ML16243A245
| ML16243A245 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 08/30/2016 |
| From: | Glenn Dentel Reactor Projects Branch 2 |
| To: | Vitale A Entergy Nuclear Operations |
| References | |
| IR 2016002 | |
| Download: ML16243A245 (54) | |
See also: IR 05000247/2016002
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
2100 RENAISSANCE BLVD.
KING OF PRUSSIA, PA 19406-2713
August 30, 2016
Mr. Anthony J. Vitale
Site Vice President
Entergy Nuclear Operations, Inc.
Indian Point Energy Center
450 Broadway, GSB
P.O. Box 249
Buchanan, NY 10511-0249
SUBJECT: INDIAN POINT NUCLEAR GENERATING - INTEGRATED INSPECTION
REPORT 05000247/2016002 AND 05000286/2016002
Dear Mr. Vitale:
On June 30, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Indian Point Nuclear Generating (Indian Point), Units 2 and 3. The enclosed inspection
report documents the inspection results, which were discussed on August 4, 2016, with Larry
Coyle and other members of your staff. Based on additional information provided, the
inspectors conducted an updated exit meeting on August 30, 2016 with John Kirkpatrick, Plant
Operations General Manager and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
This report documents three NRC-identified findings of very low safety significance (Green).
These findings involved violations of NRC requirements. However, because of the very low
safety significance, and because they are entered into your corrective action program, the NRC
is treating these findings as non-cited violations, consistent with Section 2.3.2.a of the NRC
Enforcement Policy. If you contest any non-cited violation in this report, you should provide a
response within 30 days of the date of this inspection report, with the basis for your denial, to
the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC
20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of
Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
NRC Senior Resident Inspector at Indian Point. In addition, if you disagree with the
cross-cutting aspect assigned to any finding in this report, you should provide a response within
30 days of the date of this inspection report, with the basis for your disagreement, to the
Regional Administrator, Region I, and the NRC Senior Resident Inspector at Indian Point.
A. Vitale -2-
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390 of the NRCs
Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be
available electronically for public inspection in the NRCs Public Document Room or from the
Publicly Available Records component of the NRCs Agencywide Documents Access and
Management System (ADAMS). ADAMS is accessible from the NRC website at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Glenn T. Dentel, Chief
Reactor Projects Branch 2
Division of Reactor Projects
Docket Nos. 50-247 and 50-286
License Nos. DPR-26 and DPR-64
Enclosure:
Inspection Report 05000247/2016002 and 05000286/2016002
w/Attachment: Supplementary Information
cc w/encl: Distribution via ListServ
Non-Sensitive Publicly Available
SUNSI Review
Sensitive Non-Publicly Available
OFFICE RI/DRP RI/DRP RI/DRS RI/DRP RI/DRP
BHaagensen/bh
NAME NFloyd/nf MGray/mg GDentel/gtd MScott/dlp for
DATE 8/29/16 8/24/16 8/30/16 8/30/16 8/30/16
1
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket Nos. 50-247 and 50-286
License Nos. DPR-26 and DPR-64
Report Nos. 05000247/2016002 and 05000286/2016002
Licensee: Entergy Nuclear Northeast (Entergy)
Facility: Indian Point Nuclear Generating Units 2 and 3
Location: 450 Broadway, GSB
Buchanan, NY 10511-0249
Dates: April 1, 2016, through June 30, 2016
Inspectors: B. Haagensen, Senior Resident Inspector
G. Newman, Resident Inspector
S. Rich, Resident Inspector
S. Galbreath, Reactor Inspector
J. Furia, Senior Health Physicist
N. Floyd, Senior Project Engineer
K. Mangan, Senior Reactor Inspector
J. Poehler, Senior Materials Engineer
Approved By: Glenn T. Dentel, Chief
Reactor Projects Branch 2
Division of Reactor Projects
Enclosure
2
TABLE OF CONTENTS
SUMMARY .................................................................................................................................... 3
REPORT DETAILS ....................................................................................................................... 5
1. REACTOR SAFETY .............................................................................................................. 5
1R04 Equipment Alignment .................................................................................................. 5
1R05 Fire Protection ............................................................................................................. 6
1R07 Heat Sink Performance ............................................................................................... 7
1R08 Inservice Inspection Activities ..................................................................................... 7
1R11 Licensed Operator Requalification Program ............................................................... 8
1R12 Maintenance Effectiveness ....................................................................................... 10
1R13 Maintenance Risk Assessments and Emergent Work Control .................................. 13
1R15 Operability Determinations and Functionality Assessments ..................................... 14
1R18 Plant Modifications .................................................................................................... 19
1R19 Post-Maintenance Testing ........................................................................................ 20
1R20 Refueling and Other Outage Activities ...................................................................... 21
1R22 Surveillance Testing .................................................................................................. 24
1EP6 Drill Evaluation .......................................................................................................... 25
2. RADIATION SAFETY .......................................................................................................... 25
2RS1 Radiological Hazard Assessment and Exposure Controls ........................................ 25
2RS2 Occupational As Low As Is Reasonably Achievable (ALARA) Planning
and Controls .............................................................................................................. 26
2RS7 Radiological Environmental Monitoring Program (REMP) ........................................ 26
4. OTHER ACTIVITIES ............................................................................................................ 27
4OA1 Performance Indicator Verification ............................................................................ 27
4OA2 Problem Identification and Resolution ....................................................................... 28
4OA3 Follow Up of Events and Notices of Enforcement Discretion .................................... 34
4OA5 Other Activities .......................................................................................................... 37
4OA6 Meetings, Including Exit ............................................................................................ 39
SUPPLEMENTARY INFORMATION ........................................................................................ A-1
KEY POINTS OF CONTACT .................................................................................................... A-1
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED ..................................... A-2
LIST OF DOCUMENTS REVIEWED ........................................................................................ A-3
LIST OF ACRONYMS ............................................................................................................. A-12
3
SUMMARY
Inspection Report 05000247/2016002 and 05000286/2016002; 04/01/2016 - 06/30/2016; Indian
Point Nuclear Generating (Indian Point), Units 2 and 3; Operability Determinations and
Functionality Assessments, Refueling and Other Outage Activities, and Follow Up of Events and
Notices of Enforcement Discretion.
This report covered a three-month period of inspection by resident inspectors and announced
inspections performed by regional inspectors. The inspectors identified three findings of very
low safety significance (Green), which were non-cited violations (NCVs). The significance of
most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red)
and determined using Inspection Manual Chapter (IMC) 0609, Significance Determination
Process, dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310,
Aspects within the Cross-Cutting Areas, dated December 4, 2014. All violations of
U.S. Nuclear Regulatory Commission (NRC) requirements are dispositioned in accordance with
the NRCs Enforcement Policy, dated February 4, 2015. The NRCs program for overseeing the
safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor
Oversight Process, Revision 6.
Cornerstone: Mitigating Systems
Green. The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V,
"Instructions, Procedures, and Drawings," because Entergy did not adequately accomplish
the actions prescribed by procedure EN-OP-104, Operability Determination Process, for a
degraded condition associated with the Unit 3 baffle-former bolts. Specifically, Entergy
incorrectly concluded that no degraded or non-conforming condition existed related to the
Unit 3 baffle-former bolts and exited the operability determination procedure. Entergy
subsequently performed the remaining steps in the procedure and provided appropriate
justification for their plans to examine the baffle-former bolts at the next Unit 3 refueling
outage (RFO). Entergys immediate corrective actions included entering the issue into its
corrective action program (CAP) as CR-IP3-2016-01961 and documenting an operability
evaluation to support the basis for operability of the baffle-former bolts and the emergency
core cooling system (ECCS).
This performance deficiency is more than minor because it was associated with the
equipment performance attribute of the Mitigating Systems cornerstone and affected the
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences (i.e., core damage). In
accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of
IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power,
issued June 19, 2012, the inspectors screened the finding for safety significance and
determined it to be of very low safety significance (Green), because the finding did not
represent an actual loss of system or function. After inspector questioning, Entergy
performed an operability evaluation, which provided sufficient bases to conclude the Unit 3
baffle assembly would support ECCS operability. This finding is related to the cross-cutting
aspect of Problem Identification and Resolution, Operating Experience, because Entergy did
not effectively evaluate relevant internal and external operating experience. Specifically,
Entergy did not adequately evaluate the impact of degraded baffle bolts at Unit 3 when
relevant operating experience was identified at Unit 2. [P.5 - Operating Experience]
(Section 1R15)
4
Green. The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1,
Procedures, for Entergys failure to implement procedure OAP-007, Containment Entry
and Egress. Specifically, workers transiting the inner and outer crane wall sections of
containment failed to maintain at least one (of two) flow channeling gate closed to ensure
availability of the containment sumps to provide suction for the ECCS. Entergy immediately
coached the gate monitor and restored the gates to an acceptable position. Entergy
generated CR-IP2-2016-04036 to address this issue.
This performance deficiency is more than minor because it was associated with the
configuration control (shutdown equipment lineup) attribute of the Mitigating Systems
cornerstone and affected the cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable consequences
(i.e., core damage). A detailed risk assessment was conducted and determined that the
change in core damage frequency was determined to be 7E-9, therefore, this issue
represents a Green finding. This finding had a cross-cutting aspect in the area of Human
Performance, Avoid Complacency, because Entergy did not consider potential undesired
consequences of actions before performing work and implement appropriate error-reduction
tools. Specifically, the work crew did not understand the requirements and potential
consequences prior to commencing work and the gate monitor did not enforce these
requirements to maintain at least one gate locked or pinned closed as required by OAP-007.
[H.12 - Avoid Complacency] (Section 1R20)
Cornerstone: Barrier Integrity
Green. The inspectors identified a Green NCV of 10 CFR 50.65(b)(1) for Entergys failure to
include a function of a safety-related system within the scope of the maintenance rule
program. Specifically, Entergy failed to include the feedwater isolation function performed
by the main boiler feedwater pumps (MBFPs) discharge valves, MBFPs, and feedwater
regulating valves, which are required to remain functional during and following a design
basis event to mitigate the consequence of the accident within the scope of the maintenance
rule monitoring program. Entergy initiated corrective actions to include the feedwater
isolation function performed by the MBFP discharge valves, MBFPs, and feedwater
regulating valves within the maintenance rule monitoring program. Entergy entered this
issue into the CAP as CR-IP2-2016-03963.
This performance deficiency is more than minor because it was associated with barrier
performance attribute of the Barrier Integrity cornerstone and adversely affected the
cornerstone objective to provide reasonable assurance that physical design barriers protect
the public from radionuclide releases caused by accidents or events. Specifically, the failure
to properly scope the feedwater isolation function prevented Entergy from identifying that
equipment reliability was no longer effectively controlled through preventive maintenance.
In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC
0609, Appendix A, The Significance Determination Process for Findings At-Power, issued
June 19, 2012, the inspectors determined that the finding was of very low safety significance
(Green) because the finding did not represent an actual open pathway in the physical
integrity of reactor containment, containment isolation system, and heat removal
components. This finding does not have a cross-cutting aspect since the failure to scope
this equipment into the maintenance rule program was not recognized when Entergy
combined the maintenance rule basis documents for Units 2 and 3 in 2012 and, as a result,
is not indicative of current licensee performance. (Section 4OA3)
5
REPORT DETAILS
Summary of Plant Status
Unit 2 began the inspection period during RFO 2R22 which lasted 102 days. Upon completion
of the outage, the operators restarted Unit 2 on June 14, 2016, and increased power slowly to
93 percent for fuel preconditioning. On June 23, 2016, the operators shutdown the reactor to
repair a service water leak on the 21 component cooling water (CCW) heat exchanger (Hx) inlet
line and replace switchyard breaker 9. Unit 2 returned to 100 percent power on June 29, 2016.
Unit 2 remained at or near 100 percent power for the remainder of the inspection period.
Unit 3 began the inspection period at 100 percent power. On April 26, 2016, a failed controller
caused both heater drain pumps to trip; and the operators reduced power rapidly, stabilizing the
unit at 48 percent power. Operators returned Unit 3 to 100 percent power on April 27, 2016,
and remained at or near 100 percent power for the remainder of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R04 Equipment Alignment
Partial System Walkdowns (71111.04Q - 5 samples)
a. Inspection Scope
The inspectors performed partial walkdowns of the following systems:
Unit 2
Spent fuel pool cooling system following core offload on May 19, 2016
Shutdown cooling system following core reload on June 6, 2016
CCW system following maintenance on June 28, 2016
Unit 3
32 emergency diesel generator (EDG) following maintenance on May 9, 2016 (this
sample was part of an in-depth review of the EDG system)
Residual heat removal pumps following CCW system testing on May 20, 2016
The inspectors selected these systems based on their risk-significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors reviewed
applicable operating procedures, system diagrams, the updated final safety analysis
report (UFSAR), TSs, work orders (WOs), condition reports (CRs), and the impact of
ongoing work activities on redundant trains of equipment in order to identify conditions
that could have impacted system performance of their intended safety functions. The
inspectors also performed field walkdowns of accessible portions of the systems to verify
system components and support equipment were aligned correctly and were operable.
The inspectors examined the material condition of the components and observed
operating parameters of equipment to verify that there were no deficiencies. The
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inspectors also reviewed whether Entergy had properly identified equipment issues and
entered them into the CAP for resolution with the appropriate significance
characterization. Documents reviewed for each section of this inspection report are
listed in the Attachment.
b. Findings
No findings were identified.
1R05 Fire Protection
Resident Inspector Quarterly Walkdowns (71111.05Q - 6 samples)
a. Inspection Scope
The inspectors conducted tours of the areas listed below to assess the material
condition and operational status of fire protection features. The inspectors verified that
Entergy controlled combustible materials and ignition sources in accordance with
administrative procedures. The inspectors verified that fire protection and suppression
equipment were available for use as specified in the area pre-fire plan (PFP) and
passive fire barriers were maintained in good material condition. The inspectors also
verified that station personnel implemented compensatory measures for out-of-service
(OOS), degraded, or inoperable fire protection equipment, as applicable, in accordance
with procedures.
Unit 2
Containment, 95-foot elevation, during baffle bolt repair activities with hot work in
progress (PFP-203 was reviewed) on June 2, 2016
Residual heat removal pump rooms in primary auxiliary building (PAB), 15-foot
elevation (PFP-204 was reviewed), on June 6, 2016
CCW pump room, 68-foot elevation (PFP-209 was reviewed), on June 25, 2016
PAB, 80-foot elevation, CCW heat exchanger area with hot work in progress
(PFP-211 was reviewed) on June 25, 2016
Unit 3
32 EDG room, 10-foot elevation (PFP-354 was reviewed), on May 9, 2016
480V switchgear room, 15-foot elevation (PFP-351 was reviewed), on June 30, 2016
b. Findings
No findings were identified.
