiological controlled area, posted surveys, radiation work permits, adequacy of radiological controls, radiation protection job coverage, and contamination controls Controls for highly activated or contaminated materials stored within spent fuel pools Posting and physical controls for high radiation areas and very high radiation areas

b. Findings

No findings were identified.

2RS2 Occupational As Low As Is Reasonably Achievable (ALARA) Planning and Controls

(71124.02)

a. Inspection Scope

During May 10-12 and June 13-17, 2016, the inspectors assessed performance with respect to maintaining occupational individual and collective radiation exposures ALARA. The inspectors used the requirements in 10 CFR 20, TSs, applicable industry standards, and procedures required by TSs as criteria for determining compliance.

Radiological Work Planning The inspectors reviewed:

ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements ALARA work planning, use of dose mitigation features and dose goals Work planning and the integration of ALARA requirements

b. Findings

No findings were identified.

2RS7 Radiological Environmental Monitoring Program (REMP)

Main article: IP 71124.07

a. Inspection Scope

The inspectors reviewed the REMP to validate the effectiveness of the radioactive gaseous and liquid effluent release program and implementation of the groundwater protection initiative (GPI). The inspectors used the requirements in 10 CFR 20, 40 CFR 190, 10 CFR 50, Appendix I, TSs, offsite dose calculation manual (ODCM), Nuclear Energy Institute 07-07, and procedures required by TSs as criteria for determining compliance.

Inspection Planning

The inspectors reviewed Entergy's 2014 and 2015 annual radiological environmental and effluent monitoring reports, REMP program audits, ODCM changes, land use census, the UFSAR, and inter-laboratory comparison program results.

Site Inspection The inspectors walked down various thermoluminescent dosimeter and air and water sampling locations and reviewed associated calibration and maintenance records. The inspectors observed the sampling of various environmental media as specified in the ODCM and reviewed any anomalous environmental sampling events including assessment of any positive radioactivity results. The inspectors reviewed any changes to the ODCM. The inspectors verified the operability and calibration of the meteorological tower instruments and meteorological data readouts. The inspectors reviewed environmental sample laboratory analysis results, laboratory instrument measurement detection sensitivities, laboratory quality control program audit results, and the inter- and intra-laboratory comparison program results. The inspectors reviewed the groundwater monitoring program as it applies to selected potential leaking SSCs.

GPI Implementation The inspectors reviewed groundwater monitoring results, changes to the GPI program since the last inspection, anomalous results or missed groundwater samples, leakage or spill events including entries made into the decommissioning files (10 CFR 50.75(g)), evaluations of surface water discharges, and Entergy's evaluation of any positive groundwater sample results including appropriate stakeholder notifications and effluent reporting requirements.

Identification and Resolution of Problems The inspectors evaluated whether problems associated with the REMP were identified at

an appropriate threshold and properly addressed in Entergy's CAP.

b. Findings

No findings were identified.

OTHER ACTIVITIES

4OA1 Performance Indicator Verification

Main article: IP 71151

Initiating Events Performance Indicators

a. Inspection Scope

The inspectors reviewed Entergy's submittals for the following Initiating Events cornerstone performance indicators for the period April 1, 2015, to March 31, 2016:

Unit 2 Unplanned scrams per 7000 critical hours (IE01) Unplanned power changes per 7000 critical hours (IE03) Unplanned scrams with complications (IE04)

Unit 3 Unplanned scrams (IE01) Unplanned power changes (IE03) Unplanned scrams with complications (IE04)

To determine the accuracy of the performance indicator data reported during those periods, inspectors used definitions and guidance contained in Nuclear Energy Institute 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7.

The inspectors reviewed Entergy's operator narrative logs, maintenance planning schedules, CRs, event reports, and NRC integrated inspection reports to validate the accuracy of the submittals. There were no unplanned power changes or scrams with complications during the review period.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

Main article: IP 71152

.1 Routine Review of Problem Identification and Resolution Activities

a. Inspection Scope

As required by Inspection Procedure 71152, "Problem Identification and Resolution," the

inspectors routinely reviewed issues during baseline inspection activities and plant status reviews to verify that Entergy entered issues into the CAP at an appropriate threshold, gave adequate attention to timely corrective actions, and identified and addressed adverse trends. In order to assist with the identification of repetitive equipment failures and specific human performance issues for follow up, the inspectors performed a daily screening of items entered into the CAP and periodically attended CR screening meetings. The inspectors also confirmed, on a sampling basis, that, as applicable, for identified defects and non-conformances, Entergy performed an evaluation in accordance with 10 CFR 21.

b. Findings

No findings were identified.

.2 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a semi-annual review of site issues, as required by Inspection Procedure 71152, "Problem Identification and Resolution," to identify trends that might indicate the existence of more significant safety issues. In this review, the inspectors included repetitive or closely-related issues that may have been documented by Entergy outside of the CAP, such as trend reports, performance indicators, major equipment problem lists, system health reports, maintenance rule assessments, and maintenance or CAP backlogs. The inspectors also reviewed Entergy's CAP database for the first and second quarters of 2016 to assess CRs written in various subject areas (equipment problems, human performance issues, etc.), as well as individual issues identified during the NRCs daily CR review (Section 4OA2.1). The inspectors reviewed the Entergy quarterly trend report for the first quarter of 2016 to verify that Entergy was appropriately evaluating and trending adverse conditions in accordance with applicable procedures.

b. Findings and Observations

No findings were identified.

The inspectors identified a trend in work being performed that was contrary to written work instructions and procedures, and work packages had been closed out without

documenting the deviation from the work order. While reviewing completed work order WO 447966 Task 10, Internal Coating Repair for 21 CCW HX, the inspectors found a note in the work order stating that the internal coating repair to the pipe had not been done in accordance with the engineering change. The engineering change had been written when the coating repair was expected to be small, but the actual area that was recoated was much larger. A larger area of coating increases the impact on the heat exchanger if the coating were to flake off and block the flow of service water. The work package was closed and no condition report was written. This performance deficiency is minor because the coating was applied with procedurally directed quality controls and the likelihood that it would flake off is very small; and is the same as the original smaller area specified in the work package. However, the work package was closed without documenting the deviation and no CR was written.

In another example, the inspectors noted that WO 412920 Task 15 to perform a surge test on 11 centrifugal air compressor (CENTAC) after overhaul was completed on December 22, 2015. However, the completion notes and documentation for the task showed that the test was unable to be performed due to a test equipment problem. The work package was closed and no CR was written. Subsequently, after being returned to service, the compressor failed in service due to multiple surging events on January 7, 2016. Troubleshooting under WO 433939 revealed that the motor high load limit had not been adjusted to account for the increased load due to reduced compressor clearances introduced by the overhaul. This performance deficiency is screened to minor because the 11 CENTAC is not a safety-related piece of equipment and would not affect the MC 0609 cornerstone thresholds or other generic criteria. Unit 2 and Unit 3 have dedicated instrument air compressors that are credited in the FSAR to respond to a loss of instrument air event. If the 11 CENTAC (located in Unit 1) were to fail, the unit-specific IACs would automatically start to prevent a loss of instrument air at both Units 2 and 3.

