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{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 2100 RENAISSANCE BOULEVARD, SUITE 100 KING OF PRUSSIA, PA  19406-2713  
{{#Wiki_filter:UNITED STATES  
  August 1, 2018  
NUCLEAR REGULATORY COMMISSION  
 
REGION I  
Mr. Anthony J. Vitale Site Vice President Entergy Nuclear Operations, Inc.  
2100 RENAISSANCE BOULEVARD, SUITE 100  
KING OF PRUSSIA, PA  19406-2713  
August 1, 2018  
Mr. Anthony J. Vitale  
Site Vice President  
Entergy Nuclear Operations, Inc.  
Indian Point Energy Center  
Indian Point Energy Center  
450 Broadway, General Services Building  
450 Broadway, General Services Building  
Line 26: Line 35:
Buchanan, NY 10511-0249  
Buchanan, NY 10511-0249  
   
   
SUBJECT: INDIAN POINT NUCLEAR GENERATING - INTEGRATED INSPECTION REPORT 05000247/2018002 AND 05000286/2018002  
SUBJECT:  
  Dear Mr. Vitale:  
INDIAN POINT NUCLEAR GENERATING - INTEGRATED INSPECTION  
REPORT 05000247/2018002 AND 05000286/2018002  
   
Dear Mr. Vitale:  
   
   
On June 30, 2018, the U.S. Nuclear Regulatory  
On June 30, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at  
Commission (NRC) completed an inspection at Indian Point Nuclear Generating (Indian Point), Units 2 and 3.  On July 18, 2018, the NRC inspectors discussed the results of this inspection with you and other members of your staff.   
Indian Point Nuclear Generating (Indian Point), Units 2 and 3.  On July 18, 2018, the NRC  
inspectors discussed the results of this inspection with you and other members of your staff.   
The results of this inspection are documented in the enclosed report.  
The results of this inspection are documented in the enclosed report.  
  No NRC-identified or self-revealing findings were identified during this inspection.  The NRC  
   
inspectors documented one violation, which was determined to be Severity Level IV in this report.  The NRC is treating this v
No NRC-identified or self-revealing findings were identified during this inspection.  The NRC  
iolation as a non-cited violation
inspectors documented one violation, which was determined to be Severity Level IV in this  
(NCV) consistent with Section 2.3.2.a of the Enforcement Policy.  
report.  The NRC is treating this violation as a non-cited violation (NCV) consistent with  
  If you contest the violation or significance of the NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear  
Section 2.3.2.a of the Enforcement Policy.  
   
If you contest the violation or significance of the NCV, you should provide a response within  
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear  
Regulatory Commission, ATTN:  Document Control Desk, Washington, DC 20555-0001; with  
Regulatory Commission, ATTN:  Document Control Desk, Washington, DC 20555-0001; with  
copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC  
copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC  
Resident Inspector at Indian Point.   
Resident Inspector at Indian Point.   
 
   
A. Vitale 2 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), Part 2.390, "Public Inspections, Exemptions, Requests for Withholding."
 
Sincerely, 
          /RA/  Daniel L. Schroeder, Chief Reactor Projects Branch 2


A. Vitale
2
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), Part 2.390, Public
Inspections, Exemptions, Requests for Withholding.
Sincerely,
  /RA/
Daniel L. Schroeder, Chief
Reactor Projects Branch 2
Division of Reactor Projects  
Division of Reactor Projects  
   
   
Docket Numbers:  50-247 and 50-286  
Docket Numbers:  50-247 and 50-286  
License Numbers:  DPR-26 and DPR-64  
License Numbers:  DPR-26 and DPR-64  
  Enclosure:  
   
Enclosure:  
Inspection Report 05000247/2018002 and  
Inspection Report 05000247/2018002 and  
   05000286/2018002  
   05000286/2018002  
cc w/encl:  Distribution via ListServ


cc w/encl:  Distribution via ListServ 


   ML18213A153
   ML18213A153
  SUNSI Review  
  Non-Sensitive  Sensitive  
SUNSI Review  
  Publicly Available  Non-Publicly Available  
  OFFICE RI/DRP RI/DRP RI/DRP   NAME BHaagensen via email TSetzer DSchroeder   DATE 7/25/18 7/25/18 8/1/18  
Non-Sensitive  
  Sensitive  
Publicly Available  
  Non-Publicly Available  
   
OFFICE  
RI/DRP  
RI/DRP  
RI/DRP  
NAME  
BHaagensen via email TSetzer  
DSchroeder  
DATE  
7/25/18  
7/25/18  
8/1/18  
   
   
1 Enclosure U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
 
Docket Numbers:  50-247 and 50-286
  License Numbers: DPR-26 and DPR-64


1
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
Inspection Report
Docket Numbers: 
50-247 and 50-286
License Numbers:
DPR-26 and DPR-64
Report Numbers:
05000247/2018002 and 05000286/2018002
Licensee:
Entergy Nuclear Northeast (Entergy)
Facility:
Indian Point Nuclear Generating, Units 2 and 3
Location:
450 Broadway, General Services Building
   
   
Buchanan, NY 10511-0249
   
   
Report Numbers: 05000247/2018002 and 05000286/2018002
 
Licensee: Entergy Nuclear Northeast (Entergy)
   
   
Inspection Dates:
April 1, 2018 to June 30, 2018
   
   
Facility: Indian Point Nuclear Generating, Units 2 and 3
 
Location:  450 Broadway, General Services Building
  Buchanan, NY 10511-0249
  Inspection Dates: April 1, 2018 to June 30, 2018
   
   
Inspectors:
   
   
Inspectors:  B. Haagensen, Senior Resident Inspector A. Siwy, Resident Inspector J. Vazquez, Resident Inspector J. Fuller, Acting Resident Inspector  
B. Haagensen, Senior Resident Inspector  
A. Siwy, Resident Inspector  
J. Vazquez, Resident Inspector  
J. Fuller, Acting Resident Inspector  
D. Kern, Senior Reactor Inspector  
D. Kern, Senior Reactor Inspector  
M. Modes, Senior Reactor Inspector  
M. Modes, Senior Reactor Inspector  
L. McKown, Acting Resident Inspector S. Wilson, Health Physicist  
L. McKown, Acting Resident Inspector  
S. Wilson, Health Physicist  
Approved By: 
Daniel L. Schroeder, Chief
Reactor Projects Branch 2
Division of Reactor Projects
 
2
   
   
SUMMARY
   
   
Approved By:  Daniel L. Schroeder, Chief
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring Entergys performance  
  Reactor Projects Branch 2  Division of Reactor Projects
at Indian Point Nuclear Generating, Units 2 and 3, by conducting the baseline inspections  
 
described in this report in accordance with the Reactor Oversight Process.  The Reactor  
2  SUMMARY  The U.S. Nuclear Regulatory Commission (NRC) continued monitoring Entergy's performance at Indian Point Nuclear Generating, Units 2 and 3, by conducting the baseline inspections described in this report in accordance with the Reactor Oversight Process.  The Reactor  
Oversight Process is the NRCs program for overseeing the safe operation of commercial  
Oversight Process is the NRC's program for overseeing the safe operation of commercial  
nuclear power reactors.  Refer to https://www.nrc.gov/reactors/operating/oversight.html for more  
nuclear power reactors.  Refer to https://www.nrc.gov/reactors/operating/oversight.html
information.  NRC-identified and self-revealing findings, violations, and additional items are  
for more information.  NRC-identified and self-revealing findings, violations, and additional items are summarized in the table below.
summarized in the table below.  
  List of Findings and Violations  
  Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head  
List of Findings and Violations  
   
Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head  
Penetration #3  
Penetration #3  
Cornerstone
Cornerstone  
Severity Cross-Cutting Aspect Inspection  
Severity  
Cross-Cutting Aspect  
Inspection  
Results  
Results  
Section Not Applicable Severity Level IV  
Section  
Not Applicable  
Severity Level IV  
Non-Cited Violation (NCV)  
Non-Cited Violation (NCV)  
05000286/2018-002-01  
05000286/2018-002-01  
Closed Not Applicable 71111.15 A self-revealing Severity Level IV NCV of Technical Specification (TS) 3.4.13.a, "Reactor Coolant System Operational Leakage," was identified when Entergy operated the reactor in  
Closed  
Not Applicable  
71111.15  
A self-revealing Severity Level IV NCV of Technical Specification (TS) 3.4.13.a, Reactor  
Coolant System Operational Leakage, was identified when Entergy operated the reactor in  
Mode 1 with pressure boundary leakage for a period of time longer than the allowable limiting  
Mode 1 with pressure boundary leakage for a period of time longer than the allowable limiting  
condition of operation.  Specifically, a leak in the J-weld around reactor pressure vessel  
condition of operation.  Specifically, a leak in the J-weld around reactor pressure vessel  
(RPV) head penetration #3 occurred during the last operating cycle and was not identified  
(RPV) head penetration #3 occurred during the last operating cycle and was not identified  
until after the reactor was shutdown for a refueling outage.   
until after the reactor was shutdown for a refueling outage.   
  Additional Tracking Items  
   
  Type Issue number Title Inspection  
Additional Tracking Items  
   
Type  
Issue number  
Title  
Inspection  
Results  
Results  
Section Status LER 05000286/2017-004-00
Section  
Reactor Trip Due to Main Generator Loss of Field  
Status  
  71153 Closed LER   
LER  
  05000286/2018-001-00 Reactor Trip Due to Main Generator Loss of Excitation  
05000286/2017-004-00  
71153 Closed  
Reactor Trip Due to Main  
   
Generator Loss of Field  
3   TABLE OF CONTENTS  
   
  INSPECTION SCOPES .............................................................................................................
71153  
... 4REACTOR SAFETY ................................................................................................................
Closed  
..... 4RADIATION SAFETY ..............................................................................................................
LER  
..... 8OTHER ACTIVITIES - BASELINE .............................................................................................. 9INSPECTION RESULTS ............................................................................................................ 10EXIT MEETINGS AND DEBRIEFS ............................................................................................ 15DOCUMENTS REVIEWED  
   
......................................................................................................... 16  
   
4   PLANT STATUS  
05000286/2018-001-00  
  Unit 2 began the inspection period during refueling outage 2RFO23 which lasted 33 days.  Upon completion of the outage, the operators restarted Unit 2 on April 19, 2018, and increased power slowly to approximately eight percent power when the unit was tripped by the operators due to  
Reactor Trip Due to Main  
Generator Loss of Excitation  
71153  
Closed  
 
3  
TABLE OF CONTENTS  
INSPECTION SCOPES ................................................................................................................ 4
REACTOR SAFETY ..................................................................................................................... 4
RADIATION SAFETY ................................................................................................................... 8
OTHER ACTIVITIES - BASELINE .............................................................................................. 9
INSPECTION RESULTS ............................................................................................................ 10
EXIT MEETINGS AND DEBRIEFS ............................................................................................ 15
DOCUMENTS REVIEWED ......................................................................................................... 16
 
4  
PLANT STATUS  
   
Unit 2 began the inspection period during refueling outage 2RFO23 which lasted 33 days.  Upon  
completion of the outage, the operators restarted Unit 2 on April 19, 2018, and increased power  
slowly to approximately eight percent power when the unit was tripped by the operators due to  
an issue with the main turbine control valves.  The unit was returned to rated thermal power on  
an issue with the main turbine control valves.  The unit was returned to rated thermal power on  
April 25, 2018.  On June 22, 2018, the operators reduced power to approximately 62 percent  
April 25, 2018.  On June 22, 2018, the operators reduced power to approximately 62 percent  
power to repair an oil leak on the 22 main boiler feed pump.  The unit was returned to rated  
power to repair an oil leak on the 22 main boiler feed pump.  The unit was returned to rated  
thermal power on June 23, 2018, and remained at or near rated thermal power for the remainder of the inspection period.  
thermal power on June 23, 2018, and remained at or near rated thermal power for the  
remainder of the inspection period.  
   
   
Unit 3 operated at or near rated thermal power for the entire inspection period.  
Unit 3 operated at or near rated thermal power for the entire inspection period.  
 
INSPECTION SCOPES  
INSPECTION SCOPES  
   
   
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in  
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in  
effect at the beginning of the inspection unless otherwise noted.  Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html.  Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, "Light-Water Reactor Inspection Program - Operations Phase." The inspectors performed plant status activities described in IMC 2515, Appendix D, "Plant Status" and conducted routine reviews using IP 71152, "Problem  
effect at the beginning of the inspection unless otherwise noted.  Currently approved IPs with  
Identification and Resolution." The inspectors reviewed selected procedures and records,  
their attached revision histories are located on the public website at http://www.nrc.gov/reading-
observed activities, and interviewed personnel to assess Entergy's performance and  
rm/doc-collections/insp-manual/inspection-procedure/index.html.  Samples were declared  
complete when the IP requirements most appropriate to the inspection activity were met  
consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection  
Program - Operations Phase.  The inspectors performed plant status activities described in  
IMC 2515, Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem  
Identification and Resolution.  The inspectors reviewed selected procedures and records,  
observed activities, and interviewed personnel to assess Entergys performance and  
compliance with Commission rules and regulations, license conditions, site procedures, and  
compliance with Commission rules and regulations, license conditions, site procedures, and  
standards.  
standards.  
  REACTOR SAFETY  
   
  71111.01 - Adverse Weather Protection  
REACTOR SAFETY  
   
71111.01 - Adverse Weather Protection  
   
   
Summer Readiness (1 Sample)  
Summer Readiness (1 Sample)  
  The inspectors evaluated summer readiness of offsite and alternating current power systems at Units 2 and 3 on May 17, 2018.  
   
 
The inspectors evaluated summer readiness of offsite and alternating current power  
71111.04 - Equipment Alignment  
systems at Units 2 and 3 on May 17, 2018.  
  Partial Walkdown (4 Samples)  
  The inspectors evaluated system configurations during partial walkdowns of the following  
71111.04 - Equipment Alignment  
   
Partial Walkdown (4 Samples)  
   
The inspectors evaluated system configurations during partial walkdowns of the following  
systems/trains:  
systems/trains:  
   
   
Unit 2   
Unit 2  
   
(1) Appendix R diesel generator on June 12, 2018  
(1) Appendix R diesel generator on June 12, 2018  


   
5
5  Unit 3  (2) 33 auxiliary boiler feedwater pump on June 5, 2018 (3) Service water Zurn strainers on June 18, 2018 (4) Refueling water storage tank on June 28, 2018  
 
Unit 3  
   
(2) 33 auxiliary boiler feedwater pump on June 5, 2018  
(3) Service water Zurn strainers on June 18, 2018  
(4) Refueling water storage tank on June 28, 2018  
   
   
Complete Walkdown (1 Sample)  
Complete Walkdown (1 Sample)  
   
   
The inspectors evaluated system configurations during a complete walkdown of the Unit 2 emergency diesel generators on June 27, 2018.  
The inspectors evaluated system configurations during a complete walkdown of the Unit 2  
  71111.05AQ - Fire Protection Annual/Quarterly  
emergency diesel generators on June 27, 2018.  
  Quarterly Inspection (5 Samples)  
   
71111.05AQ - Fire Protection Annual/Quarterly  
   
Quarterly Inspection (5 Samples)  
   
   
The inspectors evaluated fire protection program implementation in the following selected  
The inspectors evaluated fire protection program implementation in the following selected  
areas:  
areas:  
   
   
Unit 2   
Unit 2  
   
(1) Primary auxiliary building and fan house pre-fire plan (PFP-215) on May 15, 2018  
(1) Primary auxiliary building and fan house pre-fire plan (PFP-215) on May 15, 2018  
(2) Cable spreading room (PFP-252) on May 31, 2018  
(2) Cable spreading room (PFP-252) on May 31, 2018  
(3) Primary auxiliary building, 15-foot elevation (PFP-204), on June 7, 2018  
(3) Primary auxiliary building, 15-foot elevation (PFP-204), on June 7, 2018  
   
   
Unit 3   
Unit 3  
   
(4) Fuel storage building (PFP-315 and PFP-316) on June 6, 2018  
(4) Fuel storage building (PFP-315 and PFP-316) on June 6, 2018  
(5) Diesel generator building (PFP-354) on June 29, 2018  
(5) Diesel generator building (PFP-354) on June 29, 2018  
 
71111.07A - Heat Sink Performance  
71111.07A - Heat Sink Performance  
   
   
Heat Sink (1 Sample)  
Heat Sink (1 Sample)  
   
   
The inspectors observed Entergy's inspections and the state of cleanliness of the tubes of  
The inspectors observed Entergys inspections and the state of cleanliness of the tubes of  
the 22 component cooling water heat exchanger at Unit 2 on May 22, 2018.  
the 22 component cooling water heat exchanger at Unit 2 on May 22, 2018.  
 