7
1R07 Heat Sink Performance (71111.07A - 1 sample)
a. Inspection Scope
The inspectors reviewed the 32 EDG jacket water and lube oil heat exchanger to
determine its readiness and availability to perform its safety functions. The inspectors
reviewed the design basis for the component and verified Entergys commitments to
NRC Generic Letter 89-13, Service Water System Requirements Affecting
Safety-Related Equipment. The inspectors observed the annual cleaning and
inspection of the heat exchangers and reviewed the results of previous inspections of
the Unit 3 EDG heat exchangers. The inspectors discussed the results of the most
recent inspection with engineering staff. The inspectors verified that Entergy initiated
appropriate corrective actions for identified deficiencies. The inspectors also verified
that the number of tubes plugged within the heat exchanger did not exceed the
maximum amount allowed.
b. Findings
No findings were identified.
1R08 Inservice Inspection Activities (71111.08P - 1 sample)
a. Inspection Scope
Inspectors from the NRC Region I Office, specializing in materials and inservice
examination activities, observed portions of Entergys activities involving baffle-former
bolt examinations and replacements during Unit 2 RFO 2R22. The inspectors reviewed
work documentation and examination procedures and results, and discussed these
activities with Entergy. The inspectors were on-site from April 27 to April 28, 2016, and
on May 23, 2016. The inspectors verified that Entergy completed baffle-former bolt
examinations in accordance with their approved procedures which implemented
activities described in the Materials Reliability Program (MRP)-227-A, Pressurized
Water Reactor Internals Inspection and Evaluation Guidelines, as they relate to this
component. Specifically, the inspectors reviewed the results of the visual and volumetric
examinations of the baffle-former bolts, including capabilities, limitations, and
acceptance criteria that were performed during the current RFO.
Non-Destructive Examination Activities
The inspectors reviewed the ultrasonic testing (UT) procedure used for the examination
of the Unit 2 baffle-former bolts to verify the procedure was in accordance with the
applicable guidance in MRP-227-A and MRP-228. The inspectors reviewed the UT data
records and the detailed UT channel analysis for a sample of baffle-former bolts to verify
the examinations and evaluations were performed in accordance with approved
procedures and applicable guidance. The inspectors reviewed video recordings of the
visual examinations of the baffle-former bolts during the current RFO. The inspectors
also reviewed recorded video of visual examinations performed in 2006 at Unit 2,
completed as part of the existing inservice inspection program for the 10-year reactor
vessel examinations, to independently assess the past conditions of the baffle-former
bolts and assembly.
8
The inspectors reviewed certifications of the UT technicians performing the ultrasonic
examinations to verify the examinations were performed by qualified individuals and to
verify the results were reviewed and evaluated by certified level III non-destructive
examination personnel.
Baffle-Former Bolt Replacement Activities
The inspectors reviewed the baffle-former bolt replacement activities performed as part
of a corrective action to resolve the degraded condition identified at Unit 2. The
inspectors observed a sample of in-process bolt removal activities, which included lock
bar milling and bolt hole machining. The inspectors reviewed the documentation for
in-process and completed bolt installation activities and verified that loose parts
generated as part of the bolt replacements were properly tracked. The inspectors
verified that bolt replacement activities were performed in accordance with approved
procedures. The inspectors also reviewed the Engineering Change (EC) package
associated with the new baffle-former bolt design. This review is documented in
Section 1R18 of this report. After completion of the bolt replacement activities, the
inspectors reviewed the video of the final visual examination of the baffle assembly to
verify that the baffle-former bolt work was accomplished as planned and that there were
no visual indications of deficiencies.
b. Findings
No findings were identified.
Update to URI 05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies
This inspection was conducted to follow-up on NRC Unresolved Item (URI)05000247/2016001-01, Baffle-Former Bolts with Identified Anomalies, to determine
whether there was a performance deficiency associated with the degraded baffle-former
bolt condition discovered at Unit 2. The inspectors plan to review additional technical
information from Entergy as it becomes available, including any revisions to the root
cause evaluation. The URI remains open until review of this additional information is
completed. (URI 05000247/2016001-01, Baffle-Former Bolts with Identified
Anomalies)
1R11 Licensed Operator Requalification Program (71111.11Q - 5 samples)
Unit 2
.1 Quarterly Review of Unit 2 Licensed Operator Requalification Testing and Training
(71111.11Q - 1 sample)
a. Inspection Scope
The inspectors observed Unit 2 licensed operator simulator training on May 24, 2016,
which included reactor coolant pump seal failure with loss of normal heat sink requiring
implementation of feed and bleed cooling. The inspectors evaluated operator
performance during the simulated event and verified completion of risk significant
operator actions, including the use of abnormal and emergency operating procedures.
The inspectors assessed the clarity and effectiveness of communications,
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implementation of actions in response to alarms and degrading plant conditions, and the
oversight and direction provided by the control room supervisor. The inspectors verified
the accuracy and timeliness of the emergency classification made by the shift manager
and the TS action statements entered by the shift technical advisor. Additionally, the
inspectors assessed the ability of the crew and training staff to identify and document
crew performance problems.
b. Findings
No findings were identified.
.2 Quarterly Review of Unit 3 Licensed Operator Requalification Testing and Training
(71111.11Q - 1 sample)
a. Inspection Scope
The inspectors observed a Unit 3 licensed operator simulator requalification training
evaluated scenario on May 24, 2016, which included failure of a pressurizer pressure
instrument, charging pump trip, loss of 480V safety bus 5A, a small break loss-of-coolant
accident (LOCA), and entry into FR-C.2 core cooling. The inspectors evaluated operator
performance during the simulated event and verified completion of risk significant
operator actions, including the use of abnormal and emergency operating procedures.
The inspectors assessed the clarity and effectiveness of communications,
implementation of actions in response to alarms and degrading plant conditions, and the
oversight and direction provided by the control room supervisor. The inspectors verified
the accuracy and timeliness of the emergency classification made by the shift manager
and the TS action statements entered by the shift technical advisor. Additionally, the
inspectors assessed the ability of the crew and training staff to identify and document
crew performance problems.
b. Findings
No findings were identified.
.3 Quarterly Review of Licensed Operator Performance (71111.11Q - 3 samples)
a. Inspection Scope
The inspectors conducted a focused observation of operator performance in the main
control room. The inspectors observed pre-job briefings and control room
communications to verify they met the criteria specified in Entergys administrative
procedure EN-OP-115, Conduct of Operations. Additionally, the inspectors observed
restoration activities to verify that procedure use, crew communications, and
coordination of activities between work groups similarly met established expectations
and standards.
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Unit 2
Plant startup from RFO 2R22 on June 16, 2016 including response to a turbine trip
without a reactor trip and the subsequent turbine-generator synchronization and
transfer of plant electrical loads from offsite power to the unit auxiliary transformer.
Reactor startup and grid synchronization conducted on June 27, 2016.
Unit 3
Operator response to the feedwater transient which occurred on April 26, 2016
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12Q - 4 samples)
.1 Routine Maintenance Effectiveness
a. Inspection Scope
The inspectors reviewed the samples listed below to assess the effectiveness of
maintenance activities on SSCs performance and reliability. The inspectors reviewed
system health reports, CAP documents, maintenance WOs, and maintenance rule basis
documents to ensure that Entergy was identifying and properly evaluating performance
problems within the scope of the maintenance rule. For each SSC sample selected, the
inspectors verified that the SSC was properly scoped into the maintenance rule in
accordance with 10 CFR 50.65 and verified that the (a)(2) performance criteria
established by Entergy was reasonable. As applicable, for SSCs classified as (a)(1), the
inspectors assessed the adequacy of goals and corrective actions to return these SSCs
to (a)(2). Additionally, the inspectors ensured that Entergy was identifying and
addressing common cause failures that occurred within and across maintenance rule
system boundaries.
Unit 2 EDGs
Unit 3 EDGs (this sample was part of an in-depth review of the EDG system)
Units 2 and 3 CVCS
b. Findings
No findings were identified.
URI Opened, CVCS Goal Monitoring Under the Maintenance Rule
Introduction
The inspectors identified issues of potential concern with Entergys application of
10 CFR 50.65(a)(1), Requirements for Monitoring the Effectiveness of Maintenance at
Nuclear Plants, (the maintenance rule) in regards to the reliability of the Unit 2 CVCS
system. These concerns included the establishment of appropriate (a)(1) goals and
11
whether appropriate justification was established that the corrective actions to address
identified maintenance weaknesses were effective prior to removal from (a)(1) status.
Specifically, Entergy may have established restrictive goals without defensible
justification and may not have demonstrated their chosen goal before ending the goal
monitoring interval.
Description
The maintenance rule requires that licensees shall monitor the performance or condition
of structures, systems, or components, against licensee-established goals, in a manner
sufficient to provide reasonable assurance that these structures, systems, and
components are capable of fulfilling their intended functions. These goals shall be
established commensurate with safety and, where practical, take into account
industrywide operating experience. When the performance or condition of a structure,
system, or component does not meet established goals, appropriate corrective action
shall be taken. EN-DC-206, Maintenance Rule (a)(1) Process, provides the
requirements and processes for managing SSCs for which (a)(2) monitoring has not
demonstrated effective maintenance. EN-DC-206 specifies that (a)(1) action plans
should not be closed until effectiveness of all corrective actions has been demonstrated
by meeting performance goals through the monitoring period (or by other means
specified in the action plan).
Since 2013, there have been several repeat functional failures of equipment in the
CVCS resulting in a failure to meet the performance criterion for reliability. These
failures included:
A failure of the 23 charging pump on August 6, 2013, after the internal oil pump
discharge tubing broke causing the pump to trip on low oil pressure and a loss of
charging. The 21 charging pump had tripped for the same reason in 2010.
A failure of the 22 charging pump on January 14, 2014, due to cracked internal
check valves caused by an inadequate fill-and-vent that left air in the pump following
maintenance. The 21 charging pump had failed due to the same cause in 2013.
A failure of the Unit 2 valve FCV-110A, boric acid flow control valve, to fully open on
January 5, 2015. The valve had insufficient insulation; and as a result, boron
crystalized above the valve plug and blocked its movement. The Unit 3 FCV-110A
had failed in the same way in 2011, with earlier failures of other valves for the same
cause going back to 1997.
In each case, the CVCS for Unit 2 was already (a)(1), so Entergy either updated the
existing (a)(1) action plan or created another one to operate in parallel with the existing
one. Upon reviewing the associated (a)(1) action plans, the inspectors noted that in
each example Entergys goals may not have been in accordance with EN-DC-206(a)(1)
Process. It specifies that monitoring intervals should be at least six months for normally
operating SSCs, at least three surveillances for SSCs monitored by surveillance and
long enough to detect recurrence of the applicable failure mechanism. It also states that
performance goals that provide reasonable assurance that the SSC is capable of
performing its intended functions should be monitored throughout the time the SSC is
classified (a)(1). EN-DC-206 defines an SSC as any discreet component grouping that
has caused a monitoring failure, including any applicable extent of condition. In the
examples provided, NRC inspectors challenged whether Entergy either chose a shorter
12
monitoring interval or a goal that did not include the applicable extent of condition.
Specifically:
The (a)(1) action plan for the broken oil tubing had a goal of no noticeable decrease
in 23 charging pumps running oil pressure for the next three quarterly surveillances.
The chosen monitoring interval met the procedural expectation, but Entergy limited
the monitoring to the 23 charging pump without written justification, when the 21
charging pump had failed previously for the same reason and the other pumps were
susceptible to the same failure mechanism. During the monitoring interval, the 21
charging pump experienced low oil pressure. When Entergy performed repairs on
the 21 charging pump for an unrelated issue, they discovered that the oil tubing had
failed in the same way the 23 charging pump oil tubing had failed, although it had not
yet caused a pump trip.
The (a)(1) action plan for the cracked check valves had a goal of no check valve
failure for six months for the next charging pump that underwent maintenance. This
happened to be the 22 charging pump. Entergy chose a six-month monitoring
interval, even though only one of the three charging pumps is in service at any given
time, and the 22 charging pump only ran for four out of the six months it was
monitored. Additionally, the action plan did not justify why a single successful fill-
and-vent demonstrated adequate corrective actions. On November 19, 2014, during
the six month monitoring interval, the 21 charging pump underwent maintenance
requiring a fill-and-vent, and experienced check valve failure two weeks later on
December 4. Entergy documented this as a maintenance rule functional failure, and
discussed the possibility that it could be due to an inadequate fill-and-vent, but did
not change the (a)(1) action plan.
The (a)(1) action plan for FCV-110A specified a monitoring interval of six months to
include the winter because the previous valve failures had all occurred during the
winter months. However, the actual monitoring interval documented in the corrective
action was from April to October 2015, and therefore did not cover the winter months
as intended. In January 2016, Entergy performed maintenance on valve CH-297 on
Unit 3, which is a heat-traced boric acid valve, and did not properly restore the
insulation. The valve function was not impacted because it does not often contain
high concentrations of boric acid.
The (a)(1) action plans described above were all reviewed and approved by the
maintenance rule expert panel.
Further information regarding the performance of these SSCs is required to determine
whether these issues of concern represent performance deficiencies and whether they
are more than minor. (URI 05000247/2016002-01, CVCS Goal Monitoring Under the
Maintenance Rule)
.2 Quality Control
a. Inspection Scope
The inspectors reviewed the weld repair performed on the 21 CCW heat exchanger
service water inlet nozzle for Unit 2 to verify Entergy was properly applying quality
controls specified in their quality assurance program. The inspectors reviewed CAP
documents, maintenance WOs, ECs, and engineering procedures associated with the
weld repair. The inspectors verified Entergy specified quality control hold points in
13
accordance with their procedures, properly controlled the quality of materials used
during the repair, and adequately justified deviations from the existing design.
Additionally, the inspectors reviewed the welding procedure specification qualification by
the vendor to ensure it was in accordance with American Society of Mechanical
Engineers code.
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 7 samples)
a. Inspection Scope
The inspectors reviewed station evaluation and management of plant risk for the
maintenance and emergent work activities listed below to verify that Entergy performed
the appropriate risk assessments prior to removing equipment for work. The inspectors
selected these activities based on potential risk significance relative to the reactor safety
cornerstones. As applicable for each activity, the inspectors verified that Entergy
performed risk assessments as required by 10 CFR 50.65(a)(4) and that the
assessments were accurate and complete. When Entergy performed emergent work,
the inspectors verified that operations personnel promptly assessed and managed plant
risk. The inspectors reviewed the scope of maintenance work and discussed the results
of the assessment with the stations probabilistic risk analyst to verify plant conditions
were consistent with the risk assessment. The inspectors also reviewed the TS
requirements and inspected portions of redundant safety systems, when applicable, to
verify risk analysis assumptions were valid and applicable requirements were met.