A third recent example of work being performed contrary to written instructions occurred during 2RFO22 when the inspectors identified that the workers deviated from the surveillance procedure by demonstrating the installation of the emergency containment hatch plug without properly inflating the plug seals as directed by the procedure. This performance deficiency was previously documented in a prior inspection report as non-cited violation 05000247/05000286/2016001-02, "Failure to Adequately Implement Risk Management Actions for the Containment Key Safety Function."

In all cases, the deviations from written work instructions were directed by Entergy supervision. In addition, the inspectors noted that Entergy had self-identified similar observations where work packages or condition reports had been closed without fully completing the specified actions including CR-IP2-2015-05833, CR-IP2-2016-00103, CR-IP3-2015-04729, CR-IP3-2016-00072, CR-IP3-2016-00075, and CR-IP3-2015-04019. These CRs are further examples of work orders that were closed with deviations that were not documented or resolved. Nuclear Oversight had identified several of these condition reports. Entergy has taking immediate corrective action in response to these performance deficiencies.

.3 Annual Sample:

Maintenance Rule Self-Assessment of Corrective Actions

a. Inspection Scope

The inspectors performed an in-depth review of Entergy's corrective actions associated with self-assessment LO-IP3LO-2015-72, "Maintenance Rule (a)(3) Assessment." The self-assessment was performed to satisfy both the self-assessment criteria in EN-LI-104, "Self-Assessment and Benchmark Process," and the maintenance rule periodic assessment criteria in EN-DC-207.

The inspectors assessed Entergy's problem identification threshold, extent of condition reviews, and the prioritization and timeliness of Entergy corrective actions to determine whether Entergy was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of Entergy's CAP and 10 CFR 50, Appendix B. In addition, the inspectors interviewed engineering personnel to assess the effe ctiveness of the im plemented corrective actions.

b. Findings and Observations

No findings were identified.

Entergy identified three standard deficiencies during their self-assessment and wrote CRs to document each one. One of the standard deficiencies was that the maintenance rule basis documents were not being reviewed at least once every two years as required by procedure EN-DC-204, "Maintenance Rule Scope and Basis." The purpose of this review was to ensure that the documents were updated if the configuration of the system changed or if the performance criteria needed to be adjusted. Entergy wrote CR-IP3-2015-03628 and assigned a corrective action to create work trackers to perform the basis document reviews. They chose to use work trackers instead of corrective actions under the CAP because the work had historically been assigned using work trackers. However, because work trackers do not receive the same priority as corrective actions, some of the maintenance rule basis documents had still not been reviewed at the time of this inspection, over a year after the completion of the self-assessment. The inspectors determined that this was not a more than minor issue because the systems in question did not show signs of inadequate maintenance.

.4 Annual Sample:

Unit 2 Reactor Trip on December 5, 2015

a. Inspection Scope

The inspectors performed an in-depth review of Entergy's evaluations and corrective actions associated with CR-IP2-2015-05484 and the related apparent cause evaluation for the December 5, 2015, manual reactor trip in response to indications of multiple dropped control rods caused by the loss of control rod power due to a power supply failure. Entergy performed an apparent cause evaluation and determined the direct cause of the event was the loss of motor control center (MCC)-24 due to an internal fault at the line side leads at cubicle 2H where they connect to the bucket stab assemblies. The apparent cause was an unanticipated loss of power to the control rod system due to the degradation of the primary control rod power supply (PS1) which failed to function for more than 10 minutes when the operating alternate power supply (PS2) was deenergized.

The inspectors assessed Entergy's problem identification threshold, problem analysis, extent of condition reviews, compensatory actions, and the prioritization and timeliness of Entergy's corrective actions to determine whether Entergy was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of Entergy's CAP and 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action."

b. Findings and Observations

No findings were identified.

The inspectors found that Entergy took appropriate actions to identify the direct and apparent cause of the issue. The direct cause of the event was the loss of MCC-24 due to an internal fault at the line side leads at cubicle 2H where they connect to the bucket stab assemblies. The apparent cause was an unanticipated loss of power to the control rod system due to the degradation of the primary control rod PS1, which failed to function when PS2 was lost. Entergy replaced the degraded rod control PS1; and the MCC-24 compartments were removed to facilitate inspection and testing of the MCC bus, control wires, and MCC internal. PS2 was also restored to operation after the fault was cleared.

The inspector determined that the internal electrical fault that deenergized PS2 and the prior degradation in PS1 was not within Entergy's ability to foresee and prevent.

Therefore, there was no performance deficiency identified. Entergy's overall response to the issue was commensurate with the safety significance, was timely, and the actions taken and planned were reasonable to resolve the failure of the primary control rod PS1.

.5 Annual Sample:

Unexpected Number of Degraded Baffle-Former Bolts Discovered in the Unit 2 Reactor Pressure Vessel

a. Inspection Scope

The inspectors performed an in-depth review of Entergy's root cause evaluation and corrective actions associated with CR-IP2-2016-02348 for baffle-former ("baffle") bolts found with indications of degradation during the Indian Point Unit 2 RFO 2R22. Entergy performed ultrasonic examinations of the baffle bolts in accordance with their procedures as part of a planned activity. After an unexpected number of degraded baffle bolts were discovered, Entergy staff reported the issue to the NRC as Event Notification 51829 on March 29, 2016, because the as-found number and location of degraded bolts represented an unanalyzed condition. Entergy staff completed corrective actions to replace all of the potentially degraded baffle bolts on Unit 2. Entergy staff further replaced a population of additional bolts that exhibited no indications of degradation and performed an evaluation to determine the potential for baffle bolt failures at Unit 3.

The baffle-former bolts help secure vertical plates (also referred to as baffle plates)inside the reactor vessel, which then forms a structure surrounding the reactor fuel assemblies to orient the fuel and to direct coolant flow through the core. A sufficient number of baffle bolts are required to remain intact to secure the baffle plates in place so as to not affect control rod insertion or impede emergency core cooling flow during postulated accident conditions. Bolt heads that separate and are no longer held in place by bolt lock-tabs can also become a loose parts concern.

The inspectors determined whether Entergy's acceptable baffle bolt pattern analysis for Unit 2 was completed in accordance with the NRC-approved methodology and provided appropriate structural margin for the next cycle of operation to ensure the Unit 2 baffle plates will remain in place during both normal operation and limiting postulated accident conditions. The inspectors further determined whether Entergy's evaluations of the baffle bolts installed in Indian Point Unit 3 were technically sufficient to conclude the Unit 3 baffle assembly will perform as intended until the next planned RFO, at which time Entergy plans to examine the bolts. The inspectors reviewed Entergy's procedures for determining the functionality and operability of degraded SSC as they relate to Unit 3. The inspectors further interviewed Entergy engineering personnel and contractor staff to discuss the results of Entergy's technical evaluations and to assess the effectiveness of the implemented and planned corrective actions.