71111.08P - Inservice Inspection Activities (1 Sample)  
71111.08P - Inservice Inspection Activities (1 Sample)  
  The inspectors evaluated Entergy's non-destructive examination and welding activities at  
   
The inspectors evaluated Entergys non-destructive examination and welding activities at  
Unit 2 by reviewing the following examinations from April 24 to April 27, 2018:  
Unit 2 by reviewing the following examinations from April 24 to April 27, 2018:  
(1) Volumetric Examinations
The ultrasonic, eddy current, and liquid penetrant examination of the J-weld and
associated interfaces of RPV head penetration #3, with subsequent repair by weld
surface seal weld
Ultrasonic examination of the dissimilar welds of reactor nozzles 24-1A and 22-1A


  (1) Volumetric Examinations  The ultrasonic, eddy current, and liquid penetrant examination of the J-weld and associated interfaces of RPV head penetration #3, with subsequent repair by weld surface seal weld Ultrasonic examination of the dissimilar welds of reactor nozzles 24-1A and 22-1A
6
   
   
6  (2) The inspectors reviewed Entergy's boric acid corrosion control program performance. (3) The inspectors reviewed Entergy's steam generator aging management program, including eddy current testing of steam generator tubes. 71111.11 - Licensed Operator Requalification Program and Licensed Operator Performance  
   
Operator Requalification (1 Sample)
(2) The inspectors reviewed Entergys boric acid corrosion control program performance.  
(3) The inspectors reviewed Entergys steam generator aging management program,  
including eddy current testing of steam generator tubes.  
71111.11 - Licensed Operator Requalification Program and Licensed Operator Performance  
   
   
Unit 3  (1) The inspectors observed and evaluated a crew of licensed operators in the plant's simulator during licensed operator requalification training on June 5, 2018.  
Operator Requalification (1 Sample)
Unit 3  
   
(1) The inspectors observed and evaluated a crew of licensed operators in the plants
simulator during licensed operator requalification training on June 5, 2018.  
   
   
Operator Performance (1 Sample)  
Operator Performance (1 Sample)  
  Unit 2  (1) The inspectors observed and evaluated operator performance during 2RFO23 reactor startup on April 19, 2018.  
   
  71111.12 - Maintenance Effectiveness  
Unit 2  
  Routine Maintenance Effectiveness  
   
(1 Sample)  
(1) The inspectors observed and evaluated operator performance during 2RFO23 reactor  
  The inspectors evaluated the effectiveness of routine maintenance activities associated  
startup on April 19, 2018.  
   
71111.12 - Maintenance Effectiveness  
   
Routine Maintenance Effectiveness (1 Sample)  
   
The inspectors evaluated the effectiveness of routine maintenance activities associated  
with the following equipment and/or safety significant functions:  
with the following equipment and/or safety significant functions:  
 
Unit 2  (1) CR-IP2-2018-2246, Containment structure on June 25, 2018  
Unit 2  
 
   
71111.13 - Maintenance Risk Assessments and Emergent Work Control (8 Samples)  
(1) CR-IP2-2018-2246, Containment structure on June 25, 2018  
  The inspectors evaluated the risk assessments for the following planned and emergent  
 
71111.13 - Maintenance Risk Assessments and Emergent Work Control (8 Samples)  
   
The inspectors evaluated the risk assessments for the following planned and emergent  
work activities:  
work activities:  
Unit 2
(1) Planned Yellow risk associated with freeze seal to affect repairs to MOV-769,
component cooling water to reactor coolant pump seal cooling, isolation valve on April 7,
2018
(2) Planned Yellow risk for vacuum fill of the reactor coolant system (RCS) on April 14, 2018
(3) Risk for maintenance on containment recirculation sump level transmitter LT-939 on
April 20, 2018
(4) Unplanned Yellow risk for loss of bus 2 on 13.8kV feeder 13W93 on June 1, 2018
Unit 3
(5) 13.8kV feeders out of service for planned gas turbine auto transformer maintenance on
May 1, 2018
(6) Planned Yellow risk for 480 volt degraded voltage relay testing on May 17, 2018


7
(7) Planned Yellow risk for charging pump backflow preventer maintenance on June 6, 2018
(8) Planned Yellow risk for 32 component cooling water pump oil sampling on June 13,
2018
   
   
Unit 2  (1) Planned Yellow risk associated with freeze seal to affect repairs to MOV-769, component cooling water to reactor coolant pump seal cooling, isolation valve on April 7,
71111.15 - Operability Determinations and Functionality Assessments (5 Samples)  
2018 (2) Planned Yellow risk for vacuum fill of the reactor coolant system (RCS) on April 14, 2018
(3) Risk for maintenance on containment recirculation sump level transmitter LT-939 on April 20, 2018 (4) Unplanned Yellow risk for loss of bus 2 on 13.8kV feeder 13W93 on June 1, 2018
 
   
   
Unit 3  (5) 13.8kV feeders out of service for planned gas turbine auto transformer maintenance on May 1, 2018 (6) Planned Yellow risk for 480 volt degraded voltage relay testing on May 17, 2018 
The inspectors evaluated the following operability determinations and functionality  
7  (7) Planned Yellow risk for charging pump backflow preventer maintenance on June 6, 2018 (8) Planned Yellow risk for 32 component cooling water pump oil sampling on June 13, 2018  71111.15 - Operability Determinations and Functionality Assessments (5 Samples)
The inspectors evaluated the following operability determinations and functionality  
assessments:  
assessments:  
Unit 2  (1) CR-IP2-2018-02127, Baffle bolt operability assessment during 2RFO23 on April 17, 2018 (2) CR-IP2-2018-1944, RPV head penetration #3 through-wall leak past operability on April
17, 2018 (3) CR-IP2-2018-2359, PCV-1191 containment pressure relief valve on April 19, 2018
   
   
Unit 3  (4) CR-IP3-2018-01448, 34 atmospheric dump valve past-operability associated with unsecured nearby equipment on May 23, 2018 (5) CR-IP3-2018-01552, 33 service water pump operability during 32 emergency diesel generator testing with 35 service water pump out of service on June 5, 2018  
Unit 2
(1) CR-IP2-2018-02127, Baffle bolt operability assessment during 2RFO23 on April 17,
2018
(2) CR-IP2-2018-1944, RPV head penetration #3 through-wall leak past operability on April
17, 2018
(3) CR-IP2-2018-2359, PCV-1191 containment pressure relief valve on April 19, 2018
Unit 3  
   
(4) CR-IP3-2018-01448, 34 atmospheric dump valve past-operability associated with  
unsecured nearby equipment on May 23, 2018  
(5) CR-IP3-2018-01552, 33 service water pump operability during 32 emergency diesel  
generator testing with 35 service water pump out of service on June 5, 2018  
   
   
71111.18 - Plant Modifications (2 Samples)  
71111.18 - Plant Modifications (2 Samples)  
The inspectors evaluated the following temporary or permanent modifications:
   
   
(1) Engineering Change (EC) Package 76614 - Evaluate using freeze seal isolation on 6-inch line 13 to perform maintenance on valve 769 (temporary modification) at Unit 2 on April 7, 2018 (2) EC Package 75312 - Engineering weld-repaired leaking RHR line A-98 per ASME Code Case N-666 (permanent modification) on at Unit 2 on April 24, 2018  
The inspectors evaluated the following temporary or permanent modifications:
(1) Engineering Change (EC) Package 76614 - Evaluate using freeze seal isolation on  
6-inch line 13 to perform maintenance on valve 769 (temporary modification) at Unit 2 on  
April 7, 2018  
(2) EC Package 75312 - Engineering weld-repaired leaking RHR line A-98 per ASME Code  
Case N-666 (permanent modification) on at Unit 2 on April 24, 2018  
   
   
71111.19 - Post-Maintenance Testing (13 Samples)  
71111.19 - Post-Maintenance Testing (13 Samples)  
  The inspectors evaluated post-maintenance testing for the following maintenance/repair  
   
The inspectors evaluated post-maintenance testing for the following maintenance/repair  
activities:  
activities:  
Unit 2
(1)
MOV-744, Anchor Darling double-disc gate valve inspection following repairs on
April 2, 2018
(2)
MOV-769, Anchor Darling double-disc gate valve inspection following repairs on
April 2, 2018
(3)
MOV-746, Anchor Darling double-disc gate valve inspection following repairs on
April 2, 2018
(4)
MOV-784, Anchor Darling double-disc gate valve inspection of condition and
performance monitoring following minor valve maintenance on April 7, 2018


8
   
   
Unit 2  (1) MOV-744, Anchor Darling double-disc gate valve inspection following repairs on April 2, 2018 (2) MOV-769, Anchor Darling double-disc gate valve inspection following repairs on April 2, 2018 (3) MOV-746, Anchor Darling double-disc gate valve inspection following repairs on April 2, 2018 (4) MOV-784, Anchor Darling double-disc gate valve inspection of condition and performance monitoring following minor valve maintenance on April 7, 2018
(5)  
 
MOV-885B, Anchor Darling double-disc gate valve inspection of condition and  
8  (5) MOV-885B, Anchor Darling double-disc gate valve inspection of condition and performance monitoring following minor valve maintenance on April 7, 2018 (6) Containment pressure relief valve PCV-1191 following repairs for local leak-rate test failure on April 12, 2018 (7) RPV head penetration #3 weld repairs on April 17, 2018  
performance monitoring following minor valve maintenance on April 7, 2018  
(8) 24 reactor coolant pump motor repair on April 18, 2018  
(6)  
(9) Control rod H6 following individual rod position indication repairs on April 18, 2018.  
Containment pressure relief valve PCV-1191 following repairs for local leak-rate test  
failure on April 12, 2018  
(7)  
RPV head penetration #3 weld repairs on April 17, 2018  
(8)  
24 reactor coolant pump motor repair on April 18, 2018  
(9)  
Control rod H6 following individual rod position indication repairs on April 18, 2018.  
(10) 22 reactor coolant pump motor repair on April 18, 2018  
(10) 22 reactor coolant pump motor repair on April 18, 2018  
(11) Control rod G3 drive coil stack replacement on April 18, 2018  
(11) Control rod G3 drive coil stack replacement on April 18, 2018  
  Unit 3  (12) Control rod M4 individual rod position indication post-maintenance testing following calibration to resolve indication issues on May 8, 2018 (13) 33 atmospheric dump valve post-maintenance testing following instrument airline  
   
Unit 3  
   
(12) Control rod M4 individual rod position indication post-maintenance testing following  
calibration to resolve indication issues on May 8, 2018  
(13) 33 atmospheric dump valve post-maintenance testing following instrument airline  
repairs on May 23, 2018  
repairs on May 23, 2018  
   
   
71111.20 - Refueling and Other Outage Activities (1 Sample)  
71111.20 - Refueling and Other Outage Activities (1 Sample)  
  The inspectors evaluated Unit 2 2RFO23 activities from March 18 to April 21, 2018.  
   
The inspectors evaluated Unit 2 2RFO23 activities from March 18 to April 21, 2018.  
   
   
71111.22 - Surveillance Testing  
71111.22 - Surveillance Testing  
  The inspectors evaluated the following surveillance tests:  
   
 
The inspectors evaluated the following surveillance tests:  
   
   
Routine (3 Samples)  
Routine (3 Samples)  
  (1) 3-PT-M13B1, Reactor Protection Train B Logic Test, at Unit 3 on May 7, 2018  
   
(1) 3-PT-M13B1, Reactor Protection Train B Logic Test, at Unit 3 on May 7, 2018  
(2) 3-PT-Q120A, 31 Motor Driven Auxiliary Boiler Feedwater Pump Functional Test, at  
(2) 3-PT-Q120A, 31 Motor Driven Auxiliary Boiler Feedwater Pump Functional Test, at  
Unit 3 on May 21, 2018 (3) 0-PT-6Y001, B.5.b Diesel Driven Pump Capacity Test, at Units 2 and 3 on June 13, 2018  RADIATION SAFETY  
Unit 3 on May 21, 2018  
  71124.07 - Radiological Hazard Assessment and Exposure Controls  
(3) 0-PT-6Y001, B.5.b Diesel Driven Pump Capacity Test, at Units 2 and 3 on June 13,  
  Site Inspection (1 Sample)  
2018  
  The inspectors verified that the radiological environmental monitoring program quantified the  
   
impact of radioactive effluent releases to
RADIATION SAFETY  
the environment and sufficiently validated the integrity of the radioactive gaseous and liquid effluent release program.  The inspectors  
   
71124.07 - Radiological Hazard Assessment and Exposure Controls  
   
Site Inspection (1 Sample)  
   
The inspectors verified that the radiological environmental monitoring program quantified the  
impact of radioactive effluent releases to the environment and sufficiently validated the  
integrity of the radioactive gaseous and liquid effluent release program.  The inspectors  
observed environmental sampling of groundwater, surface water, broad leaf vegetation, and  
observed environmental sampling of groundwater, surface water, broad leaf vegetation, and  
air samples.  
air samples.  
  The inspectors verified that the radiological environmental monitoring program was  
   
consistently implemented with Entergy's TSs and/or offsite dose calculation manual and validated that the program meets the design objectives in Appendix I to Title 10 of the  
The inspectors verified that the radiological environmental monitoring program was  
Code of Federal Regulations (10 CFR), Part 2.39010.   
consistently implemented with Entergys TSs and/or offsite dose calculation manual and  
 
validated that the program meets the design objectives in Appendix I to Title 10 of the Code
9  Groundwater Protection Initiative Implementation (1 Sample)
of Federal Regulations (10 CFR), Part 2.39010.   
  The inspectors verified that Entergy is continuing to implement the voluntary Nuclear Energy Institute/Industry Groundwater Protection Initiative. 
   
OTHER ACTIVITIES - BASELINE
   
71151 - Performance Indicator Verification
  The inspectors verified Entergy's performance indicators submittals listed below for the period from April 1, 2017, through March 31, 2018 (6 Samples).


9
   
   
Unit 2  (1) Unplanned Scrams Per 7000 Critical Hours (IE01) (2) Unplanned Power Changes Per 7000 Critical Hours (IE03)   
Groundwater Protection Initiative Implementation (1 Sample)
The inspectors verified that Entergy is continuing to implement the voluntary Nuclear Energy
Institute/Industry Groundwater Protection Initiative. 
OTHER ACTIVITIES - BASELINE
71151 - Performance Indicator Verification
The inspectors verified Entergys performance indicators submittals listed below for the
period from April 1, 2017, through March 31, 2018 (6 Samples).
Unit 2  
   
(1) Unplanned Scrams Per 7000 Critical Hours (IE01)  
(2) Unplanned Power Changes Per 7000 Critical Hours (IE03)   
(3) Unplanned Scrams with Complications (IE04)  
(3) Unplanned Scrams with Complications (IE04)  
   
   
Unit 3   
Unit 3  
   
(4) Unplanned Scrams Per 7000 Critical Hours (IE01)  
(4) Unplanned Scrams Per 7000 Critical Hours (IE01)  
(5) Unplanned Power Changes Per 7000 Critical Hours (IE03)  
(5) Unplanned Power Changes Per 7000 Critical Hours (IE03)  
Line 300: Line 576:
Semiannual Trend Review (1 Sample)  
Semiannual Trend Review (1 Sample)  
   
   
The inspectors reviewed Entergy's corrective action program for trends that might be  
The inspectors reviewed Entergys corrective action program for trends that might be  
indicative of a more significant safety issue.  
indicative of a more significant safety issue.  
Annual Follow-up of Selected Issues (1 Sample)
   
   
The inspectors reviewed Entergy's implementation of its corrective action program related to  
Annual Follow-up of Selected Issues (1 Sample)
The inspectors reviewed Entergys implementation of its corrective action program related to  
the following issue:  
the following issue:  
 
(1) Several fire protection deficiencies were identified for Units 2 and 3 during the last year, including portable fire extinguisher preventive maintenance not performed documented  
(1) Several fire protection deficiencies were identified for Units 2 and 3 during the last year,  
including portable fire extinguisher preventive maintenance not performed documented  
in the Indian Point Integrated Inspection Report 05000247/2017007 and  
in the Indian Point Integrated Inspection Report 05000247/2017007 and  
05000286/2017007 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML17341A080), fire hose hydrostatic testing and preventive maintenance not performed, and pre-fire plan inaccuracy.  Condition reports associated with the above issues included CR-IP2-2017-02945, CR-IP2-2017-04090,  
05000286/2017007 (Agencywide Documents Access and Management System  
(ADAMS) Accession Number ML17341A080), fire hose hydrostatic testing and  
preventive maintenance not performed, and pre-fire plan inaccuracy.  Condition reports  
associated with the above issues included CR-IP2-2017-02945, CR-IP2-2017-04090,  
CR-IP2-2018-02800, CR-IP3-2017-04084, CR-IP3-2017-05035, CR-IP3-2017-05040,  
CR-IP2-2018-02800, CR-IP3-2017-04084, CR-IP3-2017-05035, CR-IP3-2017-05040,  
CR-IP3-2018-00463, and CR-IP3-2018-00533.  
CR-IP3-2018-00463, and CR-IP3-2018-00533.  
   