Unit 2
Temporary loss of spent fuel pool cooling due to 345-kilovolt disturbance on
April 3, 2016
Equipment hatch closure plug seal demonstration for outage risk on April 5, 2016
Reduced inventory operations during vessel reassembly on June 7, 2016
21 CCW heat exchanger OOS during mode 4 on June 25, 2016
Unit 3
32 EDG OOS while Bus Tie BT 4-5 was OOS on May 4, 2016 (this sample was part
of an in-depth review of the EDG system)
33 EDG OOS while Bus Tie BT 4-5 was OOS on June 2, 2016
31 EDG OOS while Bus Tie BT 4-5 was OOS on June 21, 2016
b. Findings
No findings were identified.
14
1R15 Operability Determinations and Functionality Assessments (71111.15 - 7 samples)
a. Inspection Scope
The inspectors reviewed operability determinations for the following degraded or
non-conforming conditions:
Unit 2
23 EDG failure to run on March 7, 2016, and subsequent failure to pass the
surveillance test on March 10, 2016, as identified in CR-IP2-2016-01260
Operability determination for N33 gamma metrics wide range nuclear instrument
channel in CR-IP2-2016-03660 on June 13, 2016
Pressurizer level transmitter LT-461 reads high in CR-IP-2016-3806 on June 14,
2016
Through-wall leak in line 411, service water inlet to the 21 CCW heat exchanger, on
June 15, 2016
Unit 3
Immediate operability determination of the degraded condition of the baffle-former
bolts identified from Unit 2 operating experience in CR-IP3-2016-01035 on April 1,
2016
Anomalies noted during digital metal impact monitoring system self-test in
CR-IP3-2015-03468 on April 1, 2016
Prompt operability determination of the degraded condition of the baffle-former bolts
identified from Unit 2 operating experience in CR-IP2-2016-03660 on June 30, 2016
The inspectors selected these issues based on the risk significance of the associated
components and systems. The inspectors evaluated the technical adequacy of the
operability determinations to assess whether TS operability was properly justified and
the subject component or system remained available such that no unrecognized
increase in risk occurred. The inspectors compared the operability and design criteria in
the appropriate sections of the TSs and UFSAR to Entergys evaluations to determine
whether the components or systems were operable.
The inspectors confirmed, where appropriate, compliance with bounding limitations
associated with the evaluations. Where compensatory measures were required to
maintain operability, the inspectors determined whether the measures in place would
function as intended and were properly controlled by Entergy. The inspectors
determined, where appropriate, compliance with bounding limitations associated with the
evaluations.
b. Findings
Introduction. The inspectors identified a Green NCV of 10 CFR 50, Appendix B,
Criterion V, "Instructions, Procedures, and Drawings," because Entergy did not
adequately accomplish the actions prescribed by procedure EN-OP-104 for a degraded
condition associated with the Unit 3 baffle-former bolts. Specifically, Entergy incorrectly
concluded that no degraded or non-conforming condition existed related to the Unit 3
15
baffle-former bolts and exited the operability determination procedure. Entergy
subsequently performed the remaining steps in the procedure and provided appropriate
justification for their plans to examine the baffle-former bolts at the next Unit 3 RFO.
Description. On March 29, 2016, Entergy identified baffle-former (baffle) bolt
degradation at Indian Point Unit 2 that was determined to be unanalyzed because it did
not meet the minimum acceptable bolt pattern analysis developed to support plant
startup. Entergy staff identified a total of 227 baffle bolts out of a population of 832 that
were potentially degraded (182 bolts had UT indications; 31 had visual indications of
failure; and 14 were inaccessible for testing and conservatively assumed to be
degraded). Entergy staff entered this problem into the CAP as CR-IP2-2016-02081,
performed a root cause evaluation, and replaced the degraded bolts on Unit 2. Due to
the number of baffle bolt indications discovered on Unit 2, Entergy staff initiated CR-IP3-
2016-01035 on April 21, 2016, and performed an immediate operability determination
(IOD) in accordance with Entergy procedure EN-OP-104 Section 5.3, to evaluate the
baffle bolts and baffle-former assembly structure on Unit 3. Entergy staff planned further
corrective actions to move up the planned Unit 3 baffle bolt ultrasonic examinations to
the next RFO in spring 2017.
The inspectors reviewed the design basis and current licensing basis documents for
Indian Point Unit 3 to identify the specific safety functions of the baffle bolts. The baffle
bolts are part of the baffle former assembly structure located in the reactor pressure
vessel. The bolts secure a series of vertical metal plates called baffle plates, which help
direct water up through the nuclear fuel assemblies to ensure proper cooling of the fuel.
A sufficient number of baffle bolts are required to secure the plates to ensure proper
core flow during normal and postulated accident conditions, and also to ensure that
control rods can be inserted to shut down the reactor.
The inspectors reviewed Entergys IOD issued on April 21, 2016, and concluded the
immediate determination was completed in accordance with Section 5.3 of procedure
EN-OP-104. The IOD provided sufficient technical detail to support the initial conclusion,
based on limited information, that the Unit 3 baffle bolts would retain sufficient capability
to perform their intended functions. Specifically, the IOD stated that Unit 2 baffle bolt
failures were likely due to irradiation-assisted stress corrosion cracking (IASCC) and that
the Unit 3 baffle bolts were also susceptible because they both utilize a baffle bolt design
with similar geometry and material to other plants with bolt failures. The IOD concluded
that Unit 3 baffle bolt degradation would likely not be as significant as Unit 2, and that
the Unit 3 baffle former assembly was currently operable pending further evaluation
because of the following differences with Unit 2: (1) less effective full power years of
operation; (2) less neutron fluence levels (i.e., irradiation); (3) less pressure differential
across the baffle plates; and (4) less fatigue-induced loading cycles on the bolts over the
operating life of the plant. The inspectors concluded that there was no immediate safety
concern.
On May 5, 2016, Entergy staff revised the operability input for CR-IP3-2016-01035 under
corrective action #2. The inspectors noted that Entergy staff concluded an operability
evaluation was not needed, in part, because the baffle-former bolts are not required by
TS and are not described in the UFSAR. The inspectors noted that while the baffle
bolts are not described in these documents, their failure in sufficient numbers could have
consequential effects on the TS-controlled ECCS if the baffle plates were to become
detached or deformed. This was described in Entergys bolt pattern analysis report
16
documenting an acceptable bolt pattern prior to the spring 2016 RFO. The inspectors
reviewed Unit 3 TS 3.5.2, ECCS - Operating, which requires multiple trains of ECCS to
be operable. The inspectors concluded that since the baffle bolts support the ECCS,
which is subject to TS, Entergys decision to not perform further evaluation of the
operability determination was inconsistent with EN-OP-104. Specifically, Section 5.1(7)
of Entergys procedure EN-OP-104 requires that an operability determination be
performed whenever a condition exists in the supporting SCC that may affect the ability
of the TS-controlled SSC to perform its specified safety function.
Further, the inspectors noted that Entergy staff concluded a degraded condition did not
exist in Unit 3, and therefore, an operability evaluation was not required as a follow-up to
the immediate determination. The documented basis provided was the differences
between the two units, plant operating data, and fuel performance. The inspectors noted
that plant operating data and fuel performance from Unit 2 did not result in identification
of the bolt degradation; therefore, the absence of indications for these problems on Unit
3 was technically insufficient to support Entergys conclusion that there was no degraded
condition on Unit 3.
The inspectors review of procedure EN-OP-104, Section 3.0, identified that examples of
the effects of equipment aging and operating experience can be sources of information
considered to enter the operability or functionality process. The inspectors
acknowledged that licensees apply judgment in these decisions. In this particular
instance, the inspectors considered that operating experience was available that showed
the Unit 3 baffle bolts were subject to IASCC and that plants of similar design (4-loop
Westinghouse pressurized water reactors with a down-flow configuration and baffle bolts
of 347 material and similar dimensions) were subject to greater amounts of bolt
degradation compared to other reactor designs. Furthermore, the inspectors noted the
baffle bolts had experienced levels of neutron radiation exposure above the threshold for
IASCC initiation as referenced in NUREG/CR-7027, Degradation of LWR Core Internal
Materials due to Neutron Irradiation.
Based on the above information available to Entergy staff, the inspectors concluded that
Entergys basis for determining that a degraded condition did not exist on Unit 3 was not
technically supported. The inspectors noted that in completing an IOD in EN-OP-104,
Step 5.3.2 states determine if there is an ongoing degradation mechanism that may
impact future operability based on changing conditions, specifically consider the SSCs
specified safety function and mission time. On May 5, 2016, Entergys basis for
concluding an operability evaluation was not required and exiting the operability
determination procedure at Step 5.3.3 was inconsistent with this procedural requirement
because their IOD concluded Unit 3 was susceptible to baffle bolt degradation, which is
time based and subject to changing conditions including fatigue inducing loading cycles
and neutron fluence. As a result, the inspectors concluded Entergy staff did not
complete the additional actions prescribed by EN-OP-104 to perform an operability
evaluation. Specifically, Step 5.3.9 states in part if an Operability Evaluation is required
then perform the following: Proceed to Subsection 5.5, Operability Evaluation.
On July 11, 2016, Entergy staff subsequently completed the steps in EN-OP-104 and
performed an operability evaluation, which assumed an estimated number of baffle-
former bolt failures based on the degradation found in Unit 2, and adjusted to take credit
for the small number of inaccessible bolts and a sample of bolts extracted with high
removal torque that indicated residual structural capacity. The inspectors determined
17
this estimated number of bolt failures was conservative because the evaluation did not
credit the baffle-edge bolts or the differences in operational history between the two units
such as neutron fluence levels or fatigue from thermal cycles. The operability evaluation
concluded that the Unit 3 baffle bolts would perform as intended to secure the baffle
plates from being dislodged. The inspectors concluded that Entergys operability
evaluation provided appropriate basis to conclude that the Unit 3 baffle assembly would
support ECCS operability until the planned Unit 3 RFO in spring 2017.
Analysis. The inspectors determined that Entergys failure to adequately accomplish the
actions prescribed in EN-OP-104 for a degraded condition and perform an operability
evaluation associated with the Unit 3 baffle-former bolts was a performance deficiency.
Specifically, Entergy incorrectly concluded that no degraded or non-conforming condition
existed related to the Unit 3 baffle-former bolts and exited the operability determination
procedure. As a result, Entergys initial documentation did not provide sufficient basis
for operability and continued operation until questioned by NRC inspectors.
This finding is more than minor because it is associated with the equipment performance
attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to
ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences (i.e., core damage). This issue was also
similar to example 3.j of IMC 0612, Appendix E, Examples of Minor Issues, because
the condition resulted in reasonable doubt of operability of the ECCS and additional
analysis was necessary to verify operability. In accordance with IMC 0609.04, Initial
Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance
Determination Process for Findings At-Power, issued June 19, 2012, the inspectors
screened the finding for safety significance and determined it to be of very low safety
significance (Green), since the finding did not represent an actual loss of system or
function. After inspector questioning, Entergy performed an operability evaluation, which
provided sufficient bases to conclude the Unit 3 baffle assembly would support ECCS
operability. This finding is related to the cross-cutting aspect of Problem Identification
and Resolution, Operating Experience, because Entergy did not effectively evaluate
relevant internal and external operating experience. Specifically, Entergy did not
adequately evaluate the impact of degraded baffle bolts at Unit 3 when relevant
operating experience was identified at Unit 2. [P.5]
Enforcement. 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, states, in part, that activities affecting quality shall be prescribed by
documented procedures of a type appropriate to the circumstances and shall be
accomplished in accordance with those procedures. The introduction to Appendix B
states that quality assurance comprises all those planned and systematic actions
necessary to provide adequate confidence that a structure, system, or component (SSC)
will perform satisfactorily in service. Procedure EN-OP-104, Step 5.3[2], related to
immediate operability, states Determine if there is an ongoing degradation mechanism
that may impact future operability based on changing conditions, specifically consider
the SSCs specified safety function and mission time. Step 5.3(3) follows with, in part If
no Degraded or Non-conforming Condition exists, then perform the following as the
Immediate Determination: Declare the SSC Operable and Exit this procedure.
Contrary to the above, from May 5, 2016 until July 11, 2016, Entergy did not adequately
accomplish actions as prescribed by EN-OP-104 for a degraded condition associated
with the Unit 3 baffle-former bolts. Specifically, Entergy incorrectly concluded that no
18
degraded or non-conforming condition existed related to the Unit 3 baffle-former bolts
and exited the operability determination procedure. The NRC determined this is contrary
to EN-OP-104 because a comparison of Unit 2 and 3 operational factors resulted in
Entergy concluding that the Unit 3 baffle bolts would likely be affected due to the same
degradation mechanism. Entergys corrective actions included entering the issue into
the CAP and documenting an operability evaluation to support the basis for operability of
the baffle bolts and ECCS. Because this issue is of very low safety significance (Green)
and Entergy entered this into their CAP as CR-IP3-2016-01961, this finding is being
treated as an NCV consistent with Section 2.3.2.a of the Enforcement Policy. (NCV
05000286/2016002-02, Failure to Follow Operability Determination Procedure for
Unit 3 Baffle-Former Bolts)
Update to URI 05000247/2016001-06, 23 Emergency Diesel Generator Automatic
Voltage Regulator Failure
Introduction. The NRC opened a URI in Inspection Report 05000247/2016001 related to
two failures of the 23 EDG to run and maintain bus voltage on March 7, 2016, and to
provide adequate control of bus voltage on March 10, 2016. This report provides an
update of the status of this URI.
Description. On March 7, 2016, approximately one hour after the trip of the 3A normal
feed breaker, the 23 EDG tripped on overcurrent while powering the 6A 480V safety bus.
The 6A bus remained de-energized for approximately one hour until the crew restored
the 6A bus via off-site power. The 23 EDG was declared inoperable. All four 480V
safety buses were restored to off-site power. Entergy replaced the overcurrent relays
and retested the 23 EDG satisfactorily on March 8, 2016. However, bench testing of the
overcurrent relays demonstrated that they were accurately calibrated.
Subsequently, on March 10, 2016, during performance of PT-R14, Automatic Safety
Injection System Electrical Load and Blackout Test, the 23 EDG exhibited anomalous
behavior during the train B load sequencing. During this test, the voltage on safety bus
6A dropped to approximately 200V when the 23 auxiliary feedwater pump was
sequenced onto the bus (CR-IP2-2016-01430) and the sequencer failed to complete the
first two sequences. The 23 EDG was again declared inoperable and the period of
inoperability was backdated to March 7, 2016, when it originally tripped. Further
troubleshooting and additional failure modes analysis by Entergy initially determined that
the cause of both events may have been a degraded resistor (R25) on the 23 EDG
automatic voltage regulator (AVR) card.