The inspectors assessed Entergy's problem identification threshold, cause analyses, extent of condition, compensatory actions, and the prioritization and timeliness of Entergy's corrective actions to determine whether Entergy staff were properly identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to Entergy's CAP, operability determination process, and the requirements of 10 CFR 50, Appendix B. The inspectors observed portions of baffle bolt replacement activities at Unit 2 and reviewed the final visual examination of the baffle bolts and plates once the work was completed.

b. Findings and Observations

The NRC responded to the initial discovery of an unexpected number of baffle bolts found degraded at Indian Point Unit 2 by implementing a comprehensive inspection plan consisting of various baseline inspection samples to assess the extent of the issue and to determine the necessary NRC actions. A follow-up inservice inspection sample (Refer to Section 1R08) was conducted to review the capability of the non-destructive examination techniques, evaluate the UT results, and observe a portion of bolt replacement activities on-site. A permanent modification sample (Refer to Section 1R18) was conducted to review the design change package and evaluations associated

with the new, replacement baffle bolts. The NRC resident inspectors reviewed Entergy's foreign material controls and loose parts analysis (Refer to Section 1R20) to address the potential for missing bolt heads and concluded it would not impact safe operation of the plant.

NRC Region I based inspectors accompanied by an expert from the NRC Office of Nuclear Reactor Regulation completed an annual problem identification and resolution inspection, documented in this section of the report, to verify that Entergy's evaluations and corrective actions to replace Unit 2 baffle bolts were completed in accordance with an NRC approved methodology to support a conclusion that the Unit 2 baffle assembly meets the plant design basis. The inspectors also determined the adequacy of Entergy's evaluations completed to determine there is a reasonable expectation that the Unit 3 baffle assembly will perform as intended during the current operating cycle.

Entergy staff determined the cause of the degraded baffle bolts was primarily due to irradiation-assisted stress-corrosion cracking (IASCC) in combination with increased fatigue loading on the baffle plates. This cause determination was based on industry operating experience related to baffle-former bolt failure in both foreign and domestic plants. IASCC is a cracking mechanism that occurs over a long period of time when susceptible metals are exposed to neutron radiation from the reactor core and stresses as part of normal design and operation. Entergy staff concluded that failure of a critical number of bolts in a localized area subsequently imposed increased loading on adjacent bolts, which propagated failures and generated the moderate clustered pattern observed in the examination results. No other contributing causes were identified.

The inspectors reviewed Entergy's root cause evaluation and the supporting operating experience related to baffle bolt failures at other plants. The inspectors determined that there is documented evidence in the existing technical literature (including materials testing of bolts from other plants) and operating experience to conclude that the likely cause is IASCC; however, the inspectors found that Entergy staff did not define the cause of the fatigue failure mechanism. The inspectors noted that Entergy staff sent a sample of baffle bolts removed from the reactor pressure vessel to a metallurgical laboratory for detailed failure analysis and materials property testing. Entergy indicated their plans to use the results of the laboratory testing to confirm the likely root cause. The inspectors concluded that Entergy staff conducted an appropriate review to identify the likely causes of the degraded baffle bolts and noted that further test results will be used to confirm these causes.

Following identification of the degraded baffle bolts on Unit 2, Entergy's immediate corrective action was to analyze the as-found condition and begin replacing bolts that either had visual indications of bolt failure (protruding bolt head for example), did not pass UT examination, or were not accessible for UT examination. The as-found number and pattern of these bolts exceeded the acceptance criteria in the plant's analysis that was prepared in advance of the baffle bolt examinations; therefore, Entergy reported this discovery to the NRC as an unanalyzed condition. Entergy staff completed corrective actions to replace all of the 227 potentially degraded baffle bolts, plus an additional 51 bolts for increased structural integrity, for a total of 278 bolts. The inspectors noted the 51 additional bolts were installed in strategic locations to prevent clustering of potential bolt failures during the next operating cycle.

The inspectors determined that Entergy staff performed an acceptable bolt pattern analysis that evaluated the replacement bolt pattern for Unit 2 and modeled the potential for future bolt failures. The inspectors found the results of the analysis accounted for a conservative failure rate of bolts and provided appropriate margin for one cycle of operation. The inspectors verified that Entergy's methodology for its acceptable bolt pattern analyses, including its determination of margin, was consistent with the NRC-approved methodology in topical report WCAP-15029-NP-A (ML15222A882). The inspectors determined that Entergy staff tracked corrective actions to re-examine the Unit 2 baffle bolts during the next planned RFO. The inspectors noted the new baffle bolts were made of a material with improved resistance to IASCC and included an improved design to reduce the stresses at the head to shank transition, both of which are enhancements compared to the original bolts.

As part of an extent of condition assessment, Entergy entered CR-IP3-2016-01035 in its CAP to evaluate the potential for degraded baffle bolts on Unit 3. Entergy operators performed an immediate operability determination and concluded that the baffle assembly was operable. Entergy staff performed a subsequent "extent of condition review" that concluded Unit 3 would experience less baffle bolt degradation than Unit 2 based on several plant factors. Entergy also conducted sensitivity analyses to show acceptable bounding conditions in the event of bolt failures. The inspectors reviewed Entergy's evaluations and noted that Entergy staff concluded there was not a degraded condition at Unit 3. In consideration of the guidance in their operability procedure and operating experience from Unit 2 and other plants, the NRC issued an NCV in this report because Entergy did not perform an operability evaluation for Unit 3 as a follow-up to the immediate determination for the potential impact on supported systems controlled by the TS (Refer to Section 1R15).

As a corrective action, Entergy staff performed an operability evaluation and demonstrated that the Unit 3 baffle former assembly remained operable. The inspectors concluded that this supplemental evaluation provided appropriate technical justification for the continued operation of Unit 3 until the next RFO in spring 2017, at which time Entergy plans to examine the baffle bolts. Entergy also implemented a corrective action as part of an enhancement to plant operations to monitor the RCS for any signs of fuel leakage, which could be an indicator of baffle bolt failures.

The inspectors reviewed Westinghouse Nuclear Safety Advisory Letter NSAL-16-1, which discussed the results of recent baffle-former bolt inspections and provided Westinghouse's recommendations on this issue. The letter described the plants as most susceptible (i.e. Tier 1a) to this degradation as Westinghouse 4-loop reactors limited to those with a down-flow configuration and using Type 347 stainless steel bolts. The inspectors noted the recommendation was to complete UT volumetric examination of the baffle bolts at the next scheduled RFO, and that Entergy had already planned this action for Unit 3. Entergy also planned a long-term corrective action to convert Units 2 and 3 from a "down-flow" baffle configuration to an "up-flow" configuration, which would

significantly reduce the load on baffle-former bolts and provide for increased structural margin of the baffle-former assembly. The inspectors determined Entergy's overall response to the issue was commensurate with the safety significance, was timely, and included appropriate compensatory actions. The inspectors concluded that the actions completed and planned were reasonable to address the ongoing aging management of baffle bolts.