   
 
10   71153 - Follow-up of Events and Notices of Enforcement Discretion  
  Events (1 Sample)  
 
10  
71153 - Follow-up of Events and Notices of Enforcement Discretion  
   
Events (1 Sample)  
   
   
The inspectors evaluated response to the following event:  
The inspectors evaluated response to the following event:  
   
   
(1) RPV head penetration #3 flaw, 50.72 report, at Unit 2 on April 2, 2018  
(1) RPV head penetration #3 flaw, 50.72 report, at Unit 2 on April 2, 2018  
   
   
Licensee Event Reports (LERs) (2 Samples)  
Licensee Event Reports (LERs) (2 Samples)  
   
   
The inspectors evaluated the following LERs:  
The inspectors evaluated the following LERs:  
   
   
(1) LER 05000286/2017-004-00, Unit 3 Reactor Trip Due to Main Generator Loss of Field (ADAMS Accession Number ML17363A212)  
(1) LER 05000286/2017-004-00, Unit 3 Reactor Trip Due to Main Generator Loss of Field  
  Entergy conducted a causal evaluation and determined that the Thyristor Firing  
(ADAMS Accession Number ML17363A212)  
   
Entergy conducted a causal evaluation and determined that the Thyristor Firing  
Module B drawer had failed.  Entergy made repairs to the failed components and  
Module B drawer had failed.  Entergy made repairs to the failed components and  
 
returned the system to service with additional monitoring equipment.  The inspectors did  
returned the system to service with additional monitoring equipment.  The inspectors did not identify any new issues during the review of the LER.  The inspectors determined that it was not reasonable to foresee and prevent the cause discussed in the LER.  Therefore, no performance deficiency was identified.  The inspectors also concluded that  
not identify any new issues during the review of the LER.  The inspectors determined  
that it was not reasonable to foresee and prevent the cause discussed in the LER.   
Therefore, no performance deficiency was identified.  The inspectors also concluded that  
no violation of NRC requirements occurred.   
no violation of NRC requirements occurred.   
  (2) LER 05000286/2018-001-00, Unit 3 Reactor Trip Due to Main Generator Loss of  
   
(2) LER 05000286/2018-001-00, Unit 3 Reactor Trip Due to Main Generator Loss of  
Excitation (ADAMS Accession Number ML18110A088)  
Excitation (ADAMS Accession Number ML18110A088)  
  Entergy conducted a causal evaluation and determined that the Thyristor Firing  
   
 
Entergy conducted a causal evaluation and determined that the Thyristor Firing  
Module A drawer had failed.  Entergy replaced both Thyristor Firing Module drawers and  
Module A drawer had failed.  Entergy replaced both Thyristor Firing Module drawers and  
 
returned the system to service with additional monitoring equipment.  The inspectors did  
returned the system to service with additional monitoring equipment.  The inspectors did not identify any new issues during the review of the LER.  The inspectors determined that it was not reasonable to foresee and prevent the cause discussed in the LER.  Therefore, no performance deficiency was identified.  The inspectors also concluded that  
not identify any new issues during the review of the LER.  The inspectors determined  
that it was not reasonable to foresee and prevent the cause discussed in the LER.   
Therefore, no performance deficiency was identified.  The inspectors also concluded that  
no violation of NRC requirements occurred.   
no violation of NRC requirements occurred.   
   
   
Personnel Performance (1 Sample)  
Personnel Performance (1 Sample)  
The inspectors evaluated response during the following non-routine evolutions or transients.
   
   
(1) Unit 2 reactor trip at 7 percent power due to a loss of feedwater during main turbine startup from 2RFO23 on April 20, 2018.   
The inspectors evaluated response during the following non-routine evolutions or transients.
  INSPECTION RESULTS  
  Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head  
(1) Unit 2 reactor trip at 7 percent power due to a loss of feedwater during main turbine  
startup from 2RFO23 on April 20, 2018.   
   
INSPECTION RESULTS  
   
Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head  
Penetration #3  
Penetration #3  
Cornerstone
Cornerstone  
Severity Cross-Cutting Aspect Report  
Severity  
Section Not Applicable Severity Level IV NCV 05000286/2018-002-01  
Cross-Cutting Aspect  
Report  
Section  
Not Applicable  
Severity Level IV  
NCV 05000286/2018-002-01  
Closed
Not Applicable
71111.15


Closed Not Applicable 71111.15  
11
11  Introduction:  A self-revealing Severity Level IV NCV of TS 3.4.13.a, "Reactor Coolant System Operational Leakage," was identified when Entergy operated the reactor in Mode 1 with pressure boundary leakage for a period of time longer than the allowable limiting condition of  
   
Introduction:  A self-revealing Severity Level IV NCV of TS 3.4.13.a, Reactor Coolant System  
Operational Leakage, was identified when Entergy operated the reactor in Mode 1 with  
pressure boundary leakage for a period of time longer than the allowable limiting condition of  
operation.  Specifically, a leak in the J-weld around RPV head penetration #3 occurred during  
operation.  Specifically, a leak in the J-weld around RPV head penetration #3 occurred during  
the last operating cycle and was not identified until after the reactor was shutdown for a refueling outage.  Description:  On March 28, 2018, during the Unit 2 bare metal exam, Entergy observed white deposits on the top of the RPV head next to (spare) penetration #3.  Entergy concluded the deposits were dried boric acid.  Subsequently, Entergy evaluated indications using ultrasonic test, eddy current test, and dye penetrant test examination techniques.  Entergy concluded that one indication had propagated into a through-wall pressure boundary leak in RPV head  
the last operating cycle and was not identified until after the reactor was shutdown for a  
refueling outage.   
Description:  On March 28, 2018, during the Unit 2 bare metal exam, Entergy observed white  
deposits on the top of the RPV head next to (spare) penetration #3.  Entergy concluded the  
deposits were dried boric acid.  Subsequently, Entergy evaluated indications using ultrasonic  
test, eddy current test, and dye penetrant test examination techniques.  Entergy concluded  
that one indication had propagated into a through-wall pressure boundary leak in RPV head  
penetration #3 caused by primary water stress cracking corrosion.  Entergy determined that  
penetration #3 caused by primary water stress cracking corrosion.  Entergy determined that  
this leakage was prohibited by TS 3.4.13, "Reactor Coolant System Operational Leakage," which requires that there be no pressure boundary leakage during operating Modes 1 through  
this leakage was prohibited by TS 3.4.13, Reactor Coolant System Operational Leakage,  
4.  The condition was reported to the NRC in
which requires that there be no pressure boundary leakage during operating Modes 1 through  
Event Notification EN-53305 on March 31, 2018, as required by 10 CFR 50.72(b)(3)(ii)(A), because it constituted a degradation of a principal  
4.  The condition was reported to the NRC in Event Notification EN-53305 on March 31, 2018,  
as required by 10 CFR 50.72(b)(3)(ii)(A), because it constituted a degradation of a principal  
safety barrier.  Entergy also reported this event in LER 05000247/2018-001-00 (ADAMS  
safety barrier.  Entergy also reported this event in LER 05000247/2018-001-00 (ADAMS  
Accession Number ML18149A126) on May 21, 2018, under 10 CFR 50.73(a)(2)(ii)(A) and  
Accession Number ML18149A126) on May 21, 2018, under 10 CFR 50.73(a)(2)(ii)(A) and  
10 CFR 50.73(a)(2)(i)(B).  These subparts describe an event or condition that resulted in the nuclear power plant including its principal safety barriers being seriously degraded, and an operation or a condition which was prohibited by the plant's TSs, respectively.   
10 CFR 50.73(a)(2)(i)(B).  These subparts describe an event or condition that resulted in the  
 
nuclear power plant including its principal safety barriers being seriously degraded, and an  
operation or a condition which was prohibited by the plants TSs, respectively.   
   
   
The inspectors reviewed the RCS leakage monitoring records, Entergy's event report, the  
The inspectors reviewed the RCS leakage monitoring records, Entergys event report, the  
causal analyses, and previous examination records.  These reviews were conducted to  
causal analyses, and previous examination records.  These reviews were conducted to  
determine whether Entergy's monitoring of operational leakage during the previous operating cycle and prior non-destructive examination activities met regulatory requirements and applicable standards.   
determine whether Entergys monitoring of operational leakage during the previous operating  
 
cycle and prior non-destructive examination activities met regulatory requirements and  
applicable standards.   
   
   
The inspectors independently reviewed the unidentified leak rate data over the previous  
The inspectors independently reviewed the unidentified leak rate data over the previous  
operating cycle and the trace boric acid deposits observed during the refueling outage.  The inspectors noted that Entergy completed a bare metal examination of the RPV the previous refueling outage and determined there were no boric acid deposits in this location.  Entergy  
operating cycle and the trace boric acid deposits observed during the refueling outage.  The  
inspectors noted that Entergy completed a bare metal examination of the RPV the previous  
refueling outage and determined there were no boric acid deposits in this location.  Entergy  
conducted isotopic analysis of the boric acid deposits and estimated the leak occurred within  
conducted isotopic analysis of the boric acid deposits and estimated the leak occurred within  
the previous 90 to 180 days.  The boron deposits were pure white in appearance indicating  
the previous 90 to 180 days.  The boron deposits were pure white in appearance indicating  
that there were no corrosion products present.  Entergy completed leakage monitoring during  
that there were no corrosion products present.  Entergy completed leakage monitoring during  
the operating cycle in accordance with standards and TS requirements.  The inspectors determined that the applicable limits were met and that the leak was not readily detectable by these methods because the change in leak rate was too small to be identified.  The  
the operating cycle in accordance with standards and TS requirements.  The inspectors  
determined that the applicable limits were met and that the leak was not readily detectable by  
these methods because the change in leak rate was too small to be identified.  The  
inspectors ultimately concluded there was no opportunity for Entergy to observe this condition  
inspectors ultimately concluded there was no opportunity for Entergy to observe this condition  
visually during the operating cycle because existing boron deposits were not visible when the  
visually during the operating cycle because existing boron deposits were not visible when the  
plant was operating.  
plant was operating.  
  The inspectors observed activities during the refueling outage to determine whether Entergy  
   
The inspectors observed activities during the refueling outage to determine whether Entergy  
completed examinations of the RPV and head penetration #3 in accordance with regulatory  
completed examinations of the RPV and head penetration #3 in accordance with regulatory  
requirements in 10 CFR Part 50, paragraph (g)(6)(ii)(D)), which conditionally endorses  
requirements in 10 CFR Part 50, paragraph (g)(6)(ii)(D)), which conditionally endorses  
American Society of Mechanical Engineers (ASME), Code Case N-729-4.  Specifically, the inspectors reviewed the results of the nondestructive examinations:  time-of-flight ultrasonic testing, liquid penetrant testing, and eddy current testing.  The inspectors also conducted discussions with the personnel who performed these tests.  The inspectors determined  
American Society of Mechanical Engineers (ASME), Code Case N-729-4.  Specifically, the  
inspectors reviewed the results of the nondestructive examinations:  time-of-flight ultrasonic  
testing, liquid penetrant testing, and eddy current testing.  The inspectors also conducted  
discussions with the personnel who performed these tests.  The inspectors determined  
Entergy staff and their contractors implemented these non-destructive examination activities  
Entergy staff and their contractors implemented these non-destructive examination activities  
in accordance with regulatory requirements and their implementing procedures.
in accordance with regulatory requirements and their implementing procedures.
 
 
12   Corrective Actions:  Entergy implemented corrective actions to repair head penetration #3 by applying a weld onlay prior to returning the RPV to service.  This alternative repair method  
12  
Corrective Actions:  Entergy implemented corrective actions to repair head penetration #3 by  
applying a weld onlay prior to returning the RPV to service.  This alternative repair method  
was the subject of relief request IP2-ISI-RR-06 which was submitted to the NRC and  
was the subject of relief request IP2-ISI-RR-06 which was submitted to the NRC and  
 
subsequently approved (ADAMS Accession Number ML18101A032).  This weld onlay  
subsequently approved (ADAMS Accession Nu
reestablished the RCS pressure boundary and addressed the primary water stress cracking  
mber ML18101A032).  This weld onlay reestablished the RCS pressure boundary and addressed the primary water stress cracking corrosion degradation mechanism.  Entergy verified that the pressure boundary had been restored using both non-destructive testing methods (dye penetration testing and ultrasonic  
corrosion degradation mechanism.  Entergy verified that the pressure boundary had been  
restored using both non-destructive testing methods (dye penetration testing and ultrasonic  
testing) and an operational pressure test.   
testing) and an operational pressure test.   
   
   
Corrective Action References:  CR-IP2-2018-01944 and CR-IP2-2018-02122  
Corrective Action References:  CR-IP2-2018-01944 and CR-IP2-2018-02122  
 
Performance Assessment:  The inspectors determined that the pressure boundary leakage was not within Entergy's
Performance Assessment:   
The inspectors determined that the pressure boundary leakage was not within Entergys
ability to foresee and prevent.  Entergy had complied with all ASME code requirements for  
ability to foresee and prevent.  Entergy had complied with all ASME code requirements for  
inspection of the RPV and there were no indications of RCS pressure boundary leakage  
inspection of the RPV and there were no indications of RCS pressure boundary leakage  
during the operating cycle.  As a result, no performance deficiency was identified.  Therefore, this violation will not be considered in the assessment process or the NRC's action matrix.  
during the operating cycle.  As a result, no performance deficiency was identified.  Therefore,  
  Enforcement:  
this violation will not be considered in the assessment process or the NRCs action matrix.  
   
Enforcement:  
   
   
Severity:  This issue is considered within the Traditional Enforcement process because there was no performance deficiency associated with the violation of NRC requirements, and the Reactor Oversight Process's significant determination process does not specifically consider  
Severity:  This issue is considered within the Traditional Enforcement process because there  
violations without performance deficiencies in its assessment of licensee performance.  In reaching this decision, the NRC determined that the issue was not within Entergy's ability to  
was no performance deficiency associated with the violation of NRC requirements, and the  
foresee and correct, Entergy's actions did not contribute to the degraded condition, and the  
Reactor Oversight Processs significant determination process does not specifically consider  
actions taken were reasonable to identify and address the condition.  Furthermore, because Entergy's actions did not contribute to this violation, it will not be considered in the  
violations without performance deficiencies in its assessment of licensee performance.  In  
assessment process or the NRC's action matrix.   
reaching this decision, the NRC determined that the issue was not within Entergys ability to  
 
foresee and correct, Entergys actions did not contribute to the degraded condition, and the  
actions taken were reasonable to identify and address the condition.  Furthermore, because  
Entergys actions did not contribute to this violation, it will not be considered in the  
assessment process or the NRCs action matrix.   
   
   
Using IMC 0612, Appendix B, "Issue Screening," the inspectors determined that the violation  
Using IMC 0612, Appendix B, Issue Screening, the inspectors determined that the violation  
is associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power  
is associated with the Equipment Performance attribute of the Initiating Events cornerstone  
and adversely affected the cornerstone objective to limit the likelihood of events that upset  
plant stability and challenge critical safety functions during shutdown as well as power  
operations.  The Barrier Integrity cornerstone was not applicable for this violation because the  
operations.  The Barrier Integrity cornerstone was not applicable for this violation because the  
violation did not involve pressurized thermal shock.     
violation did not involve pressurized thermal shock.     
   
   
The inspectors evaluated the issue using IMC 0609.04, "Initial Characterization of Finding," and determined the violation was a LOCA initiator.  The inspectors screened the safety significance to Green using IMC 0609, Appendix A, Exhibit 1, "LOCA Initiators," because,  
The inspectors evaluated the issue using IMC 0609.04, Initial Characterization of Finding,  
and determined the violation was a LOCA initiator.  The inspectors screened the safety  
significance to Green using IMC 0609, Appendix A, Exhibit 1, LOCA Initiators, because,  
after a reasonable assessment of degradation, it was determined that the violation did not  
after a reasonable assessment of degradation, it was determined that the violation did not  
result in exceeding the RCS leak rate for a small-break LOCA and did not affect any other  
result in exceeding the RCS leak rate for a small-break LOCA and did not affect any other  
systems used to mitigate a LOCA that could result in a total loss of function.  Therefore, the inspectors determined that the issue is of very low safety significance and concluded that the violation would be best characterized as Severity Level IV under the Traditional Enforcement  
systems used to mitigate a LOCA that could result in a total loss of function.  Therefore, the  
 
inspectors determined that the issue is of very low safety significance and concluded that the  
violation would be best characterized as Severity Level IV under the Traditional Enforcement  
process.  
process.  
   
   
Violation:  TS 3.4.13(A) requires, in part, that RCS operational leakage shall be limited to "no pressure boundary leakage." If pressure boundary leakage exists, then TS 3.4.13(A) limiting condition for operation action statement requires leakage to be reduced to within limits or
Violation:  TS 3.4.13(A) requires, in part, that RCS operational leakage shall be limited to no  
13  TS 3.4.13(B) requires Unit 2 to be in Mode 3 in 6 hours and TS 3.4.13(C) requires Unit 2 to be in Mode 4 within 12 hours. 
pressure boundary leakage.  If pressure boundary leakage exists, then TS 3.4.13(A) limiting  
condition for operation action statement requires leakage to be reduced to within limits or  


13
TS 3.4.13(B) requires Unit 2 to be in Mode 3 in 6 hours and TS 3.4.13(C) requires Unit 2 to
be in Mode 4 within 12 hours. 
   
   
Contrary to the above, for a period that began on an unknown date that was likely more than  
Contrary to the above, for a period that began on an unknown date that was likely more than  
12 hours before reactor shutdown on March 19, 2018, RCS pressure boundary leakage existed, and Entergy did not place Unit 2 in Mode 3 within 6 hours and in Mode 4 within 12 hours.   
12 hours before reactor shutdown on March 19, 2018, RCS pressure boundary leakage  
 
existed, and Entergy did not place Unit 2 in Mode 3 within 6 hours and in Mode 4 within  
12 hours.   
   
   
Enforcement Action:  This violation is being treated as an NCV, consistent with Section 2.3.2  
Enforcement Action:  This violation is being treated as an NCV, consistent with Section 2.3.2  
of the Enforcement Policy, because Entergy restored compliance, placed the issue in the corrective action program, was non-repetitive, was not NRC-identified, was not willful, and did not meet any minor examples in the Enforcement Manual, Section III, Appendix E.  
of the Enforcement Policy, because Entergy restored compliance, placed the issue in the  
corrective action program, was non-repetitive, was not NRC-identified, was not willful, and did  
not meet any minor examples in the Enforcement Manual, Section III, Appendix E.  
   