The 23 EDG AVR card was replaced, and the 23 EDG was again tested satisfactorily.
The voltage anomaly issues exhibited during the March 10, 2016, test were documented
in CR-IP2-2016-01430 which was closed in CR-IP2-2016-01260 to be included in the
causal assessment associated with the tripping of 23 EDG breaker on March 7, 2016.
Entergy assigned a vendor to perform laboratory bench testing and failure analysis of
the 23 EDG AVR card. The vendor report attributed the cause of the March 10, 2016,
loss of voltage control to a degraded solder joint on the AVR card. However, the vendor
report explicitly did not attribute the event on March 7, 2016, to the same cause.
Entergy assigned a corrective action in CR-IP2-2016-01260 to review the cause of the
19
23 EDG overcurrent trip on March 7, 2016, in light of the vendor report. The inspectors
determined that the issue of concern remains open as a URI until this causal
assessment has been completed by Entergy and assessed by NRC. (URI
05000247/2016001-06, 23 Emergency Diesel Generator Automatic Voltage
Regulator Failure)
1R18 Plant Modifications (71111.18 - 2 samples)
Permanent Modifications
.1 Control Rod Guide Tube Repairs in Location E-9
a. Inspection Scope
The inspectors evaluated a modification to the reactor vessel upper internals to swap
damaged control rod guide tube in location E-9 with abandoned guide tube in location
D-10. The inspectors verified that the design bases, licensing bases, and performance
capability of the affected systems were not degraded by the modification. In addition,
the inspectors reviewed modification documents associated with the design change,
including evaluation of equivalency and core flow changes, and post-modification
testing. The inspectors also reviewed revisions to the affected drawings and interviewed
refueling and engineering personnel.
b. Findings
No findings were identified.
.2 Core Baffle-Former Bolt EC 64038
a. Inspection Scope
The inspectors reviewed EC 64038, IP2 Reactor Vessel Equivalent Replacement
Baffle-to-Former Bolt. This modification was completed during RFO 2R22 and involved
the replacement of 278 baffle-former bolts out of a total of 832 located in the Unit 2
reactor vessel. Entergy replaced all of the bolts that were potentially degraded as
observed by visual indications of a protruding bolt head or lock bar problem, bolts that
did not pass UT, and bolts inaccessible for UT. Entergy staff also replaced 51 additional
bolts that passed ultrasonic and visual examinations to increase the structural margin of
the baffle-former assembly for future operating cycles.
The inspectors reviewed the equivalency evaluation completed by Entergy staff to install
baffle-former bolts of a different material and configuration than the original bolts. The
inspectors reviewed the associated EC package to determine whether the replacement
bolts form, fit, and function were maintained compared to the original bolts and whether
the change conformed to the design and licensing bases of the baffle-former assembly.
Specifically, this change involved replacing the original baffle-former bolts made of
type 347 stainless steel with bolts made of type 316 stainless steel. The baffle-former
bolt head configuration was also changed from an original internal hex and slot design
(secured with a welded lock bar) to an external hex configuration with an integral locking
cup design. The design change document further evaluated a more gradual fillet
20
geometry between the bolt head and shank intended to reduce the stress concentration
at that transition and provide for improved fatigue resistance.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19 - 8 samples)
a. Inspection Scope
The inspectors reviewed the post-maintenance tests for the maintenance activities listed
below to verify that procedures and test activities ensured system operability and
functional capability. The inspectors reviewed the test procedure to verify that the
procedure adequately tested the safety functions that may have been affected by the
maintenance activity, that the acceptance criteria in the procedure was consistent with
the information in the applicable licensing basis and/or design basis documents, and that
the test results were properly reviewed and accepted and problems were appropriately
documented. The inspectors also walked down the affected job site, observed the
pre-job brief and post-job critique where possible, confirmed work site cleanliness was
maintained, witnessed the test or reviewed test data to verify quality control hold points
were performed and checked, and that results adequately demonstrated restoration of
the affected safety functions.
Unit 2
21 EDG fuel oil transfer pump after planned maintenance on May 5, 2016
Replacement of pressurizer level transmitters LT-459 and LT-460 on May 25, 2016
21 CCW heat exchanger service water outlet weld repair on June 26, 2016
Flux mapping system drive repairs following motor failures on June 28, 2016
Unit 3
Maintenance on service water components associated with the 32 EDG on May 5,
2016 (this sample was part of an in-depth review of the EDG system)
Modification of the 32 EDG space heaters on May 5, 2016 (this sample was part of
an in-depth review of the EDG system)
Maintenance on the 32 EDG air start system on May 6, 2016 (this sample was part
of an in-depth review of the EDG system)
Replacement of failed bistable LC-427K, steam generator 32 low level mismatch trip
interlock, on May 18, 2016
b. Findings
No findings were identified.
21
1R20 Refueling and Other Outage Activities (71111.20 - 2 samples)
.1 Unit 2 RFO 2R22
a. Inspection Scope
The inspectors reviewed the stations work schedule and outage risk plan for the Unit 2
maintenance during RFO 2R22, which was conducted from March 7, 2016, to June 16,
2016. The inspectors reviewed Entergys development and implementation of outage
plans and schedules to verify that risk, industry experience, previous site-specific
problems, and defense-in-depth were considered. During the outage, the inspectors
observed portions of the shutdown and cooldown processes and monitored controls
associated with the following outage activities:
Configuration management, including maintenance of defense-in-depth,
commensurate with the outage plan for the key safety functions and compliance with
the applicable TSs when taking equipment OOS
Implementation of clearance activities and confirmation that tags were properly hung
and that equipment was appropriately configured to safely support the associated
work or testing
Installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication and instrument error accounting
Status and configuration of electrical systems and switchyard activities to ensure that
TSs were met
Monitoring of decay heat removal operations
Impact of outage work on the ability of the operators to operate the spent fuel pool
cooling system
Reactor water inventory controls, including flow paths, configurations, alternative
means for inventory additions, and controls to prevent inventory loss
Activities that could affect reactivity
Maintenance of secondary containment as required by TSs
Refueling activities, including fuel handling and fuel receipt inspections
Fatigue management
Tracking of startup prerequisites, walkdown of the primary containment to verify that
debris had not been left which could block the ECCS suction strainers, and startup
and ascension to full power operation
Foreign Object Search and Retrieval for missing baffle bolts and locking tabs
Identification and resolution of problems related to RFO activities
During this outage, Entergy replaced 278 degraded baffle bolts in the Unit 2 reactor
vessel baffle assembly. This emergent project resulted in the extension of the outage
schedule from 30 days to 102 days.
b. Findings
Introduction. The inspectors identified a Green NCV of TS 5.4.1 for Entergys failure to
implement procedure OAP-007, Containment Entry and Egress. Specifically, workers
transiting the inner and outer crane wall sections of containment on June 11, 2016, failed
to maintain at least one (of two) flow channeling gate closed to ensure availability of the
containment sumps to provide suction for the ECCS.
22
Description. On June 11, 2016, in mode 4, in preparation for reactor startup, Entergy
was performing maintenance in containment required prior to mode 3, such as reactor
coolant pump motor balancing and steam flow transmitter troubleshooting. These
activities required scaffolds to be temporarily erected for workers to safely perform
maintenance. While transiting from the inner to outer section of containment, the
inspectors noted that both flow channeling gates were maintained open simultaneously
as workers carried scaffold poles and hardware out of the area.
In the event of a postulated LOCA, Unit 2 relies on two sumps to provide a suction
source for the internal recirculation pumps and residual heat removal pumps,
respectively, after the injection phase of the accident. The sumps have cylindrical
screens with large surface area and small holes to filter small debris and maintain
adequate net positive suction head for the associated pumps. The reactor cavity sump
and large intervening barriers prevent large debris generated from the accident, such as
insulation, from reaching and blocking the recirculation and containment sump screens.
Entergy procedure OAP-007, Containment Entry and Egress, precaution and limitation
step 2.30.2, states, In mode 1, 2, 3, or 4, entry inside the crane wall shall use the
double gate entry point via gates 17 and 23. One gate shall remain shut and secured at
all times to maintain flow channeling and sump operability. Securing gates requires a
padlock or nut and bolt closure from the outside. This will require posting a gate monitor
to allow exit. The inspectors noted, while a gate monitor was posted, both gates were
maintained open during passage and not secured with a padlock or nut and bolt closure.
Upon questioning by the inspectors, Entergy immediately coached the gate monitor and
restored the gates to an acceptable position. Entergy generated CR-IP2-2016-04036 to
address this issue.
Analysis. The inspectors determined that Energys failure to maintain either gate 17 or
gate 23 closed during passage in accordance with OAP-007 was a performance
deficiency. The performance deficiency was more than minor because it is associated
with the configuration control (shutdown equipment lineup) attribute and adversely
affected the Mitigating Systems cornerstone objective to ensure the availability,
reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences (i.e., core damage). The inspectors evaluated the finding in
accordance with IMC 0609, Appendix G, Attachment 1, Exhibit 3, and determined that a
detailed risk evaluation was necessary because the finding represented a loss of system
safety function. A detailed risk assessment was conducted conservatively assuming
complete failure of the recirculation and containment sumps due to the performance
deficiency. Given that Unit 2 was in mode 4, in plant operating state 1, with a late time
window, the at-power simplified plant analysis risk model for large-break LOCAs was
determined to best model the degrade condition and plant response. An exposure time
of one day was assumed. No credit was assumed for the decrease in energy that would
be anticipated in a release during a LOCA in mode 4, nor the corresponding reduction in
debris generation. This was also considered conservative. Utilizing Systems Analysis
Program for Hands-On Integrated Reliability Evaluation, version 8.13, with Indian Point
Unit 2 Simplified Plant Analysis Risk Model, version 8.19, for the assumed conditions,
the change in core damage frequency was determined to be 7E-9. Therefore, this issue
represents a Green finding.
23
This finding had a cross-cutting aspect in the area of Human Performance, Avoid
Complacency, because Entergy did not consider potential undesired consequences of
actions before performing work and implement appropriate error-reduction tools.
Specifically, the work crew did not understand the requirements and potential
consequences prior to commencing work and the gate monitor did not enforce these
requirements to maintain at least one gate locked or pinned closed as required by
OAP-007. [H.12]
Enforcement. Unit 2 TS 5.4.1.a requires that the procedures listed in Attachment A to
Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, be
established and implemented. Attachment A states that instructions should be prepared,
as appropriate, for access to containment and changing modes of operation of the
ECCS. Entergy procedure OAP-007, Containment Entry and Egress, Step 2.30.2,
states, In mode 1, 2, 3, or 4, entry inside the crane wall shall use the double gate entry
point via gates 17 and 23. One gate shall remain shut and secured at all times to
maintain flow channeling and sump operability. Securing gates requires a padlock or nut
and bolt closure from the outside. Contrary to the above, on June 11, 2016, Entergy did
not maintain one gate secured at all times with a padlock or nut and bolt closure.
Entergy entered this issue into the CAP as CR-IP2-2016-04036. Because this violation
was of very low safety significance (Green), and Entergy entered this performance
deficiency into the CAP, the NRC is treating this as a NCV in accordance with
Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000247/2016002-03, Failure
to Maintain Flow Channeling Gates Closed in Accordance with the Containment
Procedure)
.2 Unit 2 Forced Outage
a. Inspection Scope
Unit 2 conducted a forced outage from June 24 to 27, 2016, in order to complete weld
repairs on a through-wall leak on the service water inlet line to the 21 CCW heat
exchanger. These repairs required shutting down to mode 4 in order to meet the
TS 3.7.7, Component Cooling Water (CCW) System, limiting condition for operations
for CCW operability. While these repairs were being completed, the grid operator
completed repairs to breaker 9 in the offsite switchyard. During the outage, the
inspectors observed portions of the shutdown and cooldown processes and monitored
controls associated with the following outage activities:
Configuration management, including maintenance of defense-in-depth,
commensurate with the outage plan for the key safety functions and compliance with
the applicable TSs when taking equipment OOS
Implementation of clearance activities and confirmation that tags were properly hung
and that equipment was appropriately configured to safely support the associated
work or testing
Status and configuration of electrical systems and switchyard activities to ensure that
TSs were met
Monitoring of decay heat removal operations
Reactor water inventory controls, including flow paths, configurations, alternative
means for inventory additions, and controls to prevent inventory loss
Activities that could affect reactivity
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Tracking of startup prerequisites
Identification and resolution of problems related to RFO activities
When all repairs had been completed, Entergy restarted Unit 2 on June 27, 2016.
b. Findings
No findings were identified.
1R22 Surveillance Testing (71111.22 - 6 samples)
a. Inspection Scope
The inspectors observed performance of surveillance tests and/or reviewed test data of
selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR,
and Entergys procedure requirements. The inspectors verified that test acceptance
criteria were clear, tests demonstrated operational readiness and were consistent with
design documentation, test instrumentation had current calibrations and the range and
accuracy for the application, tests were performed as written, and applicable test
prerequisites were satisfied. Upon test completion, the inspectors considered whether
the test results supported that equipment was capable of performing the required safety
functions. The inspectors reviewed the following surveillance tests:
Unit 2
WO 446385, 21 EDG AVR card inspection, on May 24, 2016
2-PT-Q013 for containment isolation valve 851B (22 safety injection (SI) pump tie to
23 SI pump discharge) on June 6, 2016
2-PT-Q029B quarterly in-service surveillance test for the 22 SI pump on June 6,
2016
Unit 3
3-PT-M079B 32 EDG monthly surveillance on May 30, 2016 (this sample was part of
an in-depth review of the EDG system)
34 steam generator pressure instrument channel check on June 21, 2016
0-SOP-LEAKRATE-001, RCS Leakrate Surveillance, Evaluation and Leak
Identification, beginning on June 28, 2016
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
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1EP6 Drill Evaluation (71114.06 - 1 sample)
Training Observations
a. Inspection Scope
The inspectors evaluated the conduct of Entergys ingestion pathway emergency
preparedness drill on April 19, 2016, to identify any weaknesses and deficiencies in the
classification, notification, and protective action recommendation development activities.
The inspectors observed emergency response operations in the emergency operations
facility to determine whether the event classification, notifications, and protective action
recommendations were performed in accordance with procedures. The inspectors also
attended the facility drill critique to compare inspector observations with those identified
by Entergy staff in order to evaluate Entergys critique and to verify whether the staff was
properly identifying weaknesses and entering them into the CAP.
b. Findings
No findings were identified.