4OA3 Follow Up of Events and Notices of Enforcement Discretion

Main article: IP 71153

.1 Plant Events

a. Inspection Scope

For the plant events listed below, the inspectors reviewed and/or observed plant parameters, reviewed personnel performance, and evaluated performance of mitigating systems. The inspectors communicated the plant events to appropriate regional personnel, and compared the event details with criteria contained in IMC 0309, "Reactive Inspection Decision Basis for Reactors," for consideration of potential reactive inspection activities. As applicable, the inspectors verified that Entergy made appropriate emergency classification assessments and properly reported the event in accordance with 10 CFR 50.72 and 50.73. The inspectors reviewed Entergy's follow-up actions related to the events to assure that Entergy implemented appropriate corrective actions commensurate with their safety significance.

Unit 2 Turbine trip occurred while synchronizing Unit 2 to the grid on June 15, 2016 Shutdown required by TS for repairs to a leak on the 21 CCW heat exchanger service water inlet on June 23, 2016 Unit 3 Rapid power reduction from 100 percent to 45 percent power in response to a loss of both heater drain pumps on May 26, 2016

b. Findings

No findings were identified.

.2 (Closed) Licensee Event Report (LER) 05000247/2015-003-00:

Manual Reactor Trip Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to a Power Supply Failure

The inspector's reviewed Entergy's actions and reportability criteria associated with LER 05000247/2015-003-00, which was submitted to the NRC on February 3, 2016. On December 5, 2015, control room operators initiated a manual reactor trip after observing indications consistent with multiple dropped control rods following an alarm for the trip of MCC-24/24A. No control rod indication was available due to MCC-24 being faulted and de-energized. The direct cause of the event was the loss of MCC-24 due to an internal fault at the line sides leads at cubicle 2H where they connect to the bucket stab assemblies. The apparent cause was an unanticipated loss of power to the control rod system due to the degradation of the primary control rod PS1 which failed to function when the operating PS2 was lost. The inspectors determined that both the unexpected failure of PS2 and the internal fault in PS1 was not within Entergy's ability to foresee and prevent and was not a performance deficiency. The inspectors reviewed the LER, the associated apparent cause evaluation analysis, and interviewed Entergy staff. This LER is closed.

.3 (Closed) LER 05000247/2016-003-00:

TS Prohibited Condition Due to an Inoperable 21 MBFP Discharge Valve for Greater Than the TS Allowed Outage Time

The inspector's reviewed Entergy's actions and reportability criteria associated with LER 05000247/2016-003-00, which was submitted to the NRC on May 6, 2016. On March 7, 2016, during the shutdown to enter 2RFO22, the control switch for the 21 MBFP was tripped from the control room but the MBFP discharge valve BFD-2-21 failed to fully close as designed. The MBFP discharge valve was declared inoperable and TS 3.7.3 Condition C was entered. The MFD-2-21 isolation valve was then manually closed. The direct cause of the failure to close the MBFP discharge valve BFD-2-21 was the motor operated valve's (MOV's) close torque switch contact finger out of position. The apparent cause was that the MOV preventative maintenance procedure lacked the level of detail and direction due to an unrecognized susceptibility associated with the orientation of the close torque switch contact finger bracket opening and spreading of the "U" shape bracket. The downward arrangement made it easier for the torque switch contact finger to move out with spreading of the "U" shaped contact holder. The inspectors reviewed the LER, the associated apparent cause evaluation analysis, and interviewed Entergy staff. This LER is closed.

Introduction.

The inspectors identified a Green NCV of 10 CFR 50.65(b)(1) for Entergy's failure to include a function of a safety-related system within the scope of the maintenance rule program. Specifically, Entergy failed to include the feedwater isolation function performed by the MBFP discharge valves, MBFPs, and feedwater regulating valves and feedwater isolation valves which are required to remain functional during and following a design basis event to mitigate the consequences of an accident, within the scope of the maintenance rule monitoring program.

Description.

On March 7, 2016, during an RFO, the control switch for the 21 MBFP was positioned to trip and the 21 MBFP tripped as designed, but the MBFP discharge valve BFD-2-21 failed to fully close. Entergy declared MBFP discharge valve BFD-2-21 inoperable and entered TS 3.7.3 Condition C. After troubleshooting, Entergy determined the MOV close torque switch contact finger was out of position within the contact holder. The misalignment allowed the contact finger to move out of the proper position causing the MOV BFD-2-21 to fail to close. This is the same failure mechanism which caused MOV BFD-2-21 to fail to close in 2010 which is referenced in CR-IP2-2010-07013. On December 5, 2015, the 21 MBFP failed to trip and required closure of the steam admission valves to secure it. This failure occurred because of contaminated control oil that prevented the solenoid valves from operating.

The inspectors reviewed Entergy's maintenance rule basis documents and identified the feedwater isolation function was not properly included in the maintenance rule monitoring program as required by 10 CFR 50.65(b)(1). The basis document for the feedwater system did identify the need to monitor the feedwater isolation function under the maintenance rule and stated that it would be monitored as a part of the vapor containment supersystem. However, the basis document for the vapor containment supersystem does not include the feedwater isolation components within the system boundaries. As a result, when component failures occurred which affected the feedwater isolation function, they were not reviewed to determine if they were maintenance rule functional failures; and Entergy was unable to identify that the performance of the main feedwater isolation equipment was not effectively controlled through preventative maintenance. Entergy entered this issue into the CAP as CR-IP2-2016-03963 and initiated actions to include the MBFP discharge valves into the maintenance rule program.

Analysis.

The failure to appropriately scope the safety-related feedwater isolation function within the maintenance rule program was a performance deficiency. This finding is more than minor because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by acci dents or events. Specifically, the failure to properly scope the feedwater isolation function prevented Entergy from identifying that equipment reliability was no longer effectively controlled through preventative maintenance. Additionally, this issue is similar to example 7.d described in IMC 0612, Appendix E, "Examples of Minor Issues," dated August 11, 2009. In accordance with IMC 0609.04, "Initial Characterization of Findings," and Exhibit 2 of IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. There are redundant methods of feedwater isolation. They include tripping the MBFPs and closing the MBFP discharge valves, closing the main feedwater regulating valves and low flow bypass valves, and closing the main feedwater isolation valves. On both December 5, 2015, and March 7, 2016, the main feedwater regulating valves and isolation valves were functional; so there was no loss of the ability to isolate feedwater to mitigate accident and transient conditions.

This finding does not have a cross-cutting aspect, since the failure to scope this equipment into the maintenance rule program was not recognized when Entergy combined the maintenance rule basis documents for Units 2 and 3 in 2012 and as a result, is not indicative of current licensee performance.

Enforcement.

10 CFR 50.65(b)(1) requires, in part, that the holders of an operating license shall include within the scope of the monitoring program, specified in 10 CFR 50.65(a)(1), SSCs that are relied upon to remain functional during and following

design basis events. Contrary to the above, since the combined maintenance rule scoping for Units 2 and 3 in 2012, Entergy failed to include within the scope of the monitoring program specified in 10 CFR 50.65(a)(1), the safety-related MBFP discharge valves. These SSCs are relied upon during and after design basis events to mitigate the consequences of a feedwater line break accident inside containment. Entergy's corrective action included entering this issue into the corrective action program.