   
  Observation
   
71152 Annual Follow-up of  
Observation  
Selected Issues The inspectors determined several aspects of the fire protection program implementation had  
71152  
Annual Follow-up of  
Selected Issues  
The inspectors determined several aspects of the fire protection program implementation had  
improved, including the following:  
improved, including the following:  
 
(1) Entergy implemented reasonable actions to address each of the fire protection deficiencies listed in Other Activities - Baseline, Inspection Procedure 71152, Annual Follow-up of Selected Issues.  The most important positive action implemented was the  
(1) Entergy implemented reasonable actions to address each of the fire protection  
deficiencies listed in Other Activities - Baseline, Inspection Procedure 71152, Annual  
Follow-up of Selected Issues.  The most important positive action implemented was the  
hiring and training of three new people to perform the fire extinguisher and fire hose  
hiring and training of three new people to perform the fire extinguisher and fire hose  
preventive maintenances.  These personnel now report directly to the respective  
preventive maintenances.  These personnel now report directly to the respective  
responsible supervisors for portable fire extinguishers and fire hoses.  
responsible supervisors for portable fire extinguishers and fire hoses.  
  (2) Entergy initiated an accelerated schedule for replacing 100 percent of the onsite fire hoses with new hoses or hoses that have  
   
been hydrostatically tested this year.  The project is scheduled to be completed by December 2018.  
(2) Entergy initiated an accelerated schedule for replacing 100 percent of the onsite fire  
hoses with new hoses or hoses that have been hydrostatically tested this year.  The  
project is scheduled to be completed by December 2018.  
    
    
Notwithstanding the significant progress, the inspectors determined several procedure gaps/deficiencies exist, which could permit more fire hoses to exceed their hydrostatic test due date prior to their next inspection/replacement.  Entergy identified some of these through  
Notwithstanding the significant progress, the inspectors determined several procedure  
gaps/deficiencies exist, which could permit more fire hoses to exceed their hydrostatic test  
due date prior to their next inspection/replacement.  Entergy identified some of these through  
the corrective action program but had not resolved them.  The inspectors identified additional  
the corrective action program but had not resolved them.  The inspectors identified additional  
deficiencies in eight fire protection hose implementing procedures and discussed the  
deficiencies in eight fire protection hose implementing procedures and discussed the  
deficiencies with fire protection engineers.  CR-IP3-2018-01465 was initiated to fix the  
deficiencies with fire protection engineers.  CR-IP3-2018-01465 was initiated to fix the  
procedure deficiencies.  The inspectors independently inspected fire hoses at 40 onsite locations (including Units 1, 2, and 3, inside and outside of the radiological controlled areas, and the intake structure) and found each hose was properly hydrostatically tested within the  
procedure deficiencies.  The inspectors independently inspected fire hoses at 40 onsite  
locations (including Units 1, 2, and 3, inside and outside of the radiological controlled areas,  
and the intake structure) and found each hose was properly hydrostatically tested within the  
required periodicity.  Minor hose station deficiencies identified were promptly corrected  
required periodicity.  Minor hose station deficiencies identified were promptly corrected  
(CR-IP3-2018-01388, CR-IP2-2018-03309, CR-IP2-2018-03310, and CR-IP2-2018-03318).   
(CR-IP3-2018-01388, CR-IP2-2018-03309, CR-IP2-2018-03310, and CR-IP2-2018-03318).   
The inspectors determined the procedure deficiencies were minor, based on the fire hose inspection results and the progress Entergy had made on the project to replace all fire hoses in 2018.  
The inspectors determined the procedure deficiencies were minor, based on the fire hose  
14   Observation
inspection results and the progress Entergy had made on the project to replace all fire hoses  
71152 Semiannual Trend  
in 2018.  
Review The inspectors performed a review of Entergy's evaluation and corrective actions associated  
 
14  
Observation  
71152  
Semiannual Trend  
Review  
The inspectors performed a review of Entergys evaluation and corrective actions associated  
with the mitigating system performance index (MSPI) for the period from July 2016 through  
with the mitigating system performance index (MSPI) for the period from July 2016 through  
June 2018.  Specifically, the inspectors reviewed the corrective actions to verify they were  
June 2018.  Specifically, the inspectors reviewed the corrective actions to verify they were  
addressing the identified conditions adverse to quality.  The inspectors assessed Entergy's problem identification threshold, extent of condition reviews, and the prioritization and timeliness of corrective actions to determine whether Entergy was appropriately identifying,  
addressing the identified conditions adverse to quality.  The inspectors assessed Entergys
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  
problem identification threshold, extent of condition reviews, and the prioritization and  
the requirements of Entergy's corrective action program and 10 CFR Part 50, Appendix B.  In addition, the inspectors interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.  
timeliness of 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 Entergys corrective action program and 10 CFR Part 50, Appendix B.  In  
addition, the inspectors interviewed engineering personnel to assess the effectiveness of the  
implemented corrective actions.  
   
   
There were licensee-identified and NRC-identified performance indicator reporting  
There were licensee-identified and NRC-identified performance indicator reporting  
discrepancies documented during the reviewed timeframe.  Of note, there were two  
discrepancies documented during the reviewed timeframe.  Of note, there were two  
NRC-identified instances in which multiple discrepancies were reported which led to Entergy reviewing the prior three years of reported data and identifying additional discrepancies.  Also, there was a licensee-identified issue with the consolidated data entry (CDE) system where the system was reporting estimated values instead of actual values even though Entergy had  
NRC-identified instances in which multiple discrepancies were reported which led to Entergy  
reviewing the prior three years of reported data and identifying additional discrepancies.  Also,  
there was a licensee-identified issue with the consolidated data entry (CDE) system where the  
system was reporting estimated values instead of actual values even though Entergy had  
reported the actual values appropriately.  
reported the actual values appropriately.  
 
Instances of discrepancies were identified by the NRC in November 2016 (CR-IP2-2016-07044) and September 2017 (CR-IP2-2017-03169, CR-IP2-2017-03196, CR-IP2-2017-03203,  
Instances of discrepancies were identified by the NRC in November 2016 (CR-IP2-2016-
07044) and September 2017 (CR-IP2-2017-03169, CR-IP2-2017-03196, CR-IP2-2017-03203,  
CR-IP3-2017-04222, and CR-IP3-2017-04268).  The instance in 2016 involved all five MSPI  
CR-IP3-2017-04222, and CR-IP3-2017-04268).  The instance in 2016 involved all five MSPI  
systems.  This resulted in the MSPI resultant values getting farther from the Green/White  
systems.  This resulted in the MSPI resultant values getting farther from the Green/White  
threshold.  Entergy reviewed the prior three  
threshold.  Entergy reviewed the prior three years, found discrepancies in every quarter, and  
years, found discrepancies in every quarter, and submitted corrected data.  The instance in 2017 involved all five MSPI systems.  This resulted in a combination of the MSPI resultant values getting closer to and farther from the  
submitted corrected data.  The instance in 2017 involved all five MSPI systems.  This resulted  
in a combination of the MSPI resultant values getting closer to and farther from the  
Green/White threshold.  Entergy reviewed the prior three years, found discrepancies in every  
Green/White threshold.  Entergy reviewed the prior three years, found discrepancies in every  
quarter, and submitted corrected data.  The scope of the completed corrective actions for the  
quarter, and submitted corrected data.  The scope of the completed corrective actions for the  
2016 instance was narrowly focused on post-maintenance testing.  The scope of the 2017  
2016 instance was narrowly focused on post-maintenance testing.  The scope of the 2017  
review was broader and found errors in the overlapping 2015 and 2016 not related to post-maintenance testing.  
review was broader and found errors in the overlapping 2015 and 2016 not related to  
post-maintenance testing.  
   
   
Instances of discrepancies were identified by Entergy in October 2017 (CR-IP2-2017-03819)  
Instances of discrepancies were identified by Entergy in October 2017 (CR-IP2-2017-03819)  
and January 2018 (CR-IP2-2018-00144).  The instance in 2017 involved the way the CDE  
and January 2018 (CR-IP2-2018-00144).  The instance in 2017 involved the way the CDE  
system was reporting data after they had already been submitted.  A small portion of the high-pressure safety injection system was showing as estimated values instead of actual values.  This resulted in a combination of the MSPI resultant values getting closer to and  
system was reporting data after they had already been submitted.  A small portion of the  
high-pressure safety injection system was showing as estimated values instead of actual  
values.  This resulted in a combination of the MSPI resultant values getting closer to and  
farther from the green/white threshold.  Entergy checked all other parameters and corrected  
farther from the green/white threshold.  Entergy checked all other parameters and corrected  
the condition so that the actual values are being reported.  It is worth noting that the CDE  
the condition so that the actual values are being reported.  It is worth noting that the CDE  
system is maintained by a third party.  The instance in 2018 involved the number of hours of
critical operation for both units which is used in the unavailability and unreliability indexes of
MSPI.  This resulted in the MSPI resultant values getting closer to the green/white threshold;
however, Entergy corrected the data prior to it being reported.
Entergy did not identify specific trends or causes for the reasons of the discrepancies. 
However, Entergy took action in each case to correct the discrepancies after discovery. 
Entergy provided coaching to the individuals who compiled and reported the data.  Entergy


system is maintained by a third party. The instance in 2018 involved the number of hours of critical operation for both units which is used in the unavailability and unreliability indexes of MSPI.  This resulted in the MSPI resultant values getting closer to the green/white threshold; however, Entergy corrected the data prior to it being reported.
15
 
   
   
   
Entergy did not identify specific trends or causes for the reasons of the discrepancies. 
also provided training in June 2018 by a third party expert to all individuals involved in the  
However, Entergy took action in each case to correct the discrepancies after discovery.  Entergy provided coaching to the individuals who compiled and reported the data.  Entergy 
MSPI reporting process as a corrective action after the NRC-identified instance in 2017.  
15  also provided training in June 2018 by a third party expert to all individuals involved in the MSPI reporting process as a corrective action after the NRC-identified instance in 2017.  
   
   
It is apparent that there have been programmatic reporting issues with MSPI at the site.  None  
It is apparent that there have been programmatic reporting issues with MSPI at the site.  None  
Line 497: Line 884:
   
   
EXIT MEETINGS AND DEBRIEFS  
EXIT MEETINGS AND DEBRIEFS  
The inspectors confirmed that proprietary information was controlled to protect from public
disclosure.
  On July 18, 2018, the inspectors presented the quarterly resident inspector inspection results to Mr. Anthony Vitale, Site Vice President, and other members of the Entergy staff. 
   
   
 
The inspectors confirmed that proprietary information was controlled to protect from public
16   DOCUMENTS REVIEWED  
disclosure.
  Common Documents Used Indian Point Units 2 and 3, Updated Final Safety Analysis Report Indian Point Units 2 and 3, Individual Plant Examination  
On July 18, 2018, the inspectors presented the quarterly resident inspector inspection
results to Mr. Anthony Vitale, Site Vice President, and other members of the Entergy staff. 
 
16  
DOCUMENTS REVIEWED  
   
Common Documents Used  
Indian Point Units 2 and 3, Updated Final Safety Analysis Report  
Indian Point Units 2 and 3, Individual Plant Examination  
Indian Point Units 2 and 3, Individual Plant Examination of External Events  
Indian Point Units 2 and 3, Individual Plant Examination of External Events  
Indian Point Units 2 and 3, Technical Specifications and Bases  
Indian Point Units 2 and 3, Technical Specifications and Bases  
Indian Point Units 2 and 3, Technical Requirements Manual  
Indian Point Units 2 and 3, Technical Requirements Manual  
Indian Point Units 2 and 3, Control Room Narrative Logs Indian Point Units 2 and 3, Plan of the Day  
Indian Point Units 2 and 3, Control Room Narrative Logs  
  71111.01  Procedures IP-SMM-OP-104, Offsite Power Continuous Monitoring and Notification, Revision 13  
Indian Point Units 2 and 3, Plan of the Day  
   
71111.01  
   
Procedures  
IP-SMM-OP-104, Offsite Power Continuous Monitoring and Notification, Revision 13  
   
   
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
2018-03277*  
2018-03277*  
   
   
71111.04  Procedures 3-COL-RW-002, Service Water System, Revision 51 3-COL-SI-001, Safety Injection System, Revision 44  
71111.04  
3-SOP-SI-002, Filling the Refueling Water Storage Tank, Revision 12 3-SOP-SI-003, Recirculation and/or Purification of the Refueling Water Storage Tank, Revision 26  
   
Procedures  
3-COL-RW-002, Service Water System, Revision 51  
3-COL-SI-001, Safety Injection System, Revision 44  
3-SOP-SI-002, Filling the Refueling Water Storage Tank, Revision 12  
3-SOP-SI-003, Recirculation and/or Purification of the Refueling Water Storage Tank,  
Revision 26  
Condition Reports (CR-IP2-) (*initiated in response to inspection)
2018-02345* 2018-03138* 2018-03700* 2018-04013* 2018-04014*
Condition Reports (CR-IP3-) (*initiated in response to inspection)
2018-01180* 2018-01768* 2018-01864*
Miscellaneous
CES-2A, Engineering Standards Manual, Seismic Design of Electrical Conduit Supports for
Indian Point Units 2 and 3, Revision 3
71111.05AQ
Procedures
0-PT-M004, Fire Extinguisher Inspection, Revision 9
PFP-215, Unit 2 Primary Auxiliary Building and Fan House, 51-Foot and 68-Foot Elevations,
Revision 8
PFP-252, Unit 2 Cable Spreading Room, Control Building, 33-Foot Elevation, Revision 11
Condition Reports (CR-IP2-) (*initiated in response to inspection)
2018-03029* 2018-03261*
   
   
Condition Reports (CR-IP2-) (*initiated in response to inspection) 2018-02345* 2018-03138* 2018-03700* 2018-04013* 2018-04014*
Condition Reports (CR-IP3-) (*initiated in response to inspection)  
Condition Reports (CR-IP3-) (*initiated in response to inspection) 2018-01180* 2018-01768* 2018-01864*  
2018-01717*  


17
   
   
Miscellaneous CES-2A, Engineering Standards Manual, Seismic Design of Electrical Conduit Supports for Indian Point Units 2 and 3, Revision 3
71111.05AQ
Procedures 0-PT-M004, Fire Extinguisher Inspection, Revision 9 PFP-215, Unit 2 Primary Auxiliary Building and Fan House, 51-Foot and 68-Foot Elevations, Revision 8 PFP-252, Unit 2 Cable Spreading Room, Control Building, 33-Foot Elevation, Revision 11
   
   
Condition Reports (CR-IP2-) (*initiated in response to inspection) 2018-03029* 2018-03261*
Calculations
 
IP3-CALC-FP-02795, Combustible Loading Calculation for Indian Point 3 Hazards Analysis,
Revision 2
   
   
Condition Reports (CR-IP3-) (*initiated in response to inspection)
Engineering Evaluations  
2018-01717* 
DC-88-207, Indian Point 2 Fire Barrier Engineering Evaluation, Evaluation No. 1, dated  
17  Calculations IP3-CALC-FP-02795, Combustible Loading Calculation for Indian Point 3 Hazards Analysis, Revision 2
Engineering Evaluations DC-88-207, Indian Point 2 Fire Barrier Engineering Evaluation, Evaluation No. 1, dated  
March 24, 1989  
March 24, 1989  
   
   
Miscellaneous PFP-204, General Floor Plan - Primary Auxiliary Building, 15-Foot Elevation, Revision 0  
Miscellaneous  
PFP-204, General Floor Plan - Primary Auxiliary Building, 15-Foot Elevation, Revision 0  
PFP-315, Fuel Storage Bay Area - Fuel Storage Building, Revision 0  
PFP-315, Fuel Storage Bay Area - Fuel Storage Building, Revision 0  
PFP-316, General Floor Plan - Fuel Storage Building, Revision 12  
PFP-316, General Floor Plan - Fuel Storage Building, Revision 12  
PFP-354, Diesel Generator Building - Diesel Generators 31, 32, and 33, 15-Foot Elevation, Revision 0  
PFP-354, Diesel Generator Building - Diesel Generators 31, 32, and 33, 15-Foot Elevation,  
  71111.07  Procedures EN-DC-316, Heat Exchanger Performance and Condition Monitoring, Revision 10  
Revision 0  
   
71111.07  
   
Procedures  
EN-DC-316, Heat Exchanger Performance and Condition Monitoring, Revision 10  
EN-MA-118, Foreign Material Exclusion, Revision 12  
EN-MA-118, Foreign Material Exclusion, Revision 12  
   
   
Condition Reports (CR-IP2-)  
Condition Reports (CR-IP2-)  
2018-03349 2018-03350  
2018-03349  
 
2018-03350  
Miscellaneous
EC-77962, Evaluate Acceptability of Allowing Temporary Tube Plugs in 22 Component Cooling
Water Heat Exchanger to Next PM Activity, Revision 0
71111.08
   
   
Miscellaneous EC-77962, Evaluate Acceptability of Allowing Temporary Tube Plugs in 22 Component Cooling Water Heat Exchanger to Next PM Activity, Revision 0
Procedures
71111.08  Procedures WDI-STD-1013 (PDI-ISI-254-SE-NB), Remote Inservice Inspection of Rector Vessel Nozzle to Safe End, Nozzle to Pipe and Safe End to Pipe Welds Using the Nozzle Scanner,  
WDI-STD-1013 (PDI-ISI-254-SE-NB), Remote Inservice Inspection of Rector Vessel Nozzle to  
Safe End, Nozzle to Pipe and Safe End to Pipe Welds Using the Nozzle Scanner,  
Revision 3  
Revision 3  
   
