ML16043A391
| ML16043A391 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 02/12/2016 |
| From: | Fred Bower Reactor Projects Branch 3 |
| To: | Dean Curtland NextEra Energy Seabrook |
| bower, fl | |
| References | |
| IR 2015001, IR 2015004 | |
| Download: ML16043A391 (50) | |
See also: IR 05000443/2015004
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
2100 RENAISSANCE BLVD., SUITE 100
KING OF PRUSSIA, PA 19406-2713
February 12, 2016
Mr. Dean Curtland
Site Vice President
Seabrook Nuclear Power Plant
NextEra Energy Seabrook, LLC
c/o Mr. Michael Ossing
P.O. Box 300
Seabrook, NH 03874
SUBJECT: SEABROOK STATION, UNIT NO. 1 - INTEGRATED INSPECTION REPORT
05000443/2015004 AND INDEPENDENT SPENT FUEL STORAGE
INSTALLATION REPORT NO. 07200063/2015001
Dear Mr. Curtland:
On December 31, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at Seabrook Station, Unit No. 1. The enclosed report documents the inspection
results, which were discussed on January 21, 2016, with you and other members of your staff.
NRC Inspectors examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, no findings of significance were identified. However, a
licensee-identified violation, which was determined to be of very low safety significance, is listed
in Section 4OA7 of this report. If you contest the non-cited violations (NCVs) 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, United States Nuclear Regulatory Commission, Washington, DC 20555-
0001; and the NRC Resident Inspector at Seabrook Station.
D. Curtland -2-
In accordance with Title 10 of the Code of Federal Regulations (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/
Fred L. Bower, III, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Docket No. 50-443
License No. NPF-86
Enclosure:
Inspection Report 05000443/2015004
and ISFSI Report 07200063/2015001
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/DRP
NAME PCataldo/ RSB for RBarkley/ RSB FBower/ FLB
DATE 02/11 /16 02/ 11 /16 02/ 11 /16
1
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No: 50-443
License No: NPF-86
Report No: 05000443/2015004
Licensee: NextEra Energy Seabrook, LLC
Facility: Seabrook Station, Unit No.1
Location: Seabrook, New Hampshire 03874
Dates: October 1, 2015 through December 31, 2015
Inspectors: P. Cataldo, Senior Resident Inspector
C. Newport, Resident Inspector
J. Vasquez, Project Engineer
D. Silk, Senior Operations Engineer
B. Dionne, Health Physicist
J. Nicholson, Senior Health Physicist
T. OHara, Reactor Inspector
B. Cook, Senior Reactor Analyst
Approved by: Fred L. Bower, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Enclosure
2
TABLE OF CONTENTS
SUMMARY ................................................................................................................................ 3
1. REACTOR SAFETY ........................................................................................................... 4
1R01 Adverse Weather Protection ..................................................................................... 4
1R04 Equipment Alignment ............................................................................................... 4
1R05 Fire Protection .......................................................................................................... 6
1R07 Heat Sink Performance ............................................................................................ 6
1R08 Inservice Inspection .................................................................................................. 6
1R11 Licensed Operator Requalification Program and Licensed Operator Performance ..10
1R12 Maintenance Effectiveness ......................................................................................12
1R13 Maintenance Risk Assessments and Emergent Work Control .................................13
1R15 Operability Determinations and Functionality Assessments .....................................14
1R18 Plant Modifications ..................................................................................................14
1R19 Post-Maintenance Testing .......................................................................................15
1R20 Refueling and Other Outage Activities .....................................................................15
1R22 Surveillance Testing ................................................................................................17
1EP2 Alert and Notification System Evaluation .................................................................17
1EP3 Emergency Response Organization Staffing and Augmentation System .................17
1EP4 Emergency Action Level and Emergency Plan Changes .........................................18
1EP5 Maintaining Emergency Preparedness ....................................................................18
2. RADIATION SAFETY.........................................................................................................19
2RS1 Radiological Hazard Assessment and Exposure Controls .......................................19
2RS2 Occupational ALARA Planning and Controls ...........................................................20
4. OTHER ACTIVITIES ..........................................................................................................21
4OA1 Performance Indicator Verification ...........................................................................21
4OA2 Problem Identification and Resolution .....................................................................22
4OA5 Other Activities ........................................................................................................28
4OA6 Meetings, Including Exit ...........................................................................................29
4OA7 Licensee-Identified Violation ....................................................................................29
SUPPLEMENTARY INFORMATION....................................................................................... A-1
KEY POINTS OF CONTACT .................................................................................................. A-1
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED .................................... A-1
LIST OF DOCUMENTS REVIEWED....................................................................................... A-2
LIST OF ACRONYMS ........................................................................................................... A-17
3
SUMMARY
Inspection Report 05000443/2015004 and Independent Spent Fuel Storage Installation (ISFSI)
Report 07200063/2015001; 10/01/2015 - 12/31/2015; Seabrook Station, Unit No. 1; Routine
Integrated Inspection Report.