Because the violation was of very low safety significance (Green) and Entergy entered this issue into their CAP as CR-IP2-2016-03963, this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000247/2016002-04, Failure to Scope Safety-Related Main Boiler Feedwater Pump Discharge Valves into the Maintenance Rule Program)

4OA5 Other Activities

.1 Groundwater Contamination

a. Inspection Scope

On February 5, 2016, Entergy notified the NRC of a significant increase in groundwater tritium levels measured at three monitoring wells (MWs) (MW-30, MW-31, and MW-32) located near the Unit 2 fuel storage building. These samples were drawn on January 26 to 27, 2016, and analyzed and confirmed on February 2 to 4, 2016. The highest concentration was detected at MW-32, which increased from 12,000 pCi/l on January 11, 2016, to 8,100,000 pCi/l on January 26, 2016, and subsequently up to 14,800,000 pCi/l on February 4, 2016. This increased tritium concentration event was documented by Entergy in CR-IP2-2016-00564 which documents its investigation of this event including a root cause evaluation. The inspectors reviewed Entergy's root cause evaluation for this event during this inspection period as well as recent groundwater monitoring results.

b. Findings and Observations

No findings were identified.

Update to URI 05000247/2016001-07, January 2016 Groundwater Contamination Entergy continues to conduct weekly, biweekly, and monthly groundwater sampling of MWs at the initial site of groundwater contamination and at downstream wells towards the Hudson River. For the initial three MWs (MW-30, MW-31, and MW-32), the general trend in tritium activity has been downward, with periodic increases seen in some weekly samples. The downstream MWs located in the Unit 2 switchyard (especially MW-55) showed an initial increase in activity up to 117,000 pCi/l, but the activity at that location has plateaued at the end of the reporting period.

Entergy documented its investigation of this event as root cause evaluation for CR-IP2-2016-00564. The inspectors reviewed Entergy's root cause evaluation for this event. Entergy concluded that the source of the groundwater contamination was from the reject water of a temporary reverse osmosis unit used to process water from the refueling water storage tank at Unit 2 in preparation for RFO 2R22. Although this analysis documents a number of issues identified during the operation of the contractor reverse osmosis unit, which is believed to be the source of the groundwater contamination, one of two leakage paths to groundwater have still not been established.

The established pathway involves leakage from two cut drain lines located above the floor on the 35-foot elevation of the PAB. Further investigation by Entergy following the conclusion of the Unit 2 RFO 2R22 must be conducted to verify the second pathway to groundwater via the floor of the fuel storage building truck bay.

Entergy's long-term corrective action for reducing tritium levels in the groundwater is the same as previously identified for the March 2014 tritium spike (CR-IP2-2015-03806), the start-up and operation of recovery well (RW)-1. Following installation of equipment and system testing, full operation of the RW system is expected later this year. This system will allow for the collection of tritiated groundwater in the vicinity of Unit 2 to be returned inside the Unit 2 PAB for processing. The NRC will be conducting an inspection in

August 2016 to review the testing plan and results of the RW-1 tests. This inspection will include a specialist region-based inspector, and a staff hydrogeologist.

The NRC's continuing assessment of the safety significance of this event focused on validating the safety impact of dose to the public from the release of tritium to the site groundwater, and ultimately to the Hudson River. The NRC verified that Entergy's bounding public dose calculations on the groundwater contamination leak was sufficiently conservative and a maximum worst case scenario would result in a dose of 0.000112 millirem per year, which represents a very small fraction of the allowable dose (liquid effluent dose objective of 3 millirem per year). This low value is due to groundwater at Indian Point not being a source of any drinking water. There are no drinking water wells on the Indian Point site, groundwater flow from the site is to the Hudson River and not to any near site drinking water wells, and the Hudson River has no downstream drinking water intakes as it is brackish water. Pathways to the public are therefore limited to the consumption of fish and river invertebrates. The inspection determined that there is no safety impact to the public as a result of this groundwater

contamination event. (URI 05000247/2016001-07, January 2016 Groundwater Contamination)

.2 Institute of Nuclear Power Operations (INPO) Report Review

a. Inspection Scope

The inspectors also reviewed the final report for the INPO equipment reliability scram review visit that was conducted to review the scrams that occurred over the past two years, conducted in June 2016. The inspectors reviewed the report to ensure that any issues identified were consistent with NRC perspectives of Entergy performance and to determine if INPO identified any significant safety issues that required further NRC follow-up.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

On August 4, 2016, the inspectors presented the inspection results to Mr. Larry Coyle, Site Vice President, and other members of Entergy. Based on additional information provided, the inspectors conducted an updated exit meeting on August 30, 2016 with John Kirkpatrick, Plant Operations General Manager and other members of Entergy.

The inspectors verified that no proprietary information was retained by the inspectors or documented in this report.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Entergy Personnel

A. Vitale, Site Vice President
J. Kirkpatrick, Plant Operations General Manager
R. Alexander, Unit 2 Shift Manager
R. Andersen, Maintenance Instrumentation and Controls Superintendent
N. Azevedo, Engineering Supervisor
J. Baker, Shift Manager
S. Bianco, Operations Fire Marshal
K. Brooks, Assistant Operations Manager
R. Burroni, Engineering Director
T. Chan, Engineering Supervisor
C. Chapin, Training Superintendent
D. Dewey, Assistant Operations Manager
J. Dignam, Unit 3 Control Room Supervisor
R. Dolansky, Inservice Inspection Program Manager
W. Durr, Outage Control Center Manager
R. Drake, Engineering Supervisor
K. Elliott, Fire Protection Engineer
J. Ferrick, Regulatory and Performance Improvement Director
L. Frink, Radiation Protection Supervisor
D. Gagnon, Security Manager
L. Glander, Emergency Preparedness Manager
D. Gray, Radiological Environmental Manager
J. Johnson, Unit 2 Control Room Supervisor
M. Johnson, Unit 3 Shift Manager
M. Khadabux, Instrumentation and Controls Supervisor
F. Kich, Performance Improvement Manager
M. Lewis, Unit 3 Assistant Operations Manager
N. Lizzo, Training Manager
S. McAllister, Baffle Bolt Replacement Project Manager
M. McCarthy, Unit 3 Control Room Supervisor
B. McCarthy, Operations Manager
F. Mitchell, Radiation Protection Manager
E. Mullek, Maintenance Manager
S. Stevens, Radiation Protection Operations Superintendent
B. Sullivan, Training Superintendent
J. Taylor, Unit 3 Shift Manager
M. Tesoriero, Outage Control Center Manager
M. Troy, Nuclear Oversight Manager
R. Walpole, Regulatory Assurance Manager
A. Zastrow, Assistant Operations Manager

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened

05000247/2016002-01 URI CVCS Goal Monitoring Under the Maintenance

Rule (Section 1R12)

Opened/Closed

05000247/2016002-03 NCV Failure to Maintain Flow Channeling Gates Closed in Accordance with the Containment Procedure

(Section 1R20)

05000247/2016002-04 NCV Failure to Scope Safety-Related Main Boiler Feedwater Pump Discharge Valves into the

Maintenance Rule Program (Section 4OA3)