   
Maintenance Orders/Work Orders WO 52568740 01 WO 52709908 01 WO 52708009 01 WO 0049815801  
Maintenance Orders/Work Orders  
 
WO 52568740 01  
WO 52709908 01  
WO 52708009 01  
WO 0049815801  
Calculation
IP-CALC-04-01420, FCX-0538, Calculation of Effective Degradation Years for the IP2 Reactor
Vessel Head by 2R18
   
   
Calculation IP-CALC-04-01420, FCX-0538, Calculation of Effective Degradation Years for the IP2 Reactor Vessel Head by 2R18
Miscellaneous
Miscellaneous ACTS; IP-001-18 Bobbin Coil:  0.720 - Standard Pull Speed  
ACTS; IP-001-18 Bobbin Coil:  0.720 - Standard Pull Speed  
ACTS; IPP 003-18 Bobbin Coil:  0.700  
ACTS; IPP 003-18 Bobbin Coil:  0.700  
ACTS; IPP-005-18 3-Coil RPC:  Mag. Biased 0.720 +PT  
ACTS; IPP-005-18 3-Coil RPC:  Mag. Biased 0.720 +PT  
ACTS; IPP-006-18 Single Coil RPC:  0.680 U-bend +PT
ACTS; PP-002-18 Bobbin Coil:  0.720 - Reduced Pull Speed
ACTS; PP-004-18 3-Coil RPC:  Standard 0.720 +PT
ACTS; PP-007-18 Bobbin Coil:  0.720-80 IPS Pull Speed
ANTS; IPP-B-18 3-Coil RPC - Standard and Mag. Biased (+PT/1 15/S080)


ACTS; IPP-006-18 Single Coil RPC:  0.680 U-bend +PT ACTS; PP-002-18 Bobbin Coil:  0.720 - Reduced Pull Speed ACTS; PP-004-18 3-Coil RPC:  Standard 0.720 +PT
18  
ACTS; PP-007-18 Bobbin Coil: 0.720-80 IPS Pull Speed
   
ANTS; IPP-B-18 3-Coil RPC - Standard and Mag. Biased (+PT/1 15/S080)  
   
18  ANTS; IPP-C-18 Single Coil +PT U-bend RPC  
ANTS; IPP-C-18 Single Coil +PT U-bend RPC  
ANTS; PP-A-18 All Bobbin Coil Probes  
ANTS; PP-A-18 All Bobbin Coil Probes  
ANTS; PP-D-18 Data Quality and General Instructions - All Probe Types Inspection Data Outlet Nozzle to DM Weld at 158 Degradation Weld No. RPVS-22-1A Inspection Data Outlet Nozzle to DM Weld at 22 Degradation Weld No. RPVS-24-1A  
ANTS; PP-D-18 Data Quality and General Instructions - All Probe Types  
Inspection Data Outlet Nozzle to DM Weld at 158 Degradation Weld No. RPVS-22-1A  
Inspection Data Outlet Nozzle to DM Weld at 22 Degradation Weld No. RPVS-24-1A  
IPP-R23-OH01-03-01, Ultrasonic Report Data Sheet  
IPP-R23-OH01-03-01, Ultrasonic Report Data Sheet  
IPP-R23-PEN3ET-SCAN2-12940, Eddy Current Report Data Sheet  
IPP-R23-PEN3ET-SCAN2-12940, Eddy Current Report Data Sheet  
Linearity Record Pulser Receiver 30028599  
Linearity Record Pulser Receiver 30028599  
Sensitivity Calibration Data Sheet No. SE-AX-DET-01 Sensitivity Calibration Data Sheet No. SE-AX-DET-03 Sensitivity Calibration Data Sheet No. SE-AX-DET-04  
Sensitivity Calibration Data Sheet No. SE-AX-DET-01  
Sensitivity Calibration Data Sheet No. SE-AX-DET-03  
Sensitivity Calibration Data Sheet No. SE-AX-DET-04  
Sensitivity Calibration Data Sheet No. SE-CIRC-DET-01  
Sensitivity Calibration Data Sheet No. SE-CIRC-DET-01  
Sensitivity Calibration Data Sheet No. SE-CIRC-DET-02  
Sensitivity Calibration Data Sheet No. SE-CIRC-DET-02  
SG-SGMP-18-2, Indian Point Unit 2 2RFO23 (Spring 2018) Steam Generator Degradation Assessment, Revision 0 WDI-CAL-002, Pulser/Receiver Linearity Procedure  
SG-SGMP-18-2, Indian Point Unit 2 2RFO23 (Spring 2018) Steam Generator Degradation  
 
Assessment, Revision 0  
71111.11  Procedures O-NF-206, Initial Criticality, Revision 8 O-NF-208, Zero Power Physics Testing, Revision 5  
WDI-CAL-002, Pulser/Receiver Linearity Procedure  
71111.11  
   
Procedures  
O-NF-206, Initial Criticality, Revision 8  
O-NF-208, Zero Power Physics Testing, Revision 5  
2-POP-1.1, Plant Heatup from Cold Shutdown, Revision 92  
2-POP-1.1, Plant Heatup from Cold Shutdown, Revision 92  
2-POP-1.2, Reactor Startup, Revision 61  
2-POP-1.2, Reactor Startup, Revision 61  
2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 96  
2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 96  
  71111.12  Condition Reports (CR-IP2-) (*initiated in response to inspection)  
   
71111.12  
   
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
2018-02246*  
2018-02246*  
  71111.13  Procedures 0-VLV-435-CWS, Annual Preventative Maintenance Testing and Replacement of Backflow  
   
Preventers, Revision 2 2-PT-R002B, Recirculation Sump Level, Revision 20 2-SOP-1.1.1, Vacuum Filling and Degassing the Reactor Coolant System, Revision 31  
71111.13  
3-PT-M62B, 480 Volt Undervoltage/Degraded Grid Protection System Bus 5A Functional, Revision 12 3-PT-M62C, 480 Volt Undervoltage/Degraded Grid Protection System Bus 2A and 3A Functional, Revision 13 3-PT-M62C, 480 Volt Undervoltage/Degraded Grid Protection System Bus 6A Functional, Revision 13 EN-OP-119, Protected Equipment Postings, Revision 9  
   
Procedures  
0-VLV-435-CWS, Annual Preventative Maintenance Testing and Replacement of Backflow  
Preventers, Revision 2  
2-PT-R002B, Recirculation Sump Level, Revision 20  
2-SOP-1.1.1, Vacuum Filling and Degassing the Reactor Coolant System, Revision 31  
3-PT-M62B, 480 Volt Undervoltage/Degraded Grid Protection System Bus 5A Functional,  
Revision 12  
3-PT-M62C, 480 Volt Undervoltage/Degraded Grid Protection System Bus 2A and 3A  
Functional, Revision 13  
3-PT-M62C, 480 Volt Undervoltage/Degraded Grid Protection System Bus 6A Functional,  
Revision 13  
EN-OP-119, Protected Equipment Postings, Revision 9  
EN-WM-104, On Line Risk Assessment, Revision 16  
EN-WM-104, On Line Risk Assessment, Revision 16  
IP-SMM-DC-910, Temporary Equipment, Revision 3 IP-SMM-DC-910, Temporary Equipment, Revision 9  
IP-SMM-DC-910, Temporary Equipment, Revision 3  
IP-SMM-DC-910, Temporary Equipment, Revision 9  
 
19
   
   
 
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
19  Condition Reports (CR-IP2-) (*initiated in response to inspection)  
2018-02240*  
2018-02240*  
  Condition Reports (CR-IP3-) (*initiated in response to inspection)  
   
Condition Reports (CR-IP3-) (*initiated in response to inspection)  
2018-01717*  
2018-01717*  
   
   
Drawings 310438, Safety Switch Level Indication and Pump Control Levels Reactor Cavity and Location Containment Sumps, Revision 10 9321-F-20343, Flow Diagram, City Water, Revision 43 9321-F-27513, Flow Diagram, Auxiliary Coolant System in Primary Auxiliary Building and Fuel Storage Building, Sheet No. 1, Revision 34 9321-F-33853, Electrical Distribution and Transmission System, Revision 21  
Drawings  
 
310438, Safety Switch Level Indication and Pump Control Levels Reactor Cavity and Location  
Miscellaneous Unit 3 Equipment Out of Service on Line Risk Assessment for May 1, 2018 Unit 3 Equipment Out of Service on Line Risk Assessment for May 17, 2018  
Containment Sumps, Revision 10  
9321-F-20343, Flow Diagram, City Water, Revision 43  
9321-F-27513, Flow Diagram, Auxiliary Coolant System in Primary Auxiliary Building and Fuel  
Storage Building, Sheet No. 1, Revision 34  
9321-F-33853, Electrical Distribution and Transmission System, Revision 21  
Miscellaneous  
Unit 3 Equipment Out of Service on Line Risk Assessment for May 1, 2018  
Unit 3 Equipment Out of Service on Line Risk Assessment for May 17, 2018  
Unit 3 Equipment Out of Service on Line Risk Assessment for June 6, 2018  
Unit 3 Equipment Out of Service on Line Risk Assessment for June 6, 2018  
Unit 3 Equipment Out of Service on Line Risk Assessment for June 17, 2018   
Unit 3 Equipment Out of Service on Line Risk Assessment for June 17, 2018   
  71111.15  Procedures IP-SMM-DC-910, Temporary Equipment, Revision 3  
   
71111.15  
   
Procedures  
IP-SMM-DC-910, Temporary Equipment, Revision 3  
OAP-035, Technical Specifications and Technical Requirements Manual - License Adherence  
OAP-035, Technical Specifications and Technical Requirements Manual - License Adherence  
and Use, Revision 15  
and Use, Revision 15  
   
   
Condition Reports (CR-IP2-) (*initiated in response to inspection) 2018-02541* 2018-02548* 2018-02581* 2018-02588*  
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
 
2018-02541* 2018-02548* 2018-02581* 2018-02588*  
Condition Reports (CR-IP3-) 2018-01448* 2018-01530 2018-01552 2018-01570 2018-01878*  
  71111.18  Condition Reports (CR-IP2-)  
Condition Reports (CR-IP3-)  
2018-01448* 2018-01530  
2018-01552  
2018-01570  
2018-01878*  
   
71111.18  
   
Condition Reports (CR-IP2-)  
2017-05238  
2017-05238  
   
   
Maintenance Orders/Work Orders  
Maintenance Orders/Work Orders  
WO 00491522  
WO 00491522  
Drawings
9321-2561, Containment Building Residual Heat Exchanger Piping Plans and Sections (A.C.S.),
Revision 23
9321-2720, Flow Diagram Auxiliary Coolant System Updated Final Safety Analysis Report,
Figure No. 9.3-1 (Sheet 2), Revision 93
Miscellaneous
EC-75309, Install Temporary Clamp to Facilitate Weld Repair at Valve A-98, Revision 2
EC-75312, Engineering to Evaluate the Repair of Weld Per Code Case N-666 on a Leaky
Socket Weld at Vent Valve A98 on Residual Heat Removal Line 361, Revision 0


  Drawings 9321-2561, Containment Building Residual Heat Exchanger Piping Plans and Sections (A.C.S.), Revision 23 9321-2720, Flow Diagram Auxiliary Coolant System Updated Final Safety Analysis Report, Figure No. 9.3-1 (Sheet 2), Revision 93
20
Miscellaneous EC-75309, Install Temporary Clamp to Facilitate Weld Repair at Valve A-98, Revision 2 EC-75312, Engineering to Evaluate the Repair of Weld Per Code Case N-666 on a Leaky Socket Weld at Vent Valve A98 on Residual Heat Removal Line 361, Revision 0  
   
20  EC-76614, Evaluate Using Freeze Seal Isolation on 6-Inch Line 13 to Perform Maintenance on  
   
EC-76614, Evaluate Using Freeze Seal Isolation on 6-Inch Line 13 to Perform Maintenance on  
Valve 769, Revision 0  
Valve 769, Revision 0  
   
   
71111.19   
71111.19  
   
Procedures  
Procedures  
2-AOP-RCP-1, Reactor Coolant Pump Malfunction, Revision 16  
2-AOP-RCP-1, Reactor Coolant Pump Malfunction, Revision 16  
2-PT-Q013, Inservice Valve Tests, Revision 52  
2-PT-Q013, Inservice Valve Tests, Revision 52  
2-PT-R027C, WCPPS Local Leak Rate, Revision 11 3-PC-R06C, Rod Position Indication System "Hot Span" Confirmation, Revision 12 3-PT-CS030, Atmospheric Steam Dump Valves Stroke Test, Revision 17  
2-PT-R027C, WCPPS Local Leak Rate, Revision 11  
3-SOP-RC-001, Full Length Rod Control and Rod Position Indication System Operation, Revision 29 EN-OP-104, Operability Determination Process, Revision 14 SEP-APJ-006, Indian Point Energy Center Primary Containment Leakage Rate Testing (Appendix J) Program Plan, Revision 4 SOP-1.3, Reactor Coolant Pump Startup and Shutdown, Revision 47   
3-PC-R06C, Rod Position Indication System Hot Span Confirmation, Revision 12  
 
3-PT-CS030, Atmospheric Steam Dump Valves Stroke Test, Revision 17  
3-SOP-RC-001, Full Length Rod Control and Rod Position Indication System Operation,  
Revision 29  
EN-OP-104, Operability Determination Process, Revision 14  
SEP-APJ-006, Indian Point Energy Center Primary Containment Leakage Rate Testing  
(Appendix J) Program Plan, Revision 4  
SOP-1.3, Reactor Coolant Pump Startup and Shutdown, Revision 47   
   
   
Condition Reports (CR-IP2-)  
Condition Reports (CR-IP2-)  
2016-01973 2016-02383 2018-02008 2018-02359  
2016-01973  
 
2016-02383  
2018-02008  
2018-02359  
   
   
Condition Reports (CR-IP3-) (*initiated in response to inspection)  
Condition Reports (CR-IP3-) (*initiated in response to inspection)  
2017-05301 2018-01261 2018-01438 2018-01463*  
2017-05301  
 
2018-01261  
2018-01438  
2018-01463*  
   
   
Maintenance Orders/Work Orders WO 00441335  WO 00501265  WO 52697548  WO 00363219-01  
Maintenance Orders/Work Orders  
 
WO 00441335   
WO 00501265   
WO 52697548   
WO 00363219-01  
WO 003623215-01  
WO 003623215-01  
   
   
Miscellaneous 2-PT-Q013, Inservice Valve Tests, Revision 52, dated Mach 19, 2018 2-PT-R027C-DS001, Containment Pressure Relief Valves PCV-1190 and PCV-1191, dated
Miscellaneous  
March 29, 2018 2-PT-R027C-DS001, Containment Pressure Relief Valves PCV-1190 and PCV-1191, dated
2-PT-Q013, Inservice Valve Tests, Revision 52, dated Mach 19, 2018  
March 31, 2016 2-PT-R027C-DS002, Containment Pressure Relief Valves PCV-1191 and PCV-1192, dated  
2-PT-R027C-DS001, Containment Pressure Relief Valves PCV-1190 and PCV-1191, dated  
March 29, 2018  
March 29, 2018  
  71111.20  Procedures  
2-PT-R027C-DS001, Containment Pressure Relief Valves PCV-1190 and PCV-1191, dated
March 31, 2016
2-PT-R027C-DS002, Containment Pressure Relief Valves PCV-1191 and PCV-1192, dated
March 29, 2018
   
71111.20  
   
Procedures  
0-NF-206, Initial Criticality, Revision 8  
0-NF-206, Initial Criticality, Revision 8  
0-NF-208, Zero Power Physics Testing, Revision 5  
0-NF-208, Zero Power Physics Testing, Revision 5  
Line 646: Line 1,180:
2-POP-1.2, Reactor Startup, Revision 61  
2-POP-1.2, Reactor Startup, Revision 61  
2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 96  
2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 96  
2-POP-3.1, Plant Shutdown from 45 Percent Power, Revision 60 2-POP-3.3, Plant Cooldown, Hot to Cold Shutdown, Revision 83 2-POP-3.4, Secondary Plant Shutdown, Revision 10  
2-POP-3.1, Plant Shutdown from 45 Percent Power, Revision 60  
2-POP-3.3, Plant Cooldown, Hot to Cold Shutdown, Revision 83  
2-POP-3.4, Secondary Plant Shutdown, Revision 10  
2-POP-4.1, Operation at Cold Shutdown, Revision 8  
2-POP-4.1, Operation at Cold Shutdown, Revision 8  


 
21
21  2-POP-4.2, Operation Below 20 Percent Pressurizer Level with Fuel in the Reactor, Revision 12 2-POP-4.3, Operation without Fuel in the Reactor, Revision 3  
 
Condition Reports (CR-IP2-) (*initiated in response to inspection) 2018-02205* 2018-02671*  
2-POP-4.2, Operation Below 20 Percent Pressurizer Level with Fuel in the Reactor, Revision 12  
  71111.22  Procedures 0-PT-6Y001, B.5.b Diesel Driven Pump Capacity Test, Revision 2  
2-POP-4.3, Operation without Fuel in the Reactor, Revision 3  
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
2018-02205* 2018-02671*  
   