2. RADIATION SAFETY
Cornerstone: Public Radiation Safety and Occupational Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)
a. Inspection Scope
During May 10-12 and June 13-17, 2016, the inspectors reviewed Entergys
performance in assessing the radiological hazards and exposure control in the
workplace. The inspectors used the requirements in 10 CFR 20, TSs, applicable
industry standards, and procedures required by TSs as criteria for determining
compliance.
Radiological Hazards Control and Work Coverage
The inspectors reviewed:
Ambient radiological conditions during tours of the radiological controlled area,
posted surveys, radiation work permits, adequacy of radiological controls, radiation
protection job coverage, and contamination controls
Controls for highly activated or contaminated materials stored within spent fuel pools
Posting and physical controls for high radiation areas and very high radiation areas
b. Findings
No findings were identified.
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2RS2 Occupational As Low As Is Reasonably Achievable (ALARA) Planning and Controls
(71124.02)
a. Inspection Scope
During May 10-12 and June 13-17, 2016, the inspectors assessed performance with
respect to maintaining occupational individual and collective radiation exposures ALARA.
The inspectors used the requirements in 10 CFR 20, TSs, applicable industry standards,
and procedures required by TSs as criteria for determining compliance.
Radiological Work Planning
The inspectors reviewed:
ALARA work activity evaluations, exposure estimates, and exposure mitigation
requirements
ALARA work planning, use of dose mitigation features and dose goals
Work planning and the integration of ALARA requirements
b. Findings
No findings were identified.
2RS7 Radiological Environmental Monitoring Program (REMP) (71124.07 - 3 samples)
a. Inspection Scope
The inspectors reviewed the REMP to validate the effectiveness of the radioactive
gaseous and liquid effluent release program and implementation of the groundwater
protection initiative (GPI). The inspectors used the requirements in 10 CFR 20,
40 CFR 190, 10 CFR 50, Appendix I, TSs, offsite dose calculation manual (ODCM),
Nuclear Energy Institute 07-07, and procedures required by TSs as criteria for
determining compliance.
Inspection Planning
The inspectors reviewed Entergys 2014 and 2015 annual radiological environmental
and effluent monitoring reports, REMP program audits, ODCM changes, land use
census, the UFSAR, and inter-laboratory comparison program results.
Site Inspection
The inspectors walked down various thermoluminescent dosimeter and air and water
sampling locations and reviewed associated calibration and maintenance records. The
inspectors observed the sampling of various environmental media as specified in the
ODCM and reviewed any anomalous environmental sampling events including
assessment of any positive radioactivity results. The inspectors reviewed any changes
to the ODCM. The inspectors verified the operability and calibration of the
meteorological tower instruments and meteorological data readouts. The inspectors
reviewed environmental sample laboratory analysis results, laboratory instrument
measurement detection sensitivities, laboratory quality control program audit results, and
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the inter- and intra-laboratory comparison program results. The inspectors reviewed the
groundwater monitoring program as it applies to selected potential leaking SSCs.
GPI Implementation
The inspectors reviewed groundwater monitoring results, changes to the GPI program
since the last inspection, anomalous results or missed groundwater samples, leakage or
spill events including entries made into the decommissioning files (10 CFR 50.75(g)),
evaluations of surface water discharges, and Entergys evaluation of any positive
groundwater sample results including appropriate stakeholder notifications and effluent
reporting requirements.
Identification and Resolution of Problems
The inspectors evaluated whether problems associated with the REMP were identified at
an appropriate threshold and properly addressed in Entergys CAP.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151 - 6 samples)
Initiating Events Performance Indicators
a. Inspection Scope
The inspectors reviewed Entergys submittals for the following Initiating Events
cornerstone performance indicators for the period April 1, 2015, to March 31, 2016:
Unit 2
Unplanned scrams per 7000 critical hours (IE01)
Unplanned power changes per 7000 critical hours (IE03)
Unplanned scrams with complications (IE04)
Unit 3
Unplanned scrams (IE01)
Unplanned power changes (IE03)
Unplanned scrams with complications (IE04)
To determine the accuracy of the performance indicator data reported during those
periods, inspectors used definitions and guidance contained in Nuclear Energy
Institute 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7.
The inspectors reviewed Entergys operator narrative logs, maintenance planning
schedules, CRs, event reports, and NRC integrated inspection reports to validate the
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accuracy of the submittals. There were no unplanned power changes or scrams with
complications during the review period.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution (71152 - 4 samples)
.1 Routine Review of Problem Identification and Resolution Activities
a. Inspection Scope
As required by Inspection Procedure 71152, Problem Identification and Resolution, the
inspectors routinely reviewed issues during baseline inspection activities and plant
status reviews to verify that Entergy entered issues into the CAP at an appropriate
threshold, gave adequate attention to timely corrective actions, and identified and
addressed adverse trends. In order to assist with the identification of repetitive
equipment failures and specific human performance issues for follow up, the inspectors
performed a daily screening of items entered into the CAP and periodically attended CR
screening meetings. The inspectors also confirmed, on a sampling basis, that, as
applicable, for identified defects and non-conformances, Entergy performed an
evaluation in accordance with 10 CFR 21.
b. Findings
No findings were identified.
.2 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a semi-annual review of site issues, as required by Inspection
Procedure 71152, Problem Identification and Resolution, to identify trends that might
indicate the existence of more significant safety issues. In this review, the inspectors
included repetitive or closely-related issues that may have been documented by Entergy
outside of the CAP, such as trend reports, performance indicators, major equipment
problem lists, system health reports, maintenance rule assessments, and maintenance
or CAP backlogs. The inspectors also reviewed Entergys CAP database for the first
and second quarters of 2016 to assess CRs written in various subject areas (equipment
problems, human performance issues, etc.), as well as individual issues identified during
the NRCs daily CR review (Section 4OA2.1). The inspectors reviewed the Entergy
quarterly trend report for the first quarter of 2016 to verify that Entergy was appropriately
evaluating and trending adverse conditions in accordance with applicable procedures.
b. Findings and Observations
No findings were identified.
The inspectors identified a trend in work being performed that was contrary to written
work instructions and procedures, and work packages had been closed out without
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documenting the deviation from the work order. While reviewing completed work order
WO 447966 Task 10, Internal Coating Repair for 21 CCW HX, the inspectors found a
note in the work order stating that the internal coating repair to the pipe had not been
done in accordance with the engineering change. The engineering change had been
written when the coating repair was expected to be small, but the actual area that was
recoated was much larger. A larger area of coating increases the impact on the heat
exchanger if the coating were to flake off and block the flow of service water. The work
package was closed and no condition report was written. This performance deficiency is
minor because the coating was applied with procedurally directed quality controls and
the likelihood that it would flake off is very small; and is the same as the original smaller
area specified in the work package. However, the work package was closed without
documenting the deviation and no CR was written.
In another example, the inspectors noted that WO 412920 Task 15 to perform a surge
test on 11 centrifugal air compressor (CENTAC) after overhaul was completed on
December 22, 2015. However, the completion notes and documentation for the task
showed that the test was unable to be performed due to a test equipment problem. The
work package was closed and no CR was written. Subsequently, after being returned to
service, the compressor failed in service due to multiple surging events on January 7,
2016. Troubleshooting under WO 433939 revealed that the motor high load limit had not
been adjusted to account for the increased load due to reduced compressor clearances
introduced by the overhaul. This performance deficiency is screened to minor because
the 11 CENTAC is not a safety-related piece of equipment and would not affect the MC
0609 cornerstone thresholds or other generic criteria. Unit 2 and Unit 3 have dedicated
instrument air compressors that are credited in the FSAR to respond to a loss of
instrument air event. If the 11 CENTAC (located in Unit 1) were to fail, the unit-specific
IACs would automatically start to prevent a loss of instrument air at both Units 2 and 3.
A third recent example of work being performed contrary to written instructions occurred
during 2RFO22 when the inspectors identified that the workers deviated from the
surveillance procedure by demonstrating the installation of the emergency containment
hatch plug without properly inflating the plug seals as directed by the procedure. This
performance deficiency was previously documented in a prior inspection report as non-
cited violation 05000247/05000286/2016001-02, Failure to Adequately Implement Risk
Management Actions for the Containment Key Safety Function.
In all cases, the deviations from written work instructions were directed by Entergy
supervision. In addition, the inspectors noted that Entergy had self-identified similar
observations where work packages or condition reports had been closed without fully
completing the specified actions including CR-IP2-2015-05833, CR-IP2-2016-00103,
CR-IP3-2015-04729, CR-IP3-2016-00072, CR-IP3-2016-00075, and CR-IP3-2015-
04019. These CRs are further examples of work orders that were closed with deviations
that were not documented or resolved. Nuclear Oversight had identified several of these
condition reports. Entergy has taking immediate corrective action in response to these
performance deficiencies.
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.3 Annual Sample: Maintenance Rule Self-Assessment of Corrective Actions
a. Inspection Scope
The inspectors performed an in-depth review of Entergys corrective actions associated
with self-assessment LO-IP3LO-2015-72, Maintenance Rule (a)(3) Assessment. The
self-assessment was performed to satisfy both the self-assessment criteria in EN-LI-104,
Self-Assessment and Benchmark Process, and the maintenance rule periodic
assessment criteria in EN-DC-207.
The inspectors assessed Entergys problem identification threshold, extent of condition
reviews, and the prioritization and timeliness of Entergy corrective actions to determine
whether Entergy was appropriately identifying, characterizing, and correcting problems
associated with this issue and whether the planned or completed corrective actions were
appropriate. The inspectors compared the actions taken to the requirements of
Entergys CAP and 10 CFR 50, Appendix B. In addition, the inspectors interviewed
engineering personnel to assess the effectiveness of the implemented corrective
actions.
b. Findings and Observations
No findings were identified.
Entergy identified three standard deficiencies during their self-assessment and wrote
CRs to document each one. One of the standard deficiencies was that the maintenance
rule basis documents were not being reviewed at least once every two years as required
by procedure EN-DC-204, Maintenance Rule Scope and Basis. The purpose of this
review was to ensure that the documents were updated if the configuration of the system
changed or if the performance criteria needed to be adjusted. Entergy wrote CR-IP3-
2015-03628 and assigned a corrective action to create work trackers to perform the
basis document reviews. They chose to use work trackers instead of corrective actions
under the CAP because the work had historically been assigned using work trackers.
However, because work trackers do not receive the same priority as corrective actions,
some of the maintenance rule basis documents had still not been reviewed at the time of
this inspection, over a year after the completion of the self-assessment. The inspectors
determined that this was not a more than minor issue because the systems in question
did not show signs of inadequate maintenance.
.4 Annual Sample: Unit 2 Reactor Trip on December 5, 2015
a. Inspection Scope
The inspectors performed an in-depth review of Entergys evaluations and corrective
actions associated with CR-IP2-2015-05484 and the related apparent cause evaluation
for the December 5, 2015, manual reactor trip in response to indications of multiple
dropped control rods caused by the loss of control rod power due to a power supply
failure. Entergy performed an apparent cause evaluation and determined the direct
cause of the event was the loss of motor control center (MCC)-24 due to an internal fault
at the line side leads at cubicle 2H where they connect to the bucket stab assemblies.
The apparent cause was an unanticipated loss of power to the control rod system due to
the degradation of the primary control rod power supply (PS1) which failed to function for
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more than 10 minutes when the operating alternate power supply (PS2) was
deenergized.
The inspectors assessed Entergys problem identification threshold, problem analysis,
extent of condition reviews, compensatory actions, and the prioritization and timeliness
of Entergy's corrective actions to determine whether Entergy was appropriately
identifying, characterizing, and correcting problems associated with this issue and
whether the planned or completed corrective actions were appropriate. The inspectors
compared the actions taken to the requirements of Entergy's CAP and 10 CFR 50,
Appendix B, Criterion XVI, Corrective Action.
b. Findings and Observations
No findings were identified.
The inspectors found that Entergy took appropriate actions to identify the direct and
apparent cause of the issue. The direct cause of the event was the loss of MCC-24 due
to an internal fault at the line side leads at cubicle 2H where they connect to the bucket
stab assemblies. The apparent cause was an unanticipated loss of power to the control
rod system due to the degradation of the primary control rod PS1, which failed to
function when PS2 was lost. Entergy replaced the degraded rod control PS1; and the
MCC-24 compartments were removed to facilitate inspection and testing of the MCC
bus, control wires, and MCC internal. PS2 was also restored to operation after the fault
was cleared.
The inspector determined that the internal electrical fault that deenergized PS2 and the
prior degradation in PS1 was not within Entergys ability to foresee and prevent.
Therefore, there was no performance deficiency identified. Entergys overall response to
the issue was commensurate with the safety significance, was timely, and the actions
taken and planned were reasonable to resolve the failure of the primary control rod PS1.
.5 Annual Sample: Unexpected Number of Degraded Baffle-Former Bolts Discovered in
the Unit 2 Reactor Pressure Vessel
a. Inspection Scope
The inspectors performed an in-depth review of Entergys root cause evaluation and
corrective actions associated with CR-IP2-2016-02348 for baffle-former (baffle) bolts
found with indications of degradation during the Indian Point Unit 2 RFO 2R22. Entergy
performed ultrasonic examinations of the baffle bolts in accordance with their procedures
as part of a planned activity. After an unexpected number of degraded baffle bolts were
discovered, Entergy staff reported the issue to the NRC as Event Notification No. 51829
on March 29, 2016, because the as-found number and location of degraded bolts
represented an unanalyzed condition. Entergy staff completed corrective actions to
replace all of the potentially degraded baffle bolts on Unit 2. Entergy staff further
replaced a population of additional bolts that exhibited no indications of degradation and
performed an evaluation to determine the potential for baffle bolt failures at Unit 3.
The baffle-former bolts help secure vertical plates (also referred to as baffle plates)
inside the reactor vessel, which then forms a structure surrounding the reactor fuel
assemblies to orient the fuel and to direct coolant flow through the core. A sufficient
32
number of baffle bolts are required to remain intact to secure the baffle plates in place so
as to not affect control rod insertion or impede emergency core cooling flow during
postulated accident conditions. Bolt heads that separate and are no longer held in place
by bolt lock-tabs can also become a loose parts concern.
The inspectors determined whether Entergys acceptable baffle bolt pattern analysis for
Unit 2 was completed in accordance with the NRC-approved methodology and provided
appropriate structural margin for the next cycle of operation to ensure the Unit 2 baffle
plates will remain in place during both normal operation and limiting postulated accident
conditions. The inspectors further determined whether Entergys evaluations of the
baffle bolts installed in Indian Point Unit 3 were technically sufficient to conclude the
Unit 3 baffle assembly will perform as intended until the next planned RFO, at which time
Entergy plans to examine the bolts. The inspectors reviewed Entergys procedures for
determining the functionality and operability of degraded SSC as they relate to Unit 3.