Closed

05000247/LER-2015-003-00 LER Manual Reactor Trip due to Indications of Multiple Dropped Control Rods Caused by Loss of Control
Rod Power Due to a Power Supply Failure

(Section 4OA3)

05000247/LER-2016-003-00 LER Technical Specification Prohibited Condition Due to an Inoperable 21 Main Boiler Feedwater
Pump Discharge Valve for Greater Than the TS
Allowed Outage Time (Section 4OA3)

Discussed

05000247/2016001-01 URI Baffle-Former Bolts with Identified Anomalies

(Section 1R08)

05000247/2016001-06 URI Emergency Diesel Generator Automatic Voltage

Regulator Failure (Section 1R15)

05000247/2016001-07 URI January 2016 Groundwater Contamination

Section (Section 4OA5)

LIST OF DOCUMENTS REVIEWED

Common Documents Used Indian Point Unit 2 and Unit 3, UFSARs Indian Point Unit 2 and Unit 3, Individual Plant Examinations Indian Point Unit 2 and Unit 3, Individual Plant Examination of External Events Indian Point Unit 2 and Unit 3, TSs and Bases Indian Point Unit 2 and Unit 3, Technical Requirements Manuals Indian Point Unit 2 and Unit 3, Control Room Narrative Logs Indian Point Unit 2 and Unit 3, Plans of the Day

Section 1R04: Equipment Alignment

Procedures

2-COL-4.2.1, Residual Heat Removal System, Revision 30
2-COL-4.3.1, Spent Fuel Pit Cooling, Revision 10
2-COL-24.1.1, Service Water System, Revision 50
3-COL-EL-005, Diesel Generators, Revision 37
OAP-019, Component Verification and System Status Control, Revision 7
OAP-044, Plant Labeling Program, Revision 3

Condition Reports

(CR-IP2)

2016-01311 2016-01505 2016-01761 2016-02330 2016-02428 2016-02470

Condition Reports

(CR-IP3)

2016-01382 2016-01810

Drawings

209762, Flow Diagram Service Water System Nuclear Steam Supply Plant, Revision 75
27781, Flow Diagram Auxiliary Coolant System, Revision 22 9321-2720, Auxiliary Coolant System, Sheet 2, Revision 22

Miscellaneous

IP3-DBD-308, CCW System, Revision 3

Section 1R05: Fire Protection

Procedures

EN-MA-133, Control of Scaffolding, Revision 12

Condition Reports

(CR-IP2)

2016-04148

Condition Reports

(CR-IP3)

2016-01272

Miscellaneous

PFP-203, Containment Building, 95-Foot Elevation (Fire Zone 86A), Revision 15
PFP-204, General Floor Plan, PAB, 15-Foot Elevation, Revision 0
PFP-209, Component Cooling Pump Room, PAB, 68-Foot Elevation, Revision 0
PFP-211, General Floor Plan, PAB, 80-Foot Elevation, Revision 14
PFP-351, 480V Switchgear Room, Revision 15

Section 1R07: Heat Sink Performance

Procedures

0-HTX-405-EDG, EDG Lube Oil and Jacket Water Heat Exchanger Maintenance, Revision 4

Condition Reports

(CR-IP3)

2010-02900 2011-03594 2011-03596 2011-03961 2012-02071 2012-03912
2013-02338 2013-02695 2013-03009 2014-00957 2014-01239 2014-03158
2014-03175 2015-00031 2015-00599 2015-02848 2015-05209 2015-05526
2016-00886 2016-00895 2016-00899
Maintenance Orders/Work Orders
WO 52489888
WO 52626563

Miscellaneous

SEP-SW-IPC-001, Indian Point Energy Center NRC Generic Letter 89-13 Service Water Program, Revision 0

Section 1R08: Inservice Inspection Activities

Procedures

GBRA-104-659, Collection of Protocols for Baffle Bolt Replacement, Revision C
GBRA-175-115, Field Service Procedure for Baffle Bolt Replacement, Revision 3
WDI-STD-088, Underwater Remote Visual Examination of Reactor Vessel Internals, Revision 13
WDI-STD-1073, Ultrasonic Test Procedure for the Inspection of Internal Hex Head Baffle-Former Bolts with Welded Lock Bars, Revision 4

Condition Reports

(CR-IP2)

2016-02081
Maintenance Orders/Work Orders
2412-13

Miscellaneous

Indian Point Unit 2 Baffle Bolt Ultrasonic Examination Expanded Analysis Report, dated April 28, 2016 IP2 Reactor Vessel Visual Examination Report, dated May 2006 Loose Parts Inventory Log for Baffle Bolt Replacements, dated May 24, 2016
MRP-227-A, Materials Reliability Program:
Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (ML120170453)
MRP-228, Materials Reliability Program:
Inspection Standard for PWR Internals - 2012 Update, Revision 1
SEP-ISI-IP2-001, IP2 Fourth Ten-Year Interval Inservice Inspection (ISI)/Containment Inservice Inspection (CISI) Program Plan, Revision 2
WDI-PJF-1315504-EPP-001, Indian Point Nuclear Power Plant
MRP-227-A Reactor Vessel Internals Examination Program Plan, Revision 0
WDI-PJF-1315505-FSR-001, Indian Point Unit 2 2R22
MRP-227-A Baffle-Former Bolt Ultrasonic Inspections Field Service Report, dated March 29, 2016
WDI-TJ-1100, Technical Justification for the Ultrasonic Inspection of Baffle-Former Bolts for Indian Point Units 2 and 3, Revision 1

Section 1R11: Licensed Operator Requalification Program

Procedures

2-AOP-480V-1, Loss of Normal Power to Any 480V Vital Bus, Revision 8
2-AOP-RCP-1, Reactor Coolant Pump Malfunction, Revision 14
2-AOP-TURB-1, Main Turbine Trip without a Reactor Trip, Revision 5
2-E-0, Reactor Trip or Safety Injection, Revision 7
2-FR-H.1, Response to Loss of Secondary Heat Sink, Revision 11
2-POP-1.2, Reactor Startup, Revision 59 2-SOP-26.4, Turbine Generator Startup, Synchronizing, Voltage Control and Shutdown, Revision 62 3-AOP-480V-1, Loss of Normal Power to Any Safeguards Bus, Revision 7
3-AOP-CVCS-1, Chemical and Volume Control System Malfunction, Revision 8
3-AOP-FW-1, Loss of Feedwater, Revision 8 3-AOP-INST-1, Instrument/Controller Failures, Revision 11
3-E-0, Reactor Trip or Safety Injection, Revision 6
3-E-1, Loss of Reactor or Secondary Coolant, Revision 4
3-FR-C.2, Response to Degraded Core Cooling, Revision 3

Condition Reports

(CR-IP2)

2016-03946 2016-04162 2016-04164 2016-04165 2016-04169 2016-04178

Condition Reports

(CR-IP3)

2016-01087 2016-01092 2016-01098 2016-01336

Miscellaneous

13SX-LOR-SES026, Licensed Operator Requalification Program Scenario Emergency Action Level Table, Revision 15.2
LRQ-SES-04, IPEC Simulator Evaluated Scenario, Revision 6