71111.22  
   
Procedures  
0-PT-6Y001, B.5.b Diesel Driven Pump Capacity Test, Revision 2  
3-PT-M13B1, Reactor Protection Logic Channel Functional Test (Reactor Power Greater Than  
3-PT-M13B1, Reactor Protection Logic Channel Functional Test (Reactor Power Greater Than  
35 Percent - P8), Revision 24 3-PT-Q120A, 31 Auxiliary Feedwater Pump, Revision 17  
35 Percent - P8), Revision 24  
3-PT-Q120A, 31 Auxiliary Feedwater Pump, Revision 17  
EN-HU-106, Procedure and Work Instruction Use and Adherence, Revision 6  
EN-HU-106, Procedure and Work Instruction Use and Adherence, Revision 6  
  Condition Reports (CR-IP2-)  
   
2018-03672 2018-03675  
Condition Reports (CR-IP2-)  
 
2018-03672  
2018-03675  
   
   
Condition Reports (CR-IP3-)  
Condition Reports (CR-IP3-)  
2018-00214  
2018-00214  
   
   
Maintenance Orders/Work Orders WO 52810279  WO 52811023  
Maintenance Orders/Work Orders  
 
WO 52810279   
71124.07  Procedures 0-CY-1930, REMP Air Samples, Revision 2  
WO 52811023  
71124.07  
   
Procedures  
0-CY-1930, REMP Air Samples, Revision 2  
0-CY-1940, REMP Broad Leaf Vegetation Samples, Revision 0  
0-CY-1940, REMP Broad Leaf Vegetation Samples, Revision 0  
0-CY-1950, REMP Collection, Preparation, and Analysis of Water Samples, Revision 0 0-EV-DD-101, Meteorological Systems Manual, Revision 1 EN-CY-111, Radiological Groundwater Protection Program, Revision 8  
0-CY-1950, REMP Collection, Preparation, and Analysis of Water Samples, Revision 0  
0-EV-DD-101, Meteorological Systems Manual, Revision 1  
EN-CY-111, Radiological Groundwater Protection Program, Revision 8  
IP-SMM-CY-110, Radiological Groundwater Monitoring Program, Revision 9  
IP-SMM-CY-110, Radiological Groundwater Monitoring Program, Revision 9  
   
   
Condition Reports (CR-IP2-)  
Condition Reports (CR-IP2-)  
2017-00635 2018-02134 2018-02272 2018-03578 2018-03624  
2017-00635  
  Condition Reports (CR-IP3-)  
2018-02134  
2018-02272  
2018-03578  
2018-03624  
   
Condition Reports (CR-IP3-)  
2017-03403  
2017-03403  
Miscellaneous
Environmental Dosimetry Company Annual Quality Assurance Status Report, dated March 7,
2018
Indian Point Energy Center Offsite Dose Calculation Manual, Revision 4
NAQA-17-0038, Audit Report 17-007 for the Audit of Environmental Dosimetry Company, dated
September 1, 2017
71151
Procedures
NEI 99-02, Regulatory Assessment Performance Indicator Guidelines, Revision 7


22
   
   
Miscellaneous Environmental Dosimetry Company Annual Quality Assurance Status Report, dated March 7, 2018 Indian Point Energy Center Offsite Dose Calculation Manual, Revision 4
NAQA-17-0038, Audit Report 17-007 for the Audit of Environmental Dosimetry Company, dated September 1, 2017
   
   
71151 
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
Procedures NEI 99-02, Regulatory Assessment Performance Indicator Guidelines, Revision 7 
2018-01129  
22  Condition Reports (CR-IP2-) (*initiated in response to inspection) 2018-01129 2018-02739* 2018-03931  
2018-02739* 2018-03931  
71152
   
   
71152 
Procedures
Procedures 0-OSP-FP-001, Hydrostatic Testing of Firehose, Revision 0  
0-OSP-FP-001, Hydrostatic Testing of Firehose, Revision 0  
0-PT-M001, Fire Brigade Equipment Inventory and Inspection, Revision 18  
0-PT-M001, Fire Brigade Equipment Inventory and Inspection, Revision 18  
0-PT-M004, Fire Extinguisher Inspection, Revision 9 2-PT-M112, Fire Protection Flow Path and Hose Station Equipment Verification, Revision 6 AP-64.1, Fire Protection/Appendix R Systems and Components Governed by Technical Requirements Manual and Technical Specifications, Revision 5 SAO-703, Fire Protection Impairment Criteria and Surveillance, Revision 35  
0-PT-M004, Fire Extinguisher Inspection, Revision 9  
 
2-PT-M112, Fire Protection Flow Path and Hose Station Equipment Verification, Revision 6  
Condition Reports (CR-IP2-) (*initiated in response to inspection) 2016-00509 2017-04090 2018-03262* 2018-03306* 2018-03309* 2018-03310*  
AP-64.1, Fire Protection/Appendix R Systems and Components Governed by Technical  
Requirements Manual and Technical Specifications, Revision 5  
SAO-703, Fire Protection Impairment Criteria and Surveillance, Revision 35  
Condition Reports (CR-IP2-) (*initiated in response to inspection)  
2016-00509  
2017-04090  
2018-03262* 2018-03306* 2018-03309* 2018-03310*  
2018-03318*  
2018-03318*  
   
   
Condition Reports (CR-IP3-) (*initiated in response to inspection) 2018-01112 2018-01388* 2018-01465*  
Condition Reports (CR-IP3-) (*initiated in response to inspection)  
 
2018-01112  
2018-01388* 2018-01465*  
   
   
Maintenance Orders/Work Orders WO 00379953  WO 00383371  WO 00417368  WO 52620568  
Maintenance Orders/Work Orders  
WO 52686482  WO 52692709  WO 52709916  WO 52723346  
WO 00379953   
 
WO 00383371   
WO 00417368   
WO 52620568  
WO 52686482   
WO 52692709   
WO 52709916   
WO 52723346  
WO 52810046  
WO 52810046  
  Miscellaneous EC-76593, Indian Point Energy Center Pre-Fire Plan Updates, Revision 0  
   
Indian Point Unit 3 Technical Requirements Manual, Section 3.7.A.5, Fire Hose Stations, Revision 3 Indian Point Unit 3 Technical Requirements Manual, Section 3.7.A.6, Yard Fire Hydrants and Hydrant Hose Houses, Revision 2 NFPA-1962, Standard for the Care, Use, Inspection, Service Testing, and Replacement of Fire Hose, Couplings, Nozzles, and Fire Hose Appliances, 2018 Edition  
Miscellaneous  
  71153  Procedures EN-FAP-MP-009, Enhanced Procurement Process for Non-Safety Related Critical Spares, Revision 3 EN-MP-100, Critical Procurements, Revision 13  
EC-76593, Indian Point Energy Center Pre-Fire Plan Updates, Revision 0  
  Condition Reports (CR-IP3-)  
Indian Point Unit 3 Technical Requirements Manual, Section 3.7.A.5, Fire Hose Stations,  
2017-05133 2018-00539
Revision 3  
Indian Point Unit 3 Technical Requirements Manual, Section 3.7.A.6, Yard Fire Hydrants and  
Hydrant Hose Houses, Revision 2  
NFPA-1962, Standard for the Care, Use, Inspection, Service Testing, and Replacement of Fire  
Hose, Couplings, Nozzles, and Fire Hose Appliances, 2018 Edition  
   
71153  
   
Procedures  
EN-FAP-MP-009, Enhanced Procurement Process for Non-Safety Related Critical Spares,  
Revision 3  
EN-MP-100, Critical Procurements, Revision 13  
   
Condition Reports (CR-IP3-)  
2017-05133  
2018-00539
}}
}}

Latest revision as of 16:04, 5 January 2025

Integrated Inspection Report 05000247/2018002 and 05000286/2018002
ML18213A153
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 08/01/2018
From: Daniel Schroeder
Reactor Projects Branch 2
To: Vitale A
Entergy Nuclear Operations
Schroeder D
References
IR 2018002
Download: ML18213A153 (25)
v • d • e


See also: IR 05000247/2018002

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION I

2100 RENAISSANCE BOULEVARD, SUITE 100

KING OF PRUSSIA, PA 19406-2713

August 1, 2018

Mr. Anthony J. Vitale

Site Vice President

Entergy Nuclear Operations, Inc.

Indian Point Energy Center

450 Broadway, General Services Building

P.O. Box 249

Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING - INTEGRATED INSPECTION

REPORT 05000247/2018002 AND 05000286/2018002

Dear Mr. Vitale:

On June 30, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at

Indian Point Nuclear Generating (Indian Point), Units 2 and 3. On July 18, 2018, the NRC

inspectors discussed the results of this inspection with you and other members of your staff.

The results of this inspection are documented in the enclosed report.

No NRC-identified or self-revealing findings were identified during this inspection. The NRC

inspectors documented one violation, which was determined to be Severity Level IV in this

report. The NRC is treating this violation as a non-cited violation (NCV) consistent with

Section 2.3.2.a of the Enforcement Policy.

If you contest the violation or significance of the NCV, you should provide a response within

30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear

Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with

copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC

Resident Inspector at Indian Point.

A. Vitale

2

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), Part 2.390, Public

Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Daniel L. Schroeder, Chief

Reactor Projects Branch 2

Division of Reactor Projects

Docket Numbers: 50-247 and 50-286

License Numbers: DPR-26 and DPR-64

Enclosure:

Inspection Report 05000247/2018002 and

05000286/2018002

cc w/encl: Distribution via ListServ

ML18213A153

SUNSI Review

Non-Sensitive

Sensitive

Publicly Available

Non-Publicly Available

OFFICE

RI/DRP

RI/DRP

RI/DRP

NAME

BHaagensen via email TSetzer

DSchroeder

DATE

7/25/18

7/25/18

8/1/18

1

Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

Inspection Report

Docket Numbers:

50-247 and 50-286

License Numbers:

DPR-26 and DPR-64

Report Numbers:

05000247/2018002 and 05000286/2018002

Licensee:

Entergy Nuclear Northeast (Entergy)

Facility:

Indian Point Nuclear Generating, Units 2 and 3

Location:

450 Broadway, General Services Building

Buchanan, NY 10511-0249

Inspection Dates:

April 1, 2018 to June 30, 2018

Inspectors:

B. Haagensen, Senior Resident Inspector

A. Siwy, Resident Inspector

J. Vazquez, Resident Inspector

J. Fuller, Acting Resident Inspector

D. Kern, Senior Reactor Inspector

M. Modes, Senior Reactor Inspector

L. McKown, Acting Resident Inspector

S. Wilson, Health Physicist

Approved By:

Daniel L. Schroeder, Chief

Reactor Projects Branch 2

Division of Reactor Projects

2

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring Entergys performance

at Indian Point Nuclear Generating, Units 2 and 3, by conducting the baseline inspections

described in this report in accordance with the Reactor Oversight Process. The Reactor

Oversight Process is the NRCs program for overseeing the safe operation of commercial

nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more

information. NRC-identified and self-revealing findings, violations, and additional items are

summarized in the table below.

List of Findings and Violations

Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head

Penetration #3

Cornerstone

Severity

Cross-Cutting Aspect

Inspection

Results

Section

Not Applicable

Severity Level IV

Non-Cited Violation (NCV)

05000286/2018-002-01

Closed

Not Applicable

71111.15

A self-revealing Severity Level IV NCV of Technical Specification (TS) 3.4.13.a, Reactor

Coolant System Operational Leakage, was identified when Entergy operated the reactor in

Mode 1 with pressure boundary leakage for a period of time longer than the allowable limiting

condition of operation. Specifically, a leak in the J-weld around reactor pressure vessel

(RPV) head penetration #3 occurred during the last operating cycle and was not identified

until after the reactor was shutdown for a refueling outage.

Additional Tracking Items

Type

Issue number

Title

Inspection

Results

Section

Status

LER 05000286/2017-004-00

Reactor Trip Due to Main

Generator Loss of Field

71153

Closed

LER 05000286/2018-001-00

Reactor Trip Due to Main

Generator Loss of Excitation

71153

Closed

3

TABLE OF CONTENTS

INSPECTION SCOPES ................................................................................................................ 4

REACTOR SAFETY ..................................................................................................................... 4

RADIATION SAFETY ................................................................................................................... 8

OTHER ACTIVITIES - BASELINE .............................................................................................. 9

INSPECTION RESULTS ............................................................................................................ 10

EXIT MEETINGS AND DEBRIEFS ............................................................................................ 15

DOCUMENTS REVIEWED ......................................................................................................... 16

4

PLANT STATUS

Unit 2 began the inspection period during refueling outage 2RFO23 which lasted 33 days. Upon

completion of the outage, the operators restarted Unit 2 on April 19, 2018, and increased power

slowly to approximately eight percent power when the unit was tripped by the operators due to

an issue with the main turbine control valves. The unit was returned to rated thermal power on

April 25, 2018. On June 22, 2018, the operators reduced power to approximately 62 percent

power to repair an oil leak on the 22 main boiler feed pump. The unit was returned to rated

thermal power on June 23, 2018, and remained at or near rated thermal power for the

remainder of the inspection period.

Unit 3 operated at or near rated thermal power for the entire inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in

effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with

their attached revision histories are located on the public website at http://www.nrc.gov/reading-

rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared

complete when the IP requirements most appropriate to the inspection activity were met

consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection

Program - Operations Phase. The inspectors performed plant status activities described in

IMC 2515, Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem

Identification and Resolution. The inspectors reviewed selected procedures and records,

observed activities, and interviewed personnel to assess Entergys performance and

compliance with Commission rules and regulations, license conditions, site procedures, and

standards.

REACTOR SAFETY

71111.01 - Adverse Weather Protection

Summer Readiness (1 Sample)

The inspectors evaluated summer readiness of offsite and alternating current power

systems at Units 2 and 3 on May 17, 2018.

71111.04 - Equipment Alignment

Partial Walkdown (4 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following

systems/trains:

Unit 2

(1) Appendix R diesel generator on June 12, 2018

5

Unit 3

(2) 33 auxiliary boiler feedwater pump on June 5, 2018

(3) Service water Zurn strainers on June 18, 2018

(4) Refueling water storage tank on June 28, 2018

Complete Walkdown (1 Sample)

The inspectors evaluated system configurations during a complete walkdown of the Unit 2

emergency diesel generators on June 27, 2018.

71111.05AQ - Fire Protection Annual/Quarterly

Quarterly Inspection (5 Samples)

The inspectors evaluated fire protection program implementation in the following selected

areas:

Unit 2

(1) Primary auxiliary building and fan house pre-fire plan (PFP-215) on May 15, 2018

(2) Cable spreading room (PFP-252) on May 31, 2018

(3) Primary auxiliary building, 15-foot elevation (PFP-204), on June 7, 2018

Unit 3

(4) Fuel storage building (PFP-315 and PFP-316) on June 6, 2018

(5) Diesel generator building (PFP-354) on June 29, 2018

71111.07A - Heat Sink Performance

Heat Sink (1 Sample)

The inspectors observed Entergys inspections and the state of cleanliness of the tubes of

the 22 component cooling water heat exchanger at Unit 2 on May 22, 2018.

71111.08P - Inservice Inspection Activities (1 Sample)

The inspectors evaluated Entergys non-destructive examination and welding activities at

Unit 2 by reviewing the following examinations from April 24 to April 27, 2018:

(1) Volumetric Examinations

The ultrasonic, eddy current, and liquid penetrant examination of the J-weld and

associated interfaces of RPV head penetration #3, with subsequent repair by weld

surface seal weld

Ultrasonic examination of the dissimilar welds of reactor nozzles 24-1A and 22-1A

6

(2) The inspectors reviewed Entergys boric acid corrosion control program performance.

(3) The inspectors reviewed Entergys steam generator aging management program,

including eddy current testing of steam generator tubes.

71111.11 - Licensed Operator Requalification Program and Licensed Operator Performance

Operator Requalification (1 Sample)

Unit 3

(1) The inspectors observed and evaluated a crew of licensed operators in the plants

simulator during licensed operator requalification training on June 5, 2018.

Operator Performance (1 Sample)

Unit 2

(1) The inspectors observed and evaluated operator performance during 2RFO23 reactor

startup on April 19, 2018.