The inspectors further interviewed Entergy engineering personnel and contractor staff to
discuss the results of Entergys technical evaluations and to assess the effectiveness of
the implemented and planned corrective actions.
The inspectors assessed Entergys problem identification threshold, cause analyses,
extent of condition, compensatory actions, and the prioritization and timeliness of
Entergys corrective actions to determine whether Entergy staff were properly identifying,
characterizing, and correcting problems associated with this issue and whether the
planned or completed corrective actions were appropriate. The inspectors compared the
actions taken to Entergys CAP, operability determination process, and the requirements
of 10 CFR 50, Appendix B. The inspectors observed portions of baffle bolt replacement
activities at Unit 2 and reviewed the final visual examination of the baffle bolts and plates
once the work was completed.
b. Findings and Observations
One Green NCV was identified and documented in Section 1R15 of this report.
The NRC responded to the initial discovery of an unexpected number of baffle bolts
found degraded at Indian Point Unit 2 by implementing a comprehensive inspection plan
consisting of various baseline inspection samples to assess the extent of the issue and
to determine the necessary NRC actions. A follow-up inservice inspection sample
(Refer to Section 1R08) was conducted to review the capability of the non-destructive
examination techniques, evaluate the UT results, and observe a portion of bolt
replacement activities on-site. A permanent modification sample (Refer to Section
1R18) was conducted to review the design change package and evaluations associated
with the new, replacement baffle bolts. The NRC resident inspectors reviewed Entergys
foreign material controls and loose parts analysis (Refer to Section 1R20) to address the
potential for missing bolt heads and concluded it would not impact safe operation of the
plant.
NRC Region I based inspectors accompanied by an expert from the NRC Office of
Nuclear Reactor Regulation completed an annual problem identification and resolution
inspection, documented in this section of the report, to verify that Entergys evaluations
and corrective actions to replace Unit 2 baffle bolts were completed in accordance with
an NRC approved methodology to support a conclusion that the Unit 2 baffle assembly
meets the plant design basis. The inspectors also determined the adequacy of
Entergys evaluations completed to determine there is a reasonable expectation that the
33
Unit 3 baffle assembly will perform as intended during the current operating cycle. The
results of this review are discussed herein and in Section 1R15 of this report.
Entergy staff determined the cause of the degraded baffle bolts was primarily due to
IASCC in combination with increased fatigue loading on the baffle plates. This cause
determination was based on industry operating experience related to baffle-former bolt
failure in both foreign and domestic plants. IASCC is a cracking mechanism that occurs
over a long period of time when susceptible metals are exposed to neutron radiation
from the reactor core and stresses as part of normal design and operation. Entergy staff
concluded that failure of a critical number of bolts in a localized area subsequently
imposed increased loading on adjacent bolts, which propagated failures and generated
the moderate clustered pattern observed in the examination results. No other
contributing causes were identified.
The inspectors reviewed Entergys root cause evaluation and the supporting operating
experience related to baffle bolt failures at other plants. The inspectors determined that
there is documented evidence in the existing technical literature (including materials
testing of bolts from other plants) and operating experience to conclude that the likely
cause is IASCC; however, the inspectors found that Entergy staff did not define the
cause of the fatigue failure mechanism. The inspectors noted that Entergy staff sent a
sample of baffle bolts removed from the reactor pressure vessel to a metallurgical
laboratory for detailed failure analysis and materials property testing. Entergy indicated
their plans to use the results of the laboratory testing to confirm the likely root cause.
The inspectors concluded that Entergy staff conducted an appropriate review to identify
the likely causes of the degraded baffle bolts and noted that further test results will be
used to confirm these causes.
Following identification of the degraded baffle bolts on Unit 2, Entergys immediate
corrective action was to analyze the as-found condition and begin replacing bolts that
either had visual indications of bolt failure (protruding bolt head for example), did not
pass UT examination, or were not accessible for UT examination. The as-found number
and pattern of these bolts exceeded the acceptance criteria in the plants analysis that
was prepared in advance of the baffle bolt examinations; therefore, Entergy reported this
discovery to the NRC as an unanalyzed condition. Entergy staff completed corrective
actions to replace all of the 227 potentially degraded baffle bolts, plus an additional 51
bolts for increased structural integrity, for a total of 278 bolts. The inspectors noted the
51 additional bolts were installed in strategic locations to prevent clustering of potential
bolt failures during the next operating cycle.
The inspectors determined that Entergy staff performed an acceptable bolt pattern
analysis that evaluated the replacement bolt pattern for Unit 2 and modeled the potential
for future bolt failures. The inspectors found the results of the analysis accounted for a
conservative failure rate of bolts and provided appropriate margin for one cycle of
operation. The inspectors verified that Entergys methodology for its acceptable bolt
pattern analyses, including its determination of margin, was consistent with the NRC-
approved methodology in topical report WCAP-15029-NP-A (ML15222A882). The
inspectors determined that Entergy staff tracked corrective actions to re-examine the
Unit 2 baffle bolts during the next planned RFO. The inspectors noted the new baffle
bolts were made of a material with improved resistance to IASCC and included an
improved design to reduce the stresses at the head to shank transition, both of which
are enhancements compared to the original bolts.
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As part of an extent of condition assessment, Entergy entered CR-IP3-2016-01035 in its
CAP to evaluate the potential for degraded baffle bolts on Unit 3. Entergy operators
performed an IOD and concluded that the baffle assembly was operable. Entergy staff
performed a subsequent extent of condition review that concluded Unit 3 would
experience less baffle bolt degradation than Unit 2 based on several plant factors.
Entergy also conducted sensitivity analyses to show acceptable bounding conditions in
the event of bolt failures. The inspectors reviewed Entergys evaluations and noted that
Entergy staff concluded there was not a degraded condition at Unit 3. In consideration
of the guidance in their operability procedure and operating experience from Unit 2 and
other plants, the NRC issued an NCV in this report because Entergy did not perform an
operability evaluation for Unit 3 as a follow-up to the immediate determination for the
potential impact on supported systems controlled by the TS (Refer to Section 1R15).
As a corrective action, Entergy staff performed an operability evaluation and
demonstrated that the Unit 3 baffle former assembly remained operable. The inspectors
concluded that this supplemental evaluation provided appropriate technical justification
for the continued operation of Unit 3 until the next RFO in spring 2017, at which time
Entergy plans to examine the baffle bolts. Entergy also implemented a corrective action
as part of an enhancement to plant operations to monitor the RCS for any signs of fuel
leakage, which could be an indicator of baffle bolt failures.
The inspectors reviewed Westinghouse Nuclear Safety Advisory Letter NSAL-16-1,
which discussed the results of recent baffle-former bolt inspections and provided
Westinghouses recommendations on this issue. The letter described the plants as most
susceptible (i.e. Tier 1a) to this degradation as Westinghouse 4-loop reactors limited to
those with a down-flow configuration and using Type 347 stainless steel bolts. The
inspectors noted the recommendation was to complete UT volumetric examination of the
baffle bolts at the next scheduled RFO, and that Entergy had already planned this action
for Unit 3. Entergy also planned a long-term corrective action to convert Units 2 and 3
from a down-flow baffle configuration to an up-flow configuration, which would
significantly reduce the load on baffle-former bolts and provide for increased structural
margin of the baffle-former assembly. The inspectors determined Entergys overall
response to the issue was commensurate with the safety significance, was timely, and
included appropriate compensatory actions. The inspectors concluded that the actions
completed and planned were reasonable to address the ongoing aging management of
4OA3 Follow Up of Events and Notices of Enforcement Discretion (71153 - 5 samples)
.1 Plant Events
a. Inspection Scope
For the plant events listed below, the inspectors reviewed and/or observed plant
parameters, reviewed personnel performance, and evaluated performance of mitigating
systems. The inspectors communicated the plant events to appropriate regional
personnel, and compared the event details with criteria contained in IMC 0309, Reactive
Inspection Decision Basis for Reactors, for consideration of potential reactive inspection
activities. As applicable, the inspectors verified that Entergy made appropriate
emergency classification assessments and properly reported the event in accordance
with 10 CFR 50.72 and 50.73. The inspectors reviewed Entergys follow-up actions
35
related to the events to assure that Entergy implemented appropriate corrective actions
commensurate with their safety significance.
Unit 2
Turbine trip occurred while synchronizing Unit 2 to the grid on June 15, 2016
Shutdown required by TS for repairs to a leak on the 21 CCW heat exchanger
service water inlet on June 23, 2016
Unit 3
Rapid power reduction from 100 percent to 45 percent power in response to a loss of
both heater drain pumps on May 26, 2016
b. Findings
No findings were identified.
.2 (Closed) Licensee Event Report (LER) 05000247/2015-003-00: Manual Reactor Trip
Due to Indications of Multiple Dropped Control Rods Caused by Loss of Control Rod
Power Due to a Power Supply Failure
The inspectors reviewed Entergys actions and reportability criteria associated with LER
05000247/2015-003-00, which was submitted to the NRC on February 3, 2016. On
December 5, 2015, control room operators initiated a manual reactor trip after observing
indications consistent with multiple dropped control rods following an alarm for the trip of
MCC-24/24A. No control rod indication was available due to MCC-24 being faulted and
de-energized. The direct cause of the event was the loss of MCC-24 due to an internal
fault at the line sides leads at cubicle 2H where they connect to the bucket stab
assemblies. The apparent cause was an unanticipated loss of power to the control rod
system due to the degradation of the primary control rod PS1 which failed to function
when the operating PS2 was lost. The inspectors determined that both the unexpected
failure of PS2 and the internal fault in PS1 was not within Entergys ability to foresee and
prevent and was not a performance deficiency. The inspectors reviewed the LER, the
associated apparent cause evaluation analysis, and interviewed Entergy staff. This LER
is closed.
.3 (Closed) LER 05000247/2016-003-00: TS Prohibited Condition Due to an Inoperable 21
MBFP Discharge Valve for Greater Than the TS Allowed Outage Time
The inspectors reviewed Entergys actions and reportability criteria associated with LER
05000247/2016-003-00, which was submitted to the NRC on May 6, 2016. On March 7,
2016, during the shutdown to enter 2RFO22, the control switch for the 21 MBFP was
tripped from the control room but the MBFP discharge valve BFD-2-21 failed to fully
close as designed. The MBFP discharge valve was declared inoperable and TS 3.7.3
Condition C was entered. The MFD-2-21 isolation valve was then manually closed. The
direct cause of the failure to close the MBFP discharge valve BFD-2-21 was the motor
operated valves (MOVs) close torque switch contact finger out of position. The
apparent cause was that the MOV preventative maintenance procedure lacked the level
of detail and direction due to an unrecognized susceptibility associated with the
orientation of the close torque switch contact finger bracket opening and spreading of
36
the U shape bracket. The downward arrangement made it easier for the torque switch
contact finger to move out with spreading of the U shaped contact holder. The
inspectors reviewed the LER, the associated apparent cause evaluation analysis, and
interviewed Entergy staff. This LER is closed.
Introduction. The inspectors identified a Green NCV of 10 CFR 50.65(b)(1) for Entergys
failure to include a function of a safety-related system within the scope of the
maintenance rule program. Specifically, Entergy failed to include the feedwater isolation
function performed by the MBFP discharge valves, MBFPs, and feedwater regulating
valves and feedwater isolation valves which are required to remain functional during and
following a design basis event to mitigate the consequences of an accident, within the
scope of the maintenance rule monitoring program.
Description. On March 7, 2016, during an RFO, the control switch for the 21 MBFP was
positioned to trip and the 21 MBFP tripped as designed, but the MBFP discharge valve
BFD-2-21 failed to fully close. Entergy declared MBFP discharge valve BFD-2-21
inoperable and entered TS 3.7.3 Condition C. After troubleshooting, Entergy determined
the MOV close torque switch contact finger was out of position within the contact holder.
The misalignment allowed the contact finger to move out of the proper position causing
the MOV BFD-2-21 to fail to close. This is the same failure mechanism which caused
MOV BFD-2-21 to fail to close in 2010 which is referenced in CR-IP2-2010-07013. On
December 5, 2015, the 21 MBFP failed to trip and required closure of the steam
admission valves to secure it. This failure occurred because of contaminated control oil
that prevented the solenoid valves from operating.
The inspectors reviewed Entergys maintenance rule basis documents and identified the
feedwater isolation function was not properly included in the maintenance rule
monitoring program as required by 10 CFR 50.65(b)(1). The basis document for the
feedwater system did identify the need to monitor the feedwater isolation function under
the maintenance rule and stated that it would be monitored as a part of the vapor
containment supersystem. However, the basis document for the vapor containment
supersystem does not include the feedwater isolation components within the system
boundaries. As a result, when component failures occurred which affected the
feedwater isolation function, they were not reviewed to determine if they were
maintenance rule functional failures; and Entergy was unable to identify that the
performance of the main feedwater isolation equipment was not effectively controlled
through preventative maintenance. Entergy entered this issue into the CAP as
CR-IP2-2016-03963 and initiated actions to include the MBFP discharge valves into the
Analysis. The failure to appropriately scope the safety-related feedwater isolation
function within the maintenance rule program was a performance deficiency. This
finding is more than minor because it is associated with the SSC and barrier
performance attribute of the Barrier Integrity cornerstone and affected the cornerstone
objective to provide reasonable assurance that physical design barriers protect the
public from radionuclide releases caused by accidents or events. Specifically, the failure
to properly scope the feedwater isolation function prevented Entergy from identifying that
equipment reliability was no longer effectively controlled through preventative
maintenance. Additionally, this issue is similar to example 7.d described in IMC 0612,
Appendix E, Examples of Minor Issues, dated August 11, 2009. In accordance with
IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix
37
A, The Significance Determination Process for Findings At-Power, issued June 19,
2012, the inspectors determined that the finding was of very low safety significance
(Green) because the finding did not represent an actual open pathway in the physical
integrity of reactor containment, containment isolation system, and heat removal
components. There are redundant methods of feedwater isolation. They include
tripping the MBFPs and closing the MBFP discharge valves, closing the main feedwater
regulating valves and low flow bypass valves, and closing the main feedwater isolation
valves. On both December 5, 2015, and March 7, 2016, the main feedwater regulating
valves and isolation valves were functional; so there was no loss of the ability to isolate
feedwater to mitigate accident and transient conditions.
This finding does not have a cross-cutting aspect, since the failure to scope this
equipment into the maintenance rule program was not recognized when Entergy
combined the maintenance rule basis documents for Units 2 and 3 in 2012 and as a
result, is not indicative of current licensee performance.