Section 1R12: Maintenance Effectiveness

Procedures

CEP-NDE-0640, Non-Section XI Liquid Penetrant Examination, Revision 9
CEP-WP-WIIR-1, Attachment 5.1, Inprocess Inspections for Installation and Replacement Welds Located Inside the ASME Section XI Boundary, Revision 3
EN-DC-206, Maintenance Rule (a)(1) Process, Revision 3

Condition Reports

(CR-IP2) 2010-00864 2013-03130 2014-00162 2014-00185 2014-01144 2014-02184

2015-00278 2016-01260 2016-01430 2016-01500

Condition Reports

(CR-IP3)

2012-03836 2013-04758 2015-01396 2015-03404 2015-03653 2015-04053
2015-04162 2015-04184 2015-04539 2015-05316 2015-05384 2015-05729
2016-00098 2016-00653 2016-00723 2016-01189 2016-01227 2016-01274
2016-01313 2016-01531 2016-01536 2016-01543 2016-02432
Maintenance Orders/Work Orders
WO 00397793
WO 00408019
WO 00414886
WO 00416091
WO 00421841
WO 00429532
WO 00429532
WO 00431497
WO 00446165
WO 00447042
WO 00447966
WO 52602429
WO 52621178

Miscellaneous

EC 65389, 21 CCW HX Leak Repair: Service Water Inlet Nozzle - Elbow Weld Configuration Change IPEC Maintenance Rule Basis Document - Chemical and Volume Control System, Revision 0
PQR 913, 134 F42
MN-GTAW ASME IX Welding Procedure Qualification Record, Revision 0 System Health Report, Unit 3, EDG, Q1-2016
Weld Map Number 447966-20-01, Revision 0
WPS 134 F42,
MN-GTAW, ASME Section IX Welding Procedure Specification, Revision 0

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Procedures

EN-OP-119, Protected Equipment, Revision 8
IP-SMM-OU-104, Attachment 9.1, Shiftly Ou tage Shutdown Safety Assessments, Revision 15
IP-SMM-OU-104, Attachment 9.2, Shiftly Ou tage Shutdown Safety Assessment Guidelines, Revision 15

Condition Reports

(CR-IP2)

2016-04141

Condition Reports

(CR-IP3)

2016-01545

Miscellaneous

EOOS Risk Assessment Software Tool

Section 1R15: Operability Determinations and Functionality Assessments

Procedures

2-PC-R3-1, Pressurizer Level Transmitters, Revision 10
3-ARP-010, Metal Impact Monitoring System, Page 10, Revision 32
3-SOP-RCS-016, Operation of the Metal Impact Monitoring System, Revision 8
EN-OP-104, Operability Determination Process, Revision 10

Condition Reports

(CR-IP2) 2016-2221 2016-2356 2016-2961 2016-3345 2016-3418 2016-3660

2016-3636 2016-3784 2016-3806 2016-3818 2016-4085

Condition Reports

(CR-IP3)

2014-01670 2015-03468
Maintenance Orders/Work Orders
WO 00327574
WO 00425980
WO 52571030

Miscellaneous

EN-LI-100, Attachment 9.1, Change Channel Check Comparison Criteria/2-PT-M100,
2-PT-D001, Revision 0

Section 1R18: Plant Modifications

Drawings

10073E87-001, Indian Point Unit 2 Baffle Bolt Replacement - Core Barrel and Baffle Assembly Elevation, Revision 0 10111D06, Indian Point Unit 2 Baffle Bolt Replacement - Replacement Baffle Bolt .625

and .750, Revision 0

Miscellaneous

EC 64308, IP2 Reactor Vessel Equivalent Replacement Baffle-to-Former Bolt, Revision 0
Process Applicability Determination Form for
EC 64308, dated April 21, 2016
WCAP-18136-P, Replacement Type 316 Cold-Worked Baffle-Former Bolt Qualification for Indian Point Unit 2, Revision 0

Section 1R19: Post-Maintenance Testing

Procedures

3-PT-M079B, 32 EDG Functional Test, Revision 52 2-PC-Q109-4, Recalibration of NIS and OT/OP Delta-T Parameters - Channel IV, Revision 44

Condition Reports

(CR-IP2)

2016-03961 2016-04266

Condition Reports

(CR-IP3)

2016-01189 2016-01199 2016-01218
Maintenance Orders/Work Orders
WO 00414886
WO 00420649
WO 00446094
WO 00447966
WO 52545181
WO 52626563
WO 52626564
WO 52630619
WO 52630620
WO 52658943
WO 00236158
WO 00277374
WO 52571030

Drawings

5651D72, Logic Diagrams Steam Generator Trip Signals, Revision 7

Miscellaneous

EC 64545, Emergency Temporary Modification to Disconnect 32 EDG Generator Space Heater Adjacent to End Plate on Outboard End of Generator FIX00091, Pressurizer Level Uncertainty - Indication, Trip Setpoints, and Annunciation Setpoints, Revision 1 E-mail from J. Michetti to G. Newman, dated July 19, 2016, Subject:
Westinghouse Report on E9

Section 1R20: Refueling and Other Outage Activities

Procedures

2-POP-1.1, Plant Heatup from Cold Shutdown, Revision 90
2-POP-1.2, Reactor Startup, Revision 59
2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 89
2-POP-3.1, Plant Shutdown from 45 Percent Power, Revision 58
2-POP-3.3, Plant Cooldown, Hot to Cold Shutdown, Revision 81
2-POP-3.4, Secondary Plant Shutdown, Revision 10 2-POP-4.1, Operation at Cold Shutdown, Revision 5 2-POP-4.2, Operation Below 20 Percent Pressurizer Level with Fuel in the Reactor, Revision 8
2-POP-4.3, Operation without Fuel in the Reactor, Revision 1

Condition Reports

(CR-IP2-)

2016-04118 2016-04119 2016-04123 2016-03124 2016-04126 2016-04129
2016-04130 2016-04131 2016-04132 2016-04139 2016-04141* 2016-04142*
2016-04144 2016-04145 2016-04146 2016-04148* 2016-04151 2016-04152
2016-04155 2016-04161 2016-04162 2016-04165 2016-04169
  • NRC identified
Maintenance Orders/Work Orders
2681465

Miscellaneous

2R22 Performance Indicator Curves Daily from March 7 to June 14, 2016
Outage Schedules and Plans of the Day from March 7 to June 14, 2016
Westinghouse
LTR-PL-16-16, Operability Assessment for Primary Side Loose Parts at Indian Point Unit 2, Revision 0, dated March 27, 2016

Section 1R22: Surveillance Testing

Procedures

0-SOP-LEAKRATE-001, RCS Leakrate Surveillance Evaluation and Leak Identification, Revision 6 2-PT-D001, Control Room Operations Surveillance Requirements, Revision 16 2-PT-M029B, 22 Safety Injection Pump, Revision 20 2-PT-Q013, Inservice Valve Tests, Revision 51
2-PT-Q013-DS040, Valve 887B Inservice Test Data Sheet, Revision 22
3-PT-M079B, 32 EDG Functional Test, Revision 52