71111.12 - Maintenance Effectiveness

Routine Maintenance Effectiveness (1 Sample)

The inspectors evaluated the effectiveness of routine maintenance activities associated

with the following equipment and/or safety significant functions:

Unit 2

(1) CR-IP2-2018-2246, Containment structure on June 25, 2018

71111.13 - Maintenance Risk Assessments and Emergent Work Control (8 Samples)

The inspectors evaluated the risk assessments for the following planned and emergent

work activities:

Unit 2

(1) Planned Yellow risk associated with freeze seal to affect repairs to MOV-769,

component cooling water to reactor coolant pump seal cooling, isolation valve on April 7,

2018

(2) Planned Yellow risk for vacuum fill of the reactor coolant system (RCS) on April 14, 2018

(3) Risk for maintenance on containment recirculation sump level transmitter LT-939 on

April 20, 2018

(4) Unplanned Yellow risk for loss of bus 2 on 13.8kV feeder 13W93 on June 1, 2018

Unit 3

(5) 13.8kV feeders out of service for planned gas turbine auto transformer maintenance on

May 1, 2018

(6) Planned Yellow risk for 480 volt degraded voltage relay testing on May 17, 2018

7

(7) Planned Yellow risk for charging pump backflow preventer maintenance on June 6, 2018

(8) Planned Yellow risk for 32 component cooling water pump oil sampling on June 13,

2018

71111.15 - Operability Determinations and Functionality Assessments (5 Samples)

The inspectors evaluated the following operability determinations and functionality

assessments:

Unit 2

(1) CR-IP2-2018-02127, Baffle bolt operability assessment during 2RFO23 on April 17,

2018

(2) CR-IP2-2018-1944, RPV head penetration #3 through-wall leak past operability on April

17, 2018

(3) CR-IP2-2018-2359, PCV-1191 containment pressure relief valve on April 19, 2018

Unit 3

(4) CR-IP3-2018-01448, 34 atmospheric dump valve past-operability associated with

unsecured nearby equipment on May 23, 2018

(5) CR-IP3-2018-01552, 33 service water pump operability during 32 emergency diesel

generator testing with 35 service water pump out of service on June 5, 2018

71111.18 - Plant Modifications (2 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) Engineering Change (EC) Package 76614 - Evaluate using freeze seal isolation on

6-inch line 13 to perform maintenance on valve 769 (temporary modification) at Unit 2 on

April 7, 2018

(2) EC Package 75312 - Engineering weld-repaired leaking RHR line A-98 per ASME Code

Case N-666 (permanent modification) on at Unit 2 on April 24, 2018

71111.19 - Post-Maintenance Testing (13 Samples)

The inspectors evaluated post-maintenance testing for the following maintenance/repair

activities:

Unit 2

(1)

MOV-744, Anchor Darling double-disc gate valve inspection following repairs on

April 2, 2018

(2)

MOV-769, Anchor Darling double-disc gate valve inspection following repairs on

April 2, 2018

(3)

MOV-746, Anchor Darling double-disc gate valve inspection following repairs on

April 2, 2018

(4)

MOV-784, Anchor Darling double-disc gate valve inspection of condition and

performance monitoring following minor valve maintenance on April 7, 2018

8

(5)

MOV-885B, Anchor Darling double-disc gate valve inspection of condition and

performance monitoring following minor valve maintenance on April 7, 2018

(6)

Containment pressure relief valve PCV-1191 following repairs for local leak-rate test

failure on April 12, 2018

(7)

RPV head penetration #3 weld repairs on April 17, 2018

(8)

24 reactor coolant pump motor repair on April 18, 2018

(9)

Control rod H6 following individual rod position indication repairs on April 18, 2018.

(10) 22 reactor coolant pump motor repair on April 18, 2018

(11) Control rod G3 drive coil stack replacement on April 18, 2018

Unit 3

(12) Control rod M4 individual rod position indication post-maintenance testing following

calibration to resolve indication issues on May 8, 2018

(13) 33 atmospheric dump valve post-maintenance testing following instrument airline

repairs on May 23, 2018

71111.20 - Refueling and Other Outage Activities (1 Sample)

The inspectors evaluated Unit 2 2RFO23 activities from March 18 to April 21, 2018.

71111.22 - Surveillance Testing

The inspectors evaluated the following surveillance tests:

Routine (3 Samples)

(1) 3-PT-M13B1, Reactor Protection Train B Logic Test, at Unit 3 on May 7, 2018

(2) 3-PT-Q120A, 31 Motor Driven Auxiliary Boiler Feedwater Pump Functional Test, at

Unit 3 on May 21, 2018

(3) 0-PT-6Y001, B.5.b Diesel Driven Pump Capacity Test, at Units 2 and 3 on June 13,

2018

RADIATION SAFETY

71124.07 - Radiological Hazard Assessment and Exposure Controls

Site Inspection (1 Sample)

The inspectors verified that the radiological environmental monitoring program quantified the

impact of radioactive effluent releases to the environment and sufficiently validated the

integrity of the radioactive gaseous and liquid effluent release program. The inspectors

observed environmental sampling of groundwater, surface water, broad leaf vegetation, and

air samples.

The inspectors verified that the radiological environmental monitoring program was

consistently implemented with Entergys TSs and/or offsite dose calculation manual and

validated that the program meets the design objectives in Appendix I to Title 10 of the Code

of Federal Regulations (10 CFR), Part 2.39010.

9

Groundwater Protection Initiative Implementation (1 Sample)

The inspectors verified that Entergy is continuing to implement the voluntary Nuclear Energy

Institute/Industry Groundwater Protection Initiative.

OTHER ACTIVITIES - BASELINE

71151 - Performance Indicator Verification

The inspectors verified Entergys performance indicators submittals listed below for the

period from April 1, 2017, through March 31, 2018 (6 Samples).

Unit 2

(1) Unplanned Scrams Per 7000 Critical Hours (IE01)

(2) Unplanned Power Changes Per 7000 Critical Hours (IE03)

(3) Unplanned Scrams with Complications (IE04)

Unit 3

(4) Unplanned Scrams Per 7000 Critical Hours (IE01)

(5) Unplanned Power Changes Per 7000 Critical Hours (IE03)

(6) Unplanned Scrams with Complications (IE04)

71152 - Problem Identification and Resolution

Semiannual Trend Review (1 Sample)

The inspectors reviewed Entergys corrective action program for trends that might be

indicative of a more significant safety issue.

Annual Follow-up of Selected Issues (1 Sample)

The inspectors reviewed Entergys implementation of its corrective action program related to

the following issue:

(1) Several fire protection deficiencies were identified for Units 2 and 3 during the last year,

including portable fire extinguisher preventive maintenance not performed documented

in the Indian Point Integrated Inspection Report 05000247/2017007 and

05000286/2017007 (Agencywide Documents Access and Management System

(ADAMS) Accession Number ML17341A080), fire hose hydrostatic testing and

preventive maintenance not performed, and pre-fire plan inaccuracy. Condition reports

associated with the above issues included CR-IP2-2017-02945, CR-IP2-2017-04090,

CR-IP2-2018-02800, CR-IP3-2017-04084, CR-IP3-2017-05035, CR-IP3-2017-05040,

CR-IP3-2018-00463, and CR-IP3-2018-00533.

10

71153 - Follow-up of Events and Notices of Enforcement Discretion

Events (1 Sample)

The inspectors evaluated response to the following event:

(1) RPV head penetration #3 flaw, 50.72 report, at Unit 2 on April 2, 2018

Licensee Event Reports (LERs) (2 Samples)

The inspectors evaluated the following LERs:

(1) LER 05000286/2017-004-00, Unit 3 Reactor Trip Due to Main Generator Loss of Field

(ADAMS Accession Number ML17363A212)

Entergy conducted a causal evaluation and determined that the Thyristor Firing

Module B drawer had failed. Entergy made repairs to the failed components and

returned the system to service with additional monitoring equipment. The inspectors did

not identify any new issues during the review of the LER. The inspectors determined

that it was not reasonable to foresee and prevent the cause discussed in the LER.

Therefore, no performance deficiency was identified. The inspectors also concluded that

no violation of NRC requirements occurred.

(2) LER 05000286/2018-001-00, Unit 3 Reactor Trip Due to Main Generator Loss of

Excitation (ADAMS Accession Number ML18110A088)

Entergy conducted a causal evaluation and determined that the Thyristor Firing

Module A drawer had failed. Entergy replaced both Thyristor Firing Module drawers and

returned the system to service with additional monitoring equipment. The inspectors did

not identify any new issues during the review of the LER. The inspectors determined

that it was not reasonable to foresee and prevent the cause discussed in the LER.

Therefore, no performance deficiency was identified. The inspectors also concluded that

no violation of NRC requirements occurred.

Personnel Performance (1 Sample)

The inspectors evaluated response during the following non-routine evolutions or transients.

(1) Unit 2 reactor trip at 7 percent power due to a loss of feedwater during main turbine

startup from 2RFO23 on April 20, 2018.

INSPECTION RESULTS

Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head

Penetration #3

Cornerstone

Severity

Cross-Cutting Aspect

Report

Section

Not Applicable

Severity Level IV

NCV 05000286/2018-002-01

Closed

Not Applicable

71111.15

11

Introduction: A self-revealing Severity Level IV NCV of TS 3.4.13.a, Reactor Coolant System

Operational Leakage, was identified when Entergy operated the reactor in Mode 1 with

pressure boundary leakage for a period of time longer than the allowable limiting condition of

operation. Specifically, a leak in the J-weld around RPV head penetration #3 occurred during

the last operating cycle and was not identified until after the reactor was shutdown for a

refueling outage.

Description: On March 28, 2018, during the Unit 2 bare metal exam, Entergy observed white

deposits on the top of the RPV head next to (spare) penetration #3. Entergy concluded the

deposits were dried boric acid. Subsequently, Entergy evaluated indications using ultrasonic

test, eddy current test, and dye penetrant test examination techniques. Entergy concluded

that one indication had propagated into a through-wall pressure boundary leak in RPV head

penetration #3 caused by primary water stress cracking corrosion. Entergy determined that

this leakage was prohibited by TS 3.4.13, Reactor Coolant System Operational Leakage,

which requires that there be no pressure boundary leakage during operating Modes 1 through

4. The condition was reported to the NRC in Event Notification EN-53305 on March 31, 2018,

as required by 10 CFR 50.72(b)(3)(ii)(A), because it constituted a degradation of a principal

safety barrier. Entergy also reported this event in LER 05000247/2018-001-00 (ADAMS

Accession Number ML18149A126) on May 21, 2018, under 10 CFR 50.73(a)(2)(ii)(A) and

10 CFR 50.73(a)(2)(i)(B). These subparts describe an event or condition that resulted in the

nuclear power plant including its principal safety barriers being seriously degraded, and an

operation or a condition which was prohibited by the plants TSs, respectively.

The inspectors reviewed the RCS leakage monitoring records, Entergys event report, the

causal analyses, and previous examination records. These reviews were conducted to

determine whether Entergys monitoring of operational leakage during the previous operating

cycle and prior non-destructive examination activities met regulatory requirements and

applicable standards.

The inspectors independently reviewed the unidentified leak rate data over the previous

operating cycle and the trace boric acid deposits observed during the refueling outage. The

inspectors noted that Entergy completed a bare metal examination of the RPV the previous

refueling outage and determined there were no boric acid deposits in this location. Entergy

conducted isotopic analysis of the boric acid deposits and estimated the leak occurred within

the previous 90 to 180 days. The boron deposits were pure white in appearance indicating

that there were no corrosion products present. Entergy completed leakage monitoring during

the operating cycle in accordance with standards and TS requirements. The inspectors

determined that the applicable limits were met and that the leak was not readily detectable by

these methods because the change in leak rate was too small to be identified. The

inspectors ultimately concluded there was no opportunity for Entergy to observe this condition

visually during the operating cycle because existing boron deposits were not visible when the

plant was operating.

The inspectors observed activities during the refueling outage to determine whether Entergy

completed examinations of the RPV and head penetration #3 in accordance with regulatory

requirements in 10 CFR Part 50, paragraph (g)(6)(ii)(D)), which conditionally endorses

American Society of Mechanical Engineers (ASME), Code Case N-729-4. Specifically, the

inspectors reviewed the results of the nondestructive examinations: time-of-flight ultrasonic

testing, liquid penetrant testing, and eddy current testing. The inspectors also conducted

discussions with the personnel who performed these tests. The inspectors determined

Entergy staff and their contractors implemented these non-destructive examination activities

in accordance with regulatory requirements and their implementing procedures.

12

Corrective Actions: Entergy implemented corrective actions to repair head penetration #3 by

applying a weld onlay prior to returning the RPV to service. This alternative repair method

was the subject of relief request IP2-ISI-RR-06 which was submitted to the NRC and

subsequently approved (ADAMS Accession Number ML18101A032). This weld onlay

reestablished the RCS pressure boundary and addressed the primary water stress cracking

corrosion degradation mechanism. Entergy verified that the pressure boundary had been

restored using both non-destructive testing methods (dye penetration testing and ultrasonic

testing) and an operational pressure test.

Corrective Action References: CR-IP2-2018-01944 and CR-IP2-2018-02122

Performance Assessment:

The inspectors determined that the pressure boundary leakage was not within Entergys

ability to foresee and prevent. Entergy had complied with all ASME code requirements for

inspection of the RPV and there were no indications of RCS pressure boundary leakage

during the operating cycle. As a result, no performance deficiency was identified. Therefore,

this violation will not be considered in the assessment process or the NRCs action matrix.

Enforcement:

Severity: This issue is considered within the Traditional Enforcement process because there

was no performance deficiency associated with the violation of NRC requirements, and the

Reactor Oversight Processs significant determination process does not specifically consider

violations without performance deficiencies in its assessment of licensee performance. In

reaching this decision, the NRC determined that the issue was not within Entergys ability to

foresee and correct, Entergys actions did not contribute to the degraded condition, and the

actions taken were reasonable to identify and address the condition. Furthermore, because

Entergys actions did not contribute to this violation, it will not be considered in the

assessment process or the NRCs action matrix.

Using IMC 0612, Appendix B, Issue Screening, the inspectors determined that the violation

is associated with the Equipment Performance attribute of the Initiating Events cornerstone

and adversely affected the cornerstone objective to limit the likelihood of events that upset

plant stability and challenge critical safety functions during shutdown as well as power

operations. The Barrier Integrity cornerstone was not applicable for this violation because the

violation did not involve pressurized thermal shock.

The inspectors evaluated the issue using IMC 0609.04, Initial Characterization of Finding,

and determined the violation was a LOCA initiator. The inspectors screened the safety

significance to Green using IMC 0609, Appendix A, Exhibit 1, LOCA Initiators, because,

after a reasonable assessment of degradation, it was determined that the violation did not

result in exceeding the RCS leak rate for a small-break LOCA and did not affect any other

systems used to mitigate a LOCA that could result in a total loss of function. Therefore, the

inspectors determined that the issue is of very low safety significance and concluded that the

violation would be best characterized as Severity Level IV under the Traditional Enforcement

process.

Violation: TS 3.4.13(A) requires, in part, that RCS operational leakage shall be limited to no

pressure boundary leakage. If pressure boundary leakage exists, then TS 3.4.13(A) limiting

condition for operation action statement requires leakage to be reduced to within limits or

13

TS 3.4.13(B) requires Unit 2 to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and TS 3.4.13(C) requires Unit 2 to

be in Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Contrary to the above, for a period that began on an unknown date that was likely more than

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> before reactor shutdown on March 19, 2018, RCS pressure boundary leakage

existed, and Entergy did not place Unit 2 in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Mode 4 within

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2

of the Enforcement Policy, because Entergy restored compliance, placed the issue in the

corrective action program, was non-repetitive, was not NRC-identified, was not willful, and did

not meet any minor examples in the Enforcement Manual,Section III, Appendix E.

Observation

71152

Annual Follow-up of

Selected Issues

The inspectors determined several aspects of the fire protection program implementation had

improved, including the following:

(1) Entergy implemented reasonable actions to address each of the fire protection

deficiencies listed in Other Activities - Baseline, Inspection Procedure 71152, Annual

Follow-up of Selected Issues. The most important positive action implemented was the

hiring and training of three new people to perform the fire extinguisher and fire hose

preventive maintenances. These personnel now report directly to the respective

responsible supervisors for portable fire extinguishers and fire hoses.

(2) Entergy initiated an accelerated schedule for replacing 100 percent of the onsite fire

hoses with new hoses or hoses that have been hydrostatically tested this year. The

project is scheduled to be completed by December 2018.

Notwithstanding the significant progress, the inspectors determined several procedure

gaps/deficiencies exist, which could permit more fire hoses to exceed their hydrostatic test

due date prior to their next inspection/replacement. Entergy identified some of these through

the corrective action program but had not resolved them. The inspectors identified additional

deficiencies in eight fire protection hose implementing procedures and discussed the

deficiencies with fire protection engineers. CR-IP3-2018-01465 was initiated to fix the

procedure deficiencies. The inspectors independently inspected fire hoses at 40 onsite

locations (including Units 1, 2, and 3, inside and outside of the radiological controlled areas,

and the intake structure) and found each hose was properly hydrostatically tested within the

required periodicity. Minor hose station deficiencies identified were promptly corrected

(CR-IP3-2018-01388, CR-IP2-2018-03309, CR-IP2-2018-03310, and CR-IP2-2018-03318).

The inspectors determined the procedure deficiencies were minor, based on the fire hose

inspection results and the progress Entergy had made on the project to replace all fire hoses

in 2018.

14

Observation

71152

Semiannual Trend

Review

The inspectors performed a review of Entergys evaluation and corrective actions associated

with the mitigating system performance index (MSPI) for the period from July 2016 through

June 2018. Specifically, the inspectors reviewed the corrective actions to verify they were

addressing the identified conditions adverse to quality. The inspectors assessed Entergys

problem identification threshold, extent of condition reviews, and the prioritization and

timeliness of 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 Entergys corrective action program and 10 CFR Part 50, Appendix B. In

addition, the inspectors interviewed engineering personnel to assess the effectiveness of the

implemented corrective actions.

There were licensee-identified and NRC-identified performance indicator reporting

discrepancies documented during the reviewed timeframe. Of note, there were two

NRC-identified instances in which multiple discrepancies were reported which led to Entergy

reviewing the prior three years of reported data and identifying additional discrepancies. Also,

there was a licensee-identified issue with the consolidated data entry (CDE) system where the

system was reporting estimated values instead of actual values even though Entergy had

reported the actual values appropriately.