Enforcement. 10 CFR 50.65(b)(1) requires, in part, that the holders of an operating
license shall include within the scope of the monitoring program, specified in
10 CFR 50.65(a)(1), SSCs that are relied upon to remain functional during and following
design basis events. Contrary to the above, since the combined maintenance rule
scoping for Units 2 and 3 in 2012, Entergy failed to include within the scope of the
monitoring program specified in 10 CFR 50.65(a)(1), the safety-related MBFP discharge
valves. These SSCs are relied upon during and after design basis events to mitigate the
consequences of a feedwater line break accident inside containment. Entergys
corrective action included entering this issue into the corrective action program.
Because the violation was of very low safety significance (Green) and Entergy entered
this issue into their CAP as CR-IP2-2016-03963, this violation is being treated as an
NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy.
(NCV 05000247/2016002-04, Failure to Scope Safety-Related Main Boiler Feedwater
Pump Discharge Valves into the Maintenance Rule Program)
4OA5 Other Activities
.1 Groundwater Contamination
a. Inspection Scope
On February 5, 2016, Entergy notified the NRC of a significant increase in groundwater
tritium levels measured at three monitoring wells (MWs) (MW-30, MW-31, and MW-32)
located near the Unit 2 fuel storage building. These samples were drawn on
January 26 to 27, 2016, and analyzed and confirmed on February 2 to 4, 2016. The
highest concentration was detected at MW-32, which increased from 12,000 pCi/l on
January 11, 2016, to 8,100,000 pCi/l on January 26, 2016, and subsequently up to
14,800,000 pCi/l on February 4, 2016. This increased tritium concentration event was
documented by Entergy in CR-IP2-2016-00564 which documents its investigation of this
event including a root cause evaluation. The inspectors reviewed Entergys root cause
evaluation for this event during this inspection period as well as recent groundwater
monitoring results.
38
b. Findings and Observations
No findings were identified.
Update to URI 05000247/2016001-07, January 2016 Groundwater Contamination
Entergy continues to conduct weekly, biweekly, and monthly groundwater sampling of
MWs at the initial site of groundwater contamination and at downstream wells towards
the Hudson River. For the initial three MWs (MW-30, MW-31, and MW-32), the general
trend in tritium activity has been downward, with periodic increases seen in some weekly
samples. The downstream MWs located in the Unit 2 switchyard (especially MW-55)
showed an initial increase in activity up to 117,000 pCi/l, but the activity at that location
has plateaued at the end of the reporting period.
Entergy documented its investigation of this event as root cause evaluation for
CR-IP2-2016-00564. The inspectors reviewed Entergys root cause evaluation for this
event. Entergy concluded that the source of the groundwater contamination was from
the reject water of a temporary reverse osmosis unit used to process water from the
refueling water storage tank at Unit 2 in preparation for RFO 2R22. Although this
analysis documents a number of issues identified during the operation of the contractor
reverse osmosis unit, which is believed to be the source of the groundwater
contamination, one of two leakage paths to groundwater have still not been established.
The established pathway involves leakage from two cut drain lines located above the
floor on the 35-foot elevation of the PAB. Further investigation by Entergy following the
conclusion of the Unit 2 RFO 2R22 must be conducted to verify the second pathway to
groundwater via the floor of the fuel storage building truck bay.
Entergys long-term corrective action for reducing tritium levels in the groundwater is the
same as previously identified for the March 2014 tritium spike (CR-IP2-2015-03806), the
start-up and operation of recovery well (RW)-1. Following installation of equipment and
system testing, full operation of the RW system is expected later this year. This system
will allow for the collection of tritiated groundwater in the vicinity of Unit 2 to be returned
inside the Unit 2 PAB for processing. The NRC will be conducting an inspection in
August 2016 to review the testing plan and results of the RW-1 tests. This inspection
will include a specialist region-based inspector, and a staff hydrogeologist.
The NRCs continuing assessment of the safety significance of this event focused on
validating the safety impact of dose to the public from the release of tritium to the site
groundwater, and ultimately to the Hudson River. The NRC verified that Entergys
bounding public dose calculations on the groundwater contamination leak was
sufficiently conservative and a maximum worst case scenario would result in a dose of
0.000112 millirem per year, which represents a very small fraction of the allowable dose
(liquid effluent dose objective of 3 millirem per year). This low value is due to
groundwater at Indian Point not being a source of any drinking water. There are no
drinking water wells on the Indian Point site, groundwater flow from the site is to the
Hudson River and not to any near site drinking water wells, and the Hudson River has
no downstream drinking water intakes as it is brackish water. Pathways to the public are
therefore limited to the consumption of fish and river invertebrates. The inspection
determined that there is no safety impact to the public as a result of this groundwater
contamination event. (URI 05000247/2016001-07, January 2016 Groundwater
Contamination)
39
.2 Institute of Nuclear Power Operations (INPO) Report Review
a. Inspection Scope
The inspectors also reviewed the final report for the INPO equipment reliability scram
review visit that was conducted to review the scrams that occurred over the past two
years, conducted in June 2016. The inspectors reviewed the report to ensure that any
issues identified were consistent with NRC perspectives of Entergy performance and to
determine if INPO identified any significant safety issues that required further NRC
follow-up.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit
On August 4, 2016, the inspectors presented the inspection results to Mr. Larry Coyle,
Site Vice President, and other members of Entergy. Based on additional information
provided, the inspectors conducted an updated exit meeting on August 30, 2016 with
John Kirkpatrick, Plant Operations General Manager and other members of Entergy.
The inspectors verified that no proprietary information was retained by the inspectors or
documented in this report.
ATTACHMENT: SUPPLEMENTARY INFORMATION
A-1
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Entergy Personnel
A. Vitale, Site Vice President
J. Kirkpatrick, Plant Operations General Manager
R. Alexander, Unit 2 Shift Manager
R. Andersen, Maintenance Instrumentation and Controls Superintendent
N. Azevedo, Engineering Supervisor
J. Baker, Shift Manager
S. Bianco, Operations Fire Marshal
K. Brooks, Assistant Operations Manager
R. Burroni, Engineering Director
T. Chan, Engineering Supervisor
C. Chapin, Training Superintendent
D. Dewey, Assistant Operations Manager
J. Dignam, Unit 3 Control Room Supervisor
R. Dolansky, Inservice Inspection Program Manager
W. Durr, Outage Control Center Manager
R. Drake, Engineering Supervisor
K. Elliott, Fire Protection Engineer
J. Ferrick, Regulatory and Performance Improvement Director
L. Frink, Radiation Protection Supervisor
D. Gagnon, Security Manager
L. Glander, Emergency Preparedness Manager
D. Gray, Radiological Environmental Manager
J. Johnson, Unit 2 Control Room Supervisor
M. Johnson, Unit 3 Shift Manager
M. Khadabux, Instrumentation and Controls Supervisor
F. Kich, Performance Improvement Manager
M. Lewis, Unit 3 Assistant Operations Manager
N. Lizzo, Training Manager
S. McAllister, Baffle Bolt Replacement Project Manager
M. McCarthy, Unit 3 Control Room Supervisor
B. McCarthy, Operations Manager
F. Mitchell, Radiation Protection Manager
E. Mullek, Maintenance Manager
S. Stevens, Radiation Protection Operations Superintendent
B. Sullivan, Training Superintendent
J. Taylor, Unit 3 Shift Manager
M. Tesoriero, Outage Control Center Manager
M. Troy, Nuclear Oversight Manager
R. Walpole, Regulatory Assurance Manager
A. Zastrow, Assistant Operations Manager
Attachment
A-2
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened
05000247/2016002-01 URI CVCS Goal Monitoring Under the Maintenance
Rule (Section 1R12)
Opened/Closed
05000286/2016002-02 NCV Failure to Follow Operability Determination
Procedure for Unit 3 Baffle-Former Bolts
(Section 1R15)05000247/2016002-03 NCV Failure to Maintain Flow Channeling Gates Closed
in Accordance with the Containment Procedure
(Section 1R20)05000247/2016002-04 NCV Failure to Scope Safety-Related Main Boiler
Feedwater Pump Discharge Valves into the
Maintenance Rule Program (Section 4OA3)
Closed
05000247/2015-003-00 LER Manual Reactor Trip due to Indications of Multiple
Dropped Control Rods Caused by Loss of Control
Rod Power Due to a Power Supply Failure
(Section 4OA3)
05000247/2016-003-00 LER Technical Specification Prohibited Condition
Due to an Inoperable 21 Main Boiler Feedwater
Pump Discharge Valve for Greater Than the TS
Allowed Outage Time (Section 4OA3)
Discussed
05000247/2016001-01 URI Baffle-Former Bolts with Identified Anomalies
(Section 1R08)05000247/2016001-06 URI Emergency Diesel Generator Automatic Voltage
Regulator Failure (Section 1R15)05000247/2016001-07 URI January 2016 Groundwater Contamination
Section (Section 4OA5)
A-3
LIST OF DOCUMENTS REVIEWED
Common Documents Used
Indian Point Unit 2 and Unit 3, UFSARs
Indian Point Unit 2 and Unit 3, Individual Plant Examinations
Indian Point Unit 2 and Unit 3, Individual Plant Examination of External Events
Indian Point Unit 2 and Unit 3, TSs and Bases
Indian Point Unit 2 and Unit 3, Technical Requirements Manuals
Indian Point Unit 2 and Unit 3, Control Room Narrative Logs
Indian Point Unit 2 and Unit 3, Plans of the Day
Section 1R04: Equipment Alignment
Procedures
2-COL-4.2.1, Residual Heat Removal System, Revision 30
2-COL-4.3.1, Spent Fuel Pit Cooling, Revision 10
2-COL-24.1.1, Service Water System, Revision 50
3-COL-EL-005, Diesel Generators, Revision 37
OAP-019, Component Verification and System Status Control, Revision 7
OAP-044, Plant Labeling Program, Revision 3
Condition Reports (CR-IP2)
2016-01311 2016-01505 2016-01761 2016-02330 2016-02428 2016-02470
Condition Reports (CR-IP3)
2016-01382 2016-01810
Drawings
209762, Flow Diagram Service Water System Nuclear Steam Supply Plant, Revision 75
227781, Flow Diagram Auxiliary Coolant System, Revision 22
9321-2720, Auxiliary Coolant System, Sheet 2, Revision 22
Miscellaneous
IP3-DBD-308, CCW System, Revision 3
Section 1R05: Fire Protection
Procedures
EN-MA-133, Control of Scaffolding, Revision 12
Condition Reports (CR-IP2)
2016-04148
Condition Reports (CR-IP3)
2016-01272
Miscellaneous
PFP-203, Containment Building, 95-Foot Elevation (Fire Zone 86A), Revision 15
PFP-204, General Floor Plan, PAB, 15-Foot Elevation, Revision 0
PFP-209, Component Cooling Pump Room, PAB, 68-Foot Elevation, Revision 0
PFP-211, General Floor Plan, PAB, 80-Foot Elevation, Revision 14
PFP-351, 480V Switchgear Room, Revision 15
A-4
Section 1R07: Heat Sink Performance
Procedures
0-HTX-405-EDG, EDG Lube Oil and Jacket Water Heat Exchanger Maintenance, Revision 4
Condition Reports (CR-IP3)
2010-02900 2011-03594 2011-03596 2011-03961 2012-02071 2012-03912
2013-02338 2013-02695 2013-03009 2014-00957 2014-01239 2014-03158
2014-03175 2015-00031 2015-00599 2015-02848 2015-05209 2015-05526
2016-00886 2016-00895 2016-00899
Maintenance Orders/Work Orders
Miscellaneous
SEP-SW-IPC-001, Indian Point Energy Center NRC Generic Letter 89-13 Service Water
Program, Revision 0
Section 1R08: Inservice Inspection Activities
Procedures
GBRA-104-659, Collection of Protocols for Baffle Bolt Replacement, Revision C
GBRA-175-115, Field Service Procedure for Baffle Bolt Replacement, Revision 3
WDI-STD-088, Underwater Remote Visual Examination of Reactor Vessel Internals,
Revision 13
WDI-STD-1073, Ultrasonic Test Procedure for the Inspection of Internal Hex Head
Baffle-Former Bolts with Welded Lock Bars, Revision 4
Condition Reports (CR-IP2)
2016-02081
Maintenance Orders/Work Orders
442412-13
Miscellaneous
Indian Point Unit 2 Baffle Bolt Ultrasonic Examination Expanded Analysis Report, dated
April 28, 2016
IP2 Reactor Vessel Visual Examination Report, dated May 2006
Loose Parts Inventory Log for Baffle Bolt Replacements, dated May 24, 2016
MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and
Evaluation Guidelines (ML120170453)
MRP-228, Materials Reliability Program: Inspection Standard for PWR Internals - 2012 Update,
Revision 1
SEP-ISI-IP2-001, IP2 Fourth Ten-Year Interval Inservice Inspection (ISI)/Containment Inservice
Inspection (CISI) Program Plan, Revision 2
WDI-PJF-1315504-EPP-001, Indian Point Nuclear Power Plant MRP-227-A Reactor Vessel
Internals Examination Program Plan, Revision 0
WDI-PJF-1315505-FSR-001, Indian Point Unit 2 2R22 MRP-227-A Baffle-Former Bolt
Ultrasonic Inspections Field Service Report, dated March 29, 2016
WDI-TJ-1100, Technical Justification for the Ultrasonic Inspection of Baffle-Former Bolts for