Condition Reports

(CR-IP2)

2016-03360 2016-03363

Condition Reports

(CR-IP3)

2016-01716 2016-01752
Maintenance Orders/Work Orders
WO 00443040
WO 00446385
WO 00446867
WO 52681652-01
WO 52681646-01

Miscellaneous

EC 64855, EC Reply to Provide Input Regarding MPR Maintenance Bulletin
MB-2007-01 for Auto Voltage Regulator Solder Joints
MB-2007-01, Potential for Solder Joint Cracks on Basler SBSR Auto Voltage Regulator Cards and Technical Manual Addendum
TM-2007-01, November 5, 2007 Unit 3 RCS Routine Activity Sample, 28-June-16-10006

Section 1EP6: Drill Evaluation

Procedures

IP-EP-120, Emergency Classification, Revision 10
IP-EP-410, Protective Action Recommendations, Revision 11

Section 2RS7: Radiological Environmental Monitoring Program

Procedures

0-CY-1920, REMP Land Use Census, Revision 1
0-CY-1980, Preparation, Placement and Collection of Site Environmental Thermoluminescent Dosimeters, Revision 2

Condition Reports

(CR-IP2)

2014-05319 2015-00948 2015-01300 2015-02687 2015-02800 2015-02987
2015-03271 2015-03396 2016-02313

Condition Reports

(CR-IP3)

2016-00514

Miscellaneous

2014 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3
2015 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3 Environmental Dosimetry Company, Annual Quality Assurance Status Report, January to December 2015 Indian Point Energy Center ODCM, Revision 4
June 2015 to May 2016 Meteorological Data Recovery Met One Instruments, Inc. Certificates of Calibration for Temperature, Wind Direction, and Wind Speed Teledyne Brown Engineering Environmental Serv ices Annual 2015 Quality Assurance Report
Exelon PowerLabs Certificates of Calibration for Gas Meters
3471875
3482909
3471871
3471867
3482920
3471873
3482910
3482916
3471877
3482914
3482918
3482921
3471881
3471879
3471872
3471869
3471880
3482908
Quality Assurance Quality Assurance Audit Report
QA-2-6-2015-IP-1, Chemistry, Effluents, and Environmental Monitoring Snapshot Self-Assessment,
LO-IP3LO-2015-00126, Chemistry-REMP

Section 4OA2: Problem Identification and Resolution

Procedures

EN-DC-204, Maintenance Rule Scope and Basis, Revision 3
EN-DC-204, Maintenance Rule Scope and Basis, Revision 3
EN-DC-207, Maintenance Rule Periodic Assessment, Revision 3
EN-LI-102, Corrective Action Program, Revision 26
EN-LI-104, Self-Assessment and Benchmark Process, Revision 11
EN-LI-110-01, Equipment Failure Evaluation, Revision 0
EN-LI-119, Apparent Cause Evaluation Process, Revision 11
EN-OP-104, Operability Determination Process, Revision 10

Condition Reports

(CR-IP2)

2010-07013 2015-04574 2015-05458 2015-05460 2015-05461 2015-05464
2015-05466 2015-05467 2016-01374 2016-02348

Condition Reports

(CR-IP3) 2015-3628 2016-01035 2016-01961

Maintenance Orders/Work Orders
WO 00442412
Apparent Cause Evaluations
IP2-2015-05458

Drawings

504405, Sheet 1 of 2, Replacement Baffle Bolt Location Matrix, Revision 0
504405, Sheet 2 of 2, Replacement Baffle Bolt Location Matrix, Revision 0

Miscellaneous

61265, Temporary Modification Control Form-Install Additional Temporary Spare Power Supply Inside Rod Control Cabinet 2BD to Restore Margin, Revision 0 Appendix A to Facility Operating License
DPR-64, Technical Specifications and Basis For The Indian Point 3 Nuclear Generating Station Unit No.3, Through Amendment 260
CN-RIDA-15-43, Indian Point Units 2 and 3 Acceptable Baffle-Former Bolting LOCA and Seismic Analysis, Revision 2 Engineering Change 63938,
As-left condition of the baffle-former plate assembly following the replacement of degraded bolts, Revision 0 EPRI
TR-112209, Analysis of Baffle Former Bolt Cracking in EDF CPO Plants (PWRMRP-03), dated June 1999 Indian Point Entergy Center (IPEC) Unit 3, Updated Final Safety Analysis Report, dated May
2013
IP-RPT-16-00025, Evaluation of Indian Point Unit 3 Reactor Core Baffle Bolting Following
MRP-227-A Inspection Findings at Indian Point Unit 2 during 2R22, Revision 0
LO-IP3LO-2015-72
LTR-PL-16-21, Transmittal of Indian Point Unit 3 Final Engineering Evaluations Supporting Extent of Condition Review, Revision 0
LTR-RIDA-16-103, Indian Point Unit 2 Baffle Bolting Anti-Clustering Pattern and Margin Assessment, Revision 0
LTR-RIDA-16-152, Indian Point Unit 3 Baffle Bolt Leak Before Break Operability Assessment, Revision 0
LTR-RIDA-16-60, Indian Point Unit 2 Baffle Bolting One Cycle Replacement Pattern Summary Letter, Revision 0
MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (ML120170453)
Operation Decision Making Issue Action Plan for IP3 Baffle Bolt Monitoring, dated May 19, 2016
WCAP-15029-NP-A, Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions, Revision 0

(ML15222A882)

WCAP-17949-P, Background and Technical Basis Supporting Engineering Flaw Acceptance Criteria for Indian Point Unit 2 Reactor Vessel Internals
MRP-227-A Primary and Expansion Components, Revision 1
WCAP-18048-P, Determination of Acceptable Baffle-Former Bolting for Indian Point Units 2 and
3, Revision 0

Section 4OA5: Other Activities

Miscellaneous

INPO Letter, INPO Equipment Reliabilit

y Scram Review Visit, May 31, 2016

Root Cause Evaluation for CR-IP2-2016-00564

LIST OF ACRONYMS

CFR Title 10 of the

Code of Federal Regulations

ALARA as low as is reasonably achievable
AVR automatic voltage regulator
CAP corrective action program
CCW component cooling water
CR condition report
CVCS chemical and volume control system
EC engineering change
ECCS emergency core cooling system
EDG emergency diesel generator
GPI groundwater protection initiative
IASCC irradiation-assisted stress-corrosion cracking
IMC Inspection Manual Chapter
INPO Institute of Nuclear Power Operations
LER licensee event report
LOCA loss-of-coolant accident
MBFP main boiler feedwater pump
MCC motor control center
MOV motor operated valve
MRP materials reliability program
MW monitoring well
NCV non-cited violation
NRC Nuclear Regulatory Commission,
U.S. [[]]
ODCM offsite dose calculation manual
OOS out of service
PAB primary auxiliary building
PFP pre-fire plan
RCS reactor coolant system
REMP radiological environmental monitoring program
RFO refueling outage
RW recovery well
SI safety injection
SSC structure, system, and component
TS technical specification
UFSAR updated final safety evaluation report
URI unresolved item
UT ultrasonic testing
WO work order
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