Instances of discrepancies were identified by the NRC in November 2016 (CR-IP2-2016-

07044) and September 2017 (CR-IP2-2017-03169, CR-IP2-2017-03196, CR-IP2-2017-03203,

CR-IP3-2017-04222, and CR-IP3-2017-04268). The instance in 2016 involved all five MSPI

systems. This resulted in the MSPI resultant values getting farther from the Green/White

threshold. Entergy reviewed the prior three years, found discrepancies in every quarter, and

submitted corrected data. The instance in 2017 involved all five MSPI systems. This resulted

in a combination of the MSPI resultant values getting closer to and farther from the

Green/White threshold. Entergy reviewed the prior three years, found discrepancies in every

quarter, and submitted corrected data. The scope of the completed corrective actions for the

2016 instance was narrowly focused on post-maintenance testing. The scope of the 2017

review was broader and found errors in the overlapping 2015 and 2016 not related to

post-maintenance testing.

Instances of discrepancies were identified by Entergy in October 2017 (CR-IP2-2017-03819)

and January 2018 (CR-IP2-2018-00144). The instance in 2017 involved the way the CDE

system was reporting data after they had already been submitted. A small portion of the

high-pressure safety injection system was showing as estimated values instead of actual

values. This resulted in a combination of the MSPI resultant values getting closer to and

farther from the green/white threshold. Entergy checked all other parameters and corrected

the condition so that the actual values are being reported. It is worth noting that the CDE

system is maintained by a third party. The instance in 2018 involved the number of hours of

critical operation for both units which is used in the unavailability and unreliability indexes of

MSPI. This resulted in the MSPI resultant values getting closer to the green/white threshold;

however, Entergy corrected the data prior to it being reported.

Entergy did not identify specific trends or causes for the reasons of the discrepancies.

However, Entergy took action in each case to correct the discrepancies after discovery.

Entergy provided coaching to the individuals who compiled and reported the data. Entergy

15

also provided training in June 2018 by a third party expert to all individuals involved in the

MSPI reporting process as a corrective action after the NRC-identified instance in 2017.

It is apparent that there have been programmatic reporting issues with MSPI at the site. None

of the issues listed above led to a change in the performance indicators crossing a threshold.

EXIT MEETINGS AND DEBRIEFS

The inspectors confirmed that proprietary information was controlled to protect from public

disclosure.

On July 18, 2018, the inspectors presented the quarterly resident inspector inspection

results to Mr. Anthony Vitale, Site Vice President, and other members of the Entergy staff.

16

DOCUMENTS REVIEWED

Common Documents Used

Indian Point Units 2 and 3, Updated Final Safety Analysis Report

Indian Point Units 2 and 3, Individual Plant Examination

Indian Point Units 2 and 3, Individual Plant Examination of External Events

Indian Point Units 2 and 3, Technical Specifications and Bases

Indian Point Units 2 and 3, Technical Requirements Manual

Indian Point Units 2 and 3, Control Room Narrative Logs

Indian Point Units 2 and 3, Plan of the Day

71111.01

Procedures

IP-SMM-OP-104, Offsite Power Continuous Monitoring and Notification, Revision 13

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-03277*

71111.04

Procedures

3-COL-RW-002, Service Water System, Revision 51

3-COL-SI-001, Safety Injection System, Revision 44

3-SOP-SI-002, Filling the Refueling Water Storage Tank, Revision 12

3-SOP-SI-003, Recirculation and/or Purification of the Refueling Water Storage Tank,

Revision 26

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-02345* 2018-03138* 2018-03700* 2018-04013* 2018-04014*

Condition Reports (CR-IP3-) (*initiated in response to inspection)

2018-01180* 2018-01768* 2018-01864*

Miscellaneous

CES-2A, Engineering Standards Manual, Seismic Design of Electrical Conduit Supports for

Indian Point Units 2 and 3, Revision 3

71111.05AQ

Procedures

0-PT-M004, Fire Extinguisher Inspection, Revision 9

PFP-215, Unit 2 Primary Auxiliary Building and Fan House, 51-Foot and 68-Foot Elevations,

Revision 8

PFP-252, Unit 2 Cable Spreading Room, Control Building, 33-Foot Elevation, Revision 11

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-03029* 2018-03261*

Condition Reports (CR-IP3-) (*initiated in response to inspection)

2018-01717*

17

Calculations

IP3-CALC-FP-02795, Combustible Loading Calculation for Indian Point 3 Hazards Analysis,

Revision 2

Engineering Evaluations

DC-88-207, Indian Point 2 Fire Barrier Engineering Evaluation, Evaluation No. 1, dated

March 24, 1989

Miscellaneous

PFP-204, General Floor Plan - Primary Auxiliary Building, 15-Foot Elevation, Revision 0

PFP-315, Fuel Storage Bay Area - Fuel Storage Building, Revision 0

PFP-316, General Floor Plan - Fuel Storage Building, Revision 12

PFP-354, Diesel Generator Building - Diesel Generators 31, 32, and 33, 15-Foot Elevation,

Revision 0

71111.07

Procedures

EN-DC-316, Heat Exchanger Performance and Condition Monitoring, Revision 10

EN-MA-118, Foreign Material Exclusion, Revision 12

Condition Reports (CR-IP2-)

2018-03349

2018-03350

Miscellaneous

EC-77962, Evaluate Acceptability of Allowing Temporary Tube Plugs in 22 Component Cooling

Water Heat Exchanger to Next PM Activity, Revision 0

71111.08

Procedures

WDI-STD-1013 (PDI-ISI-254-SE-NB), Remote Inservice Inspection of Rector Vessel Nozzle to

Safe End, Nozzle to Pipe and Safe End to Pipe Welds Using the Nozzle Scanner,

Revision 3

Maintenance Orders/Work Orders

WO 52568740 01

WO 52709908 01

WO 52708009 01

WO 0049815801

Calculation

IP-CALC-04-01420, FCX-0538, Calculation of Effective Degradation Years for the IP2 Reactor

Vessel Head by 2R18

Miscellaneous

ACTS; IP-001-18 Bobbin Coil: 0.720 - Standard Pull Speed

ACTS; IPP 003-18 Bobbin Coil: 0.700

ACTS; IPP-005-18 3-Coil RPC: Mag. Biased 0.720 +PT

ACTS; IPP-006-18 Single Coil RPC: 0.680 U-bend +PT

ACTS; PP-002-18 Bobbin Coil: 0.720 - Reduced Pull Speed

ACTS; PP-004-18 3-Coil RPC: Standard 0.720 +PT

ACTS; PP-007-18 Bobbin Coil: 0.720-80 IPS Pull Speed

ANTS; IPP-B-18 3-Coil RPC - Standard and Mag. Biased (+PT/1 15/S080)

18

ANTS; IPP-C-18 Single Coil +PT U-bend RPC

ANTS; PP-A-18 All Bobbin Coil Probes

ANTS; PP-D-18 Data Quality and General Instructions - All Probe Types

Inspection Data Outlet Nozzle to DM Weld at 158 Degradation Weld No. RPVS-22-1A

Inspection Data Outlet Nozzle to DM Weld at 22 Degradation Weld No. RPVS-24-1A

IPP-R23-OH01-03-01, Ultrasonic Report Data Sheet

IPP-R23-PEN3ET-SCAN2-12940, Eddy Current Report Data Sheet

Linearity Record Pulser Receiver 30028599

Sensitivity Calibration Data Sheet No. SE-AX-DET-01

Sensitivity Calibration Data Sheet No. SE-AX-DET-03

Sensitivity Calibration Data Sheet No. SE-AX-DET-04

Sensitivity Calibration Data Sheet No. SE-CIRC-DET-01

Sensitivity Calibration Data Sheet No. SE-CIRC-DET-02

SG-SGMP-18-2, Indian Point Unit 2 2RFO23 (Spring 2018) Steam Generator Degradation

Assessment, Revision 0

WDI-CAL-002, Pulser/Receiver Linearity Procedure

71111.11

Procedures

O-NF-206, Initial Criticality, Revision 8

O-NF-208, Zero Power Physics Testing, Revision 5

2-POP-1.1, Plant Heatup from Cold Shutdown, Revision 92

2-POP-1.2, Reactor Startup, Revision 61

2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 96

71111.12

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-02246*

71111.13

Procedures

0-VLV-435-CWS, Annual Preventative Maintenance Testing and Replacement of Backflow

Preventers, Revision 2

2-PT-R002B, Recirculation Sump Level, Revision 20

2-SOP-1.1.1, Vacuum Filling and Degassing the Reactor Coolant System, Revision 31

3-PT-M62B, 480 Volt Undervoltage/Degraded Grid Protection System Bus 5A Functional,

Revision 12

3-PT-M62C, 480 Volt Undervoltage/Degraded Grid Protection System Bus 2A and 3A

Functional, Revision 13

3-PT-M62C, 480 Volt Undervoltage/Degraded Grid Protection System Bus 6A Functional,

Revision 13

EN-OP-119, Protected Equipment Postings, Revision 9

EN-WM-104, On Line Risk Assessment, Revision 16

IP-SMM-DC-910, Temporary Equipment, Revision 3

IP-SMM-DC-910, Temporary Equipment, Revision 9

19

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-02240*

Condition Reports (CR-IP3-) (*initiated in response to inspection)

2018-01717*

Drawings

310438, Safety Switch Level Indication and Pump Control Levels Reactor Cavity and Location

Containment Sumps, Revision 10

9321-F-20343, Flow Diagram, City Water, Revision 43

9321-F-27513, Flow Diagram, Auxiliary Coolant System in Primary Auxiliary Building and Fuel

Storage Building, Sheet No. 1, Revision 34

9321-F-33853, Electrical Distribution and Transmission System, Revision 21

Miscellaneous

Unit 3 Equipment Out of Service on Line Risk Assessment for May 1, 2018

Unit 3 Equipment Out of Service on Line Risk Assessment for May 17, 2018

Unit 3 Equipment Out of Service on Line Risk Assessment for June 6, 2018

Unit 3 Equipment Out of Service on Line Risk Assessment for June 17, 2018

71111.15

Procedures

IP-SMM-DC-910, Temporary Equipment, Revision 3

OAP-035, Technical Specifications and Technical Requirements Manual - License Adherence

and Use, Revision 15

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-02541* 2018-02548* 2018-02581* 2018-02588*

Condition Reports (CR-IP3-)

2018-01448* 2018-01530

2018-01552

2018-01570

2018-01878*

71111.18

Condition Reports (CR-IP2-)

2017-05238

Maintenance Orders/Work Orders

WO 00491522

Drawings

9321-2561, Containment Building Residual Heat Exchanger Piping Plans and Sections (A.C.S.),

Revision 23

9321-2720, Flow Diagram Auxiliary Coolant System Updated Final Safety Analysis Report,

Figure No. 9.3-1 (Sheet 2), Revision 93

Miscellaneous

EC-75309, Install Temporary Clamp to Facilitate Weld Repair at Valve A-98, Revision 2

EC-75312, Engineering to Evaluate the Repair of Weld Per Code Case N-666 on a Leaky

Socket Weld at Vent Valve A98 on Residual Heat Removal Line 361, Revision 0

20

EC-76614, Evaluate Using Freeze Seal Isolation on 6-Inch Line 13 to Perform Maintenance on

Valve 769, Revision 0

71111.19

Procedures

2-AOP-RCP-1, Reactor Coolant Pump Malfunction, Revision 16

2-PT-Q013, Inservice Valve Tests, Revision 52

2-PT-R027C, WCPPS Local Leak Rate, Revision 11

3-PC-R06C, Rod Position Indication System Hot Span Confirmation, Revision 12

3-PT-CS030, Atmospheric Steam Dump Valves Stroke Test, Revision 17

3-SOP-RC-001, Full Length Rod Control and Rod Position Indication System Operation,

Revision 29

EN-OP-104, Operability Determination Process, Revision 14

SEP-APJ-006, Indian Point Energy Center Primary Containment Leakage Rate Testing

(Appendix J) Program Plan, Revision 4

SOP-1.3, Reactor Coolant Pump Startup and Shutdown, Revision 47

Condition Reports (CR-IP2-)

2016-01973

2016-02383

2018-02008

2018-02359

Condition Reports (CR-IP3-) (*initiated in response to inspection)

2017-05301

2018-01261

2018-01438

2018-01463*

Maintenance Orders/Work Orders

WO 00441335

WO 00501265

WO 52697548

WO 00363219-01

WO 003623215-01

Miscellaneous

2-PT-Q013, Inservice Valve Tests, Revision 52, dated Mach 19, 2018

2-PT-R027C-DS001, Containment Pressure Relief Valves PCV-1190 and PCV-1191, dated

March 29, 2018

2-PT-R027C-DS001, Containment Pressure Relief Valves PCV-1190 and PCV-1191, dated

March 31, 2016

2-PT-R027C-DS002, Containment Pressure Relief Valves PCV-1191 and PCV-1192, dated

March 29, 2018

71111.20

Procedures

0-NF-206, Initial Criticality, Revision 8

0-NF-208, Zero Power Physics Testing, Revision 5

0-SOP-CB-001, Containment Entry and Egress, Revision 1

2-POP-1.1, Plant Heatup from Cold Shutdown, Revision 92

2-POP-1.2, Reactor Startup, Revision 61

2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 96

2-POP-3.1, Plant Shutdown from 45 Percent Power, Revision 60

2-POP-3.3, Plant Cooldown, Hot to Cold Shutdown, Revision 83

2-POP-3.4, Secondary Plant Shutdown, Revision 10

2-POP-4.1, Operation at Cold Shutdown, Revision 8

21

2-POP-4.2, Operation Below 20 Percent Pressurizer Level with Fuel in the Reactor, Revision 12

2-POP-4.3, Operation without Fuel in the Reactor, Revision 3

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-02205* 2018-02671*

71111.22

Procedures

0-PT-6Y001, B.5.b Diesel Driven Pump Capacity Test, Revision 2

3-PT-M13B1, Reactor Protection Logic Channel Functional Test (Reactor Power Greater Than

35 Percent - P8), Revision 24

3-PT-Q120A, 31 Auxiliary Feedwater Pump, Revision 17

EN-HU-106, Procedure and Work Instruction Use and Adherence, Revision 6

Condition Reports (CR-IP2-)

2018-03672

2018-03675

Condition Reports (CR-IP3-)

2018-00214

Maintenance Orders/Work Orders

WO 52810279

WO 52811023

71124.07

Procedures

0-CY-1930, REMP Air Samples, Revision 2

0-CY-1940, REMP Broad Leaf Vegetation Samples, Revision 0

0-CY-1950, REMP Collection, Preparation, and Analysis of Water Samples, Revision 0

0-EV-DD-101, Meteorological Systems Manual, Revision 1

EN-CY-111, Radiological Groundwater Protection Program, Revision 8

IP-SMM-CY-110, Radiological Groundwater Monitoring Program, Revision 9

Condition Reports (CR-IP2-)

2017-00635

2018-02134

2018-02272

2018-03578

2018-03624

Condition Reports (CR-IP3-)

2017-03403

Miscellaneous

Environmental Dosimetry Company Annual Quality Assurance Status Report, dated March 7,

2018

Indian Point Energy Center Offsite Dose Calculation Manual, Revision 4

NAQA-17-0038, Audit Report 17-007 for the Audit of Environmental Dosimetry Company, dated

September 1, 2017

71151

Procedures

NEI 99-02, Regulatory Assessment Performance Indicator Guidelines, Revision 7

22

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2018-01129

2018-02739* 2018-03931

71152

Procedures

0-OSP-FP-001, Hydrostatic Testing of Firehose, Revision 0

0-PT-M001, Fire Brigade Equipment Inventory and Inspection, Revision 18

0-PT-M004, Fire Extinguisher Inspection, Revision 9

2-PT-M112, Fire Protection Flow Path and Hose Station Equipment Verification, Revision 6

AP-64.1, Fire Protection/Appendix R Systems and Components Governed by Technical

Requirements Manual and Technical Specifications, Revision 5

SAO-703, Fire Protection Impairment Criteria and Surveillance, Revision 35

Condition Reports (CR-IP2-) (*initiated in response to inspection)

2016-00509

2017-04090

2018-03262* 2018-03306* 2018-03309* 2018-03310*

2018-03318*

Condition Reports (CR-IP3-) (*initiated in response to inspection)

2018-01112

2018-01388* 2018-01465*

Maintenance Orders/Work Orders

WO 00379953

WO 00383371

WO 00417368

WO 52620568

WO 52686482

WO 52692709

WO 52709916

WO 52723346

WO 52810046

Miscellaneous

EC-76593, Indian Point Energy Center Pre-Fire Plan Updates, Revision 0

Indian Point Unit 3 Technical Requirements Manual, Section 3.7.A.5, Fire Hose Stations,

Revision 3

Indian Point Unit 3 Technical Requirements Manual, Section 3.7.A.6, Yard Fire Hydrants and

Hydrant Hose Houses, Revision 2

NFPA-1962, Standard for the Care, Use, Inspection, Service Testing, and Replacement of Fire

Hose, Couplings, Nozzles, and Fire Hose Appliances, 2018 Edition

71153

Procedures

EN-FAP-MP-009, Enhanced Procurement Process for Non-Safety Related Critical Spares,

Revision 3

EN-MP-100, Critical Procurements, Revision 13

Condition Reports (CR-IP3-)

2017-05133

2018-00539

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