Indian Point Units 2 and 3, Revision 1
A-5
Section 1R11: Licensed Operator Requalification Program
Procedures
2-AOP-480V-1, Loss of Normal Power to Any 480V Vital Bus, Revision 8
2-AOP-RCP-1, Reactor Coolant Pump Malfunction, Revision 14
2-AOP-TURB-1, Main Turbine Trip without a Reactor Trip, Revision 5
2-E-0, Reactor Trip or Safety Injection, Revision 7
2-FR-H.1, Response to Loss of Secondary Heat Sink, Revision 11
2-POP-1.2, Reactor Startup, Revision 59
2-SOP-26.4, Turbine Generator Startup, Synchronizing, Voltage Control and Shutdown,
Revision 62
3-AOP-480V-1, Loss of Normal Power to Any Safeguards Bus, Revision 7
3-AOP-CVCS-1, Chemical and Volume Control System Malfunction, Revision 8
3-AOP-FW-1, Loss of Feedwater, Revision 8
3-AOP-INST-1, Instrument/Controller Failures, Revision 11
3-E-0, Reactor Trip or Safety Injection, Revision 6
3-E-1, Loss of Reactor or Secondary Coolant, Revision 4
3-FR-C.2, Response to Degraded Core Cooling, Revision 3
Condition Reports (CR-IP2)
2016-03946 2016-04162 2016-04164 2016-04165 2016-04169 2016-04178
Condition Reports (CR-IP3)
2016-01087 2016-01092 2016-01098 2016-01336
Miscellaneous
13SX-LOR-SES026, Licensed Operator Requalification Program Scenario
Emergency Action Level Table, Revision 15.2
LRQ-SES-04, IPEC Simulator Evaluated Scenario, Revision 6
Section 1R12: Maintenance Effectiveness
Procedures
CEP-NDE-0640, Non-Section XI Liquid Penetrant Examination, Revision 9
CEP-WP-WIIR-1, Attachment 5.1, Inprocess Inspections for Installation and Replacement
Welds Located Inside the ASME Section XI Boundary, Revision 3
EN-DC-206, Maintenance Rule (a)(1) Process, Revision 3
Condition Reports (CR-IP2)
2010-00864 2013-03130 2014-00162 2014-00185 2014-01144 2014-02184
2015-00278 2016-01260 2016-01430 2016-01500
Condition Reports (CR-IP3)
2012-03836 2013-04758 2015-01396 2015-03404 2015-03653 2015-04053
2015-04162 2015-04184 2015-04539 2015-05316 2015-05384 2015-05729
A-6
2016-00098 2016-00653 2016-00723 2016-01189 2016-01227 2016-01274
2016-01313 2016-01531 2016-01536 2016-01543 2016-02432
Maintenance Orders/Work Orders
WO 00397793 WO 00408019 WO 00414886 WO 00416091
WO 00421841 WO 00429532 WO 00429532 WO 00431497
WO 00446165 WO 00447042 WO 00447966 WO 52602429
WO 52621178
Miscellaneous
EC 65389, 21 CCW HX Leak Repair: Service Water Inlet Nozzle - Elbow Weld Configuration
Change
IPEC Maintenance Rule Basis Document - Chemical and Volume Control System, Revision 0
PQR 913, 134 F42 MN-GTAW ASME IX Welding Procedure Qualification Record, Revision 0
System Health Report, Unit 3, EDG, Q1-2016
Weld Map Number 447966-20-01, Revision 0
WPS 134 F42, MN-GTAW, ASME Section IX Welding Procedure Specification, Revision 0
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
EN-OP-119, Protected Equipment, Revision 8
IP-SMM-OU-104, Attachment 9.1, Shiftly Outage Shutdown Safety Assessments, Revision 15
IP-SMM-OU-104, Attachment 9.2, Shiftly Outage Shutdown Safety Assessment Guidelines,
Revision 15
Condition Reports (CR-IP2)
2016-04141
Condition Reports (CR-IP3)
2016-01545
Miscellaneous
EOOS Risk Assessment Software Tool
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
2-PC-R3-1, Pressurizer Level Transmitters, Revision 10
3-ARP-010, Metal Impact Monitoring System, Page 10, Revision 32
3-SOP-RCS-016, Operation of the Metal Impact Monitoring System, Revision 8
EN-OP-104, Operability Determination Process, Revision 10
Condition Reports (CR-IP2)
2016-2221 2016-2356 2016-2961 2016-3345 2016-3418 2016-3660
2016-3636 2016-3784 2016-3806 2016-3818 2016-4085
Condition Reports (CR-IP3)
2014-01670 2015-03468
A-7
Maintenance Orders/Work Orders
WO 00327574 WO 00425980 WO 52571030
Miscellaneous
EN-LI-100, Attachment 9.1, Change Channel Check Comparison Criteria/2-PT-M100,
2-PT-D001, Revision 0
Section 1R18: Plant Modifications
Drawings
10073E87-001, Indian Point Unit 2 Baffle Bolt Replacement - Core Barrel and Baffle Assembly
Elevation, Revision 0
10111D06, Indian Point Unit 2 Baffle Bolt Replacement - Replacement Baffle Bolt .625
and .750, Revision 0
Miscellaneous
EC 64308, IP2 Reactor Vessel Equivalent Replacement Baffle-to-Former Bolt, Revision 0
Process Applicability Determination Form for EC 64308, dated April 21, 2016
WCAP-18136-P, Replacement Type 316 Cold-Worked Baffle-Former Bolt Qualification for
Indian Point Unit 2, Revision 0
Section 1R19: Post-Maintenance Testing
Procedures
3-PT-M079B, 32 EDG Functional Test, Revision 52
2-PC-Q109-4, Recalibration of NIS and OT/OP Delta-T Parameters - Channel IV, Revision 44
Condition Reports (CR-IP2)
2016-03961 2016-04266
Condition Reports (CR-IP3)
2016-01189 2016-01199 2016-01218
Maintenance Orders/Work Orders
WO 00414886 WO 00420649 WO 00446094 WO 00447966
WO 52545181 WO 52626563 WO 52626564 WO 52630619
WO 52630620 WO 52658943 WO 00236158 WO 00277374
WO 52571030
Drawings
5651D72, Logic Diagrams Steam Generator Trip Signals, Revision 7
Miscellaneous
EC 64545, Emergency Temporary Modification to Disconnect 32 EDG Generator Space Heater
Adjacent to End Plate on Outboard End of Generator
FIX00091, Pressurizer Level Uncertainty - Indication, Trip Setpoints, and Annunciation
Setpoints, Revision 1
E-mail from J. Michetti to G. Newman, dated July 19, 2016, Subject: Westinghouse Report
on E9
A-8
Section 1R20: Refueling and Other Outage Activities
Procedures
2-POP-1.1, Plant Heatup from Cold Shutdown, Revision 90
2-POP-1.2, Reactor Startup, Revision 59
2-POP-1.3, Plant Startup from Zero to 45 Percent Power, Revision 89
2-POP-3.1, Plant Shutdown from 45 Percent Power, Revision 58
2-POP-3.3, Plant Cooldown, Hot to Cold Shutdown, Revision 81
2-POP-3.4, Secondary Plant Shutdown, Revision 10
2-POP-4.1, Operation at Cold Shutdown, Revision 5
2-POP-4.2, Operation Below 20 Percent Pressurizer Level with Fuel in the Reactor, Revision 8
2-POP-4.3, Operation without Fuel in the Reactor, Revision 1
Condition Reports (CR-IP2-)
2016-04118 2016-04119 2016-04123 2016-03124 2016-04126 2016-04129
2016-04130 2016-04131 2016-04132 2016-04139 2016-04141* 2016-04142*
2016-04144 2016-04145 2016-04146 2016-04148* 2016-04151 2016-04152
2016-04155 2016-04161 2016-04162 2016-04165 2016-04169
- NRC identified
Maintenance Orders/Work Orders
52681465
Miscellaneous
2R22 Performance Indicator Curves Daily from March 7 to June 14, 2016
Outage Schedules and Plans of the Day from March 7 to June 14, 2016
Westinghouse LTR-PL-16-16, Operability Assessment for Primary Side Loose Parts at Indian
Point Unit 2, Revision 0, dated March 27, 2016
Section 1R22: Surveillance Testing
Procedures
0-SOP-LEAKRATE-001, RCS Leakrate Surveillance Evaluation and Leak Identification,
Revision 6
2-PT-D001, Control Room Operations Surveillance Requirements, Revision 16
2-PT-M029B, 22 Safety Injection Pump, Revision 20
2-PT-Q013, Inservice Valve Tests, Revision 51
2-PT-Q013-DS040, Valve 887B Inservice Test Data Sheet, Revision 22
3-PT-M079B, 32 EDG Functional Test, Revision 52
Condition Reports (CR-IP2)
2016-03360 2016-03363
Condition Reports (CR-IP3)
2016-01716 2016-01752
Maintenance Orders/Work Orders
WO 00443040 WO 00446385 WO 00446867 WO 52681652-01
WO 52681646-01
A-9
Miscellaneous
EC 64855, EC Reply to Provide Input Regarding MPR Maintenance Bulletin MB-2007-01 for
Auto Voltage Regulator Solder Joints
MB-2007-01, Potential for Solder Joint Cracks on Basler SBSR Auto Voltage Regulator Cards
and Technical Manual Addendum TM-2007-01, November 5, 2007
Unit 3 RCS Routine Activity Sample, 28-June-16-10006
Section 1EP6: Drill Evaluation
Procedures
IP-EP-120, Emergency Classification, Revision 10
IP-EP-410, Protective Action Recommendations, Revision 11
Section 2RS7: Radiological Environmental Monitoring Program
Procedures
0-CY-1920, REMP Land Use Census, Revision 1
0-CY-1980, Preparation, Placement and Collection of Site Environmental Thermoluminescent
Dosimeters, Revision 2
Condition Reports (CR-IP2)
2014-05319 2015-00948 2015-01300 2015-02687 2015-02800 2015-02987
2015-03271 2015-03396 2016-02313
Condition Reports (CR-IP3)
2016-00514
Miscellaneous
2014 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3
2015 Annual Radiological Environmental Operating Report, Indian Point Nos. 1, 2, and 3
Environmental Dosimetry Company, Annual Quality Assurance Status Report,
January to December 2015
Indian Point Energy Center ODCM, Revision 4
June 2015 to May 2016 Meteorological Data Recovery
Met One Instruments, Inc. Certificates of Calibration for Temperature, Wind Direction, and Wind
Speed
Teledyne Brown Engineering Environmental Services Annual 2015 Quality Assurance Report
Exelon PowerLabs Certificates of Calibration for Gas Meters
3471875 3482909 3471871 3471867 3482920 3471873
3482910 3482916 3471877 3482914 3482918 3482921
3471881 3471879 3471872 3471869 3471880 3482908
Quality Assurance
Quality Assurance Audit Report QA-2-6-2015-IP-1, Chemistry, Effluents, and Environmental
Monitoring Snapshot Self-Assessment, LO-IP3LO-2015-00126, Chemistry-REMP
Section 4OA2: Problem Identification and Resolution
Procedures
EN-DC-204, Maintenance Rule Scope and Basis, Revision 3
EN-DC-204, Maintenance Rule Scope and Basis, Revision 3
EN-DC-207, Maintenance Rule Periodic Assessment, Revision 3
A-10
EN-LI-102, Corrective Action Program, Revision 26
EN-LI-104, Self-Assessment and Benchmark Process, Revision 11
EN-LI-110-01, Equipment Failure Evaluation, Revision 0
EN-LI-119, Apparent Cause Evaluation Process, Revision 11
EN-OP-104, Operability Determination Process, Revision 10
Condition Reports (CR-IP2)
2010-07013 2015-04574 2015-05458 2015-05460 2015-05461 2015-05464
2015-05466 2015-05467 2016-01374 2016-02348
Condition Reports (CR-IP3)
2015-3628 2016-01035 2016-01961
Maintenance Orders/Work Orders
Apparent Cause Evaluations
Drawings
504405, Sheet 1 of 2, Replacement Baffle Bolt Location Matrix, Revision 0
504405, Sheet 2 of 2, Replacement Baffle Bolt Location Matrix, Revision 0
Miscellaneous
61265, Temporary Modification Control Form-Install Additional Temporary Spare Power Supply
Inside Rod Control Cabinet 2BD to Restore Margin, Revision 0
Appendix A to Facility Operating License DPR-64, Technical Specifications and Basis For The
Indian Point 3 Nuclear Generating Station Unit No.3, Through Amendment 260
CN-RIDA-15-43, Indian Point Units 2 and 3 Acceptable Baffle-Former Bolting LOCA and
Seismic Analysis, Revision 2
Engineering Change 63938, As-left condition of the baffle-former plate assembly following the
replacement of degraded bolts, Revision 0
EPRI TR-112209, Analysis of Baffle Former Bolt Cracking in EDF CPO Plants (PWRMRP-03),
dated June 1999
Indian Point Entergy Center (IPEC) Unit 3, Updated Final Safety Analysis Report, dated May
2013
IP-RPT-16-00025, Evaluation of Indian Point Unit 3 Reactor Core Baffle Bolting Following MRP-
227-A Inspection Findings at Indian Point Unit 2 during 2R22, Revision 0
LTR-PL-16-21, Transmittal of Indian Point Unit 3 Final Engineering Evaluations Supporting
Extent of Condition Review, Revision 0
LTR-RIDA-16-103, Indian Point Unit 2 Baffle Bolting Anti-Clustering Pattern and Margin
Assessment, Revision 0
LTR-RIDA-16-152, Indian Point Unit 3 Baffle Bolt Leak Before Break Operability Assessment,
Revision 0
LTR-RIDA-16-60, Indian Point Unit 2 Baffle Bolting One Cycle Replacement Pattern Summary
Letter, Revision 0
MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and
Evaluation Guidelines (ML120170453)
Operation Decision Making Issue Action Plan for IP3 Baffle Bolt Monitoring, dated May 19, 2016
WCAP-15029-NP-A, Westinghouse Methodology for Evaluating the Acceptability of Baffle-
Former-Barrel Bolting Distributions Under Faulted Load Conditions, Revision 0
A-11
WCAP-17949-P, Background and Technical Basis Supporting Engineering Flaw Acceptance
Criteria for Indian Point Unit 2 Reactor Vessel Internals MRP-227-A Primary and
Expansion Components, Revision 1
WCAP-18048-P, Determination of Acceptable Baffle-Former Bolting for Indian Point Units 2 and
3, Revision 0
Section 4OA5: Other Activities
Miscellaneous
INPO Letter, INPO Equipment Reliability Scram Review Visit, May 31, 2016
Root Cause Evaluation for CR-IP2-2016-00564
A-12
LIST OF ACRONYMS
10 CFR Title 10 of the Code of Federal Regulations
ADAMS Agencywide Document Access and Management System
ALARA as low as is reasonably achievable
AVR automatic voltage regulator
CAP corrective action program
CCW component cooling water
CR condition report
CVCS chemical and volume control system
EC engineering change
ECCS emergency core cooling system
EDG emergency diesel generator
GPI groundwater protection initiative
IASCC irradiation-assisted stress-corrosion cracking
IMC Inspection Manual Chapter
INPO Institute of Nuclear Power Operations
LER licensee event report
LOCA loss-of-coolant accident
MBFP main boiler feedwater pump
MCC motor control center
MOV motor operated valve
MRP materials reliability program
MW monitoring well
NCV non-cited violation
NRC Nuclear Regulatory Commission, U.S.
ODCM offsite dose calculation manual
OOS out of service
PAB primary auxiliary building
PFP pre-fire plan
REMP radiological environmental monitoring program
RFO refueling outage
RW recovery well
SI safety injection
SSC structure, system, and component
TS technical specification
UFSAR updated final safety evaluation report
URI unresolved item
UT ultrasonic testing
WO work order