This report covered a three-month period of inspection by resident inspectors and announced
inspections performed by regional inspectors. The NRCs program for overseeing the safe
operation of commercial nuclear power reactors is described in NUREG-1649, Reactor
Oversight Process, Revision 5.
Other Findings
A violation of very low safety significance was identified by NextEra and reviewed by the
inspectors. Corrective actions taken or planned by NextEra have been entered into the
corrective action program (CAP). This violation and corrective action tracking number are listed
in Section 4OA7 of this report.
4
REPORT DETAILS
Summary of Plant Status
Seabrook began the assessment period at less than 20 percent reactor power, continuing the
coastdown for shutdown and entry into Refueling Outage No. 17 (OR17). Seabrook remained
shutdown during performance of OR17, until reactor criticality was achieved on November 12,
2015. Subsequently, Seabrook synchronized to the grid on November 13, 2015, and achieved
full power (100 percent) on November 17, 2015. Seabrook operated at full power for the
remainder of the assessment period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01 - 1 sample)
Readiness for Seasonal Extreme Weather Conditions
a. Inspection Scope
The inspectors reviewed NextEras readiness for the onset of seasonal cold
temperatures. The review focused on the service water (SW) cooling tower (CT) and the
supplemental emergency power system generators, as well as other areas of cold
weather protection for various areas of interest onsite, such as weather enclosures for
the condensate storage tank. The inspectors reviewed the updated final safety analysis
report (UFSAR), technical specifications (TSs), control room logs, and the CAP to
determine what temperatures or other seasonal weather could challenge these systems,
and to ensure NextEra personnel had adequately prepared for these challenges. The
inspectors reviewed station procedures, including NextEras seasonal weather
preparation procedure and applicable operating procedures. The inspectors performed
walkdowns of the selected systems to ensure station personnel identified issues that
could challenge the operability of the systems during cold weather conditions.
Documents reviewed for each section of this report are listed in the Attachment.
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Partial System Walkdowns (71111.04 - 3 samples)
a. Inspection Scope
The inspectors performed partial walkdowns of the following systems:
'A' SW return to service on October 29
'B' residual heat removal (RHR) return to service on November 17
'B' charging system during 'A' train maintenance on December 17
5
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 UFSAR, TSs, work orders
(WOs), condition reports (CRs), and the impact of ongoing work activities on redundant
trains of equipment to identify conditions that could have impacted the systems
performance of its intended safety functions. The inspectors also performed field
walkdowns of accessible portions of the systems to verify system components and
support equipment were aligned correctly and were operable. The inspectors examined
the material condition of the components and observed operating parameters of
equipment to verify that there were no deficiencies. The inspectors also reviewed
whether NextEra staff had properly identified equipment issues and entered them into
the CAP for resolution with the appropriate significance characterization.
b. Findings
No findings were identified.
.2 Full System Walkdown (71111.04S - 1 sample)
a. Inspection Scope
During October and November 2015, the inspectors performed a complete system
walkdown of accessible portions of the primary component cooling water system to
verify the existing equipment lineup was correct. The inspectors reviewed operating
procedures, surveillance tests, system diagrams, equipment line-up check-off lists, WOs,
work requests, CRs, TSs, and the UFSAR to verify the system was aligned to perform its
required safety functions. The inspectors also reviewed electrical power availability,
component lubrication and equipment cooling, hanger and support functionality, and
operability of support systems. The inspectors performed field walkdowns of accessible
portions of the systems to verify system components and support equipment were
aligned correctly and operable. The inspectors examined the material condition of the
components and observed operating parameters of equipment to verify that there were
no deficiencies. The inspectors also reviewed whether NextEra staff had appropriately
evaluated and resolved any equipment issues and other performance deficiencies and
entered them into their CAP for resolution with the appropriate significance
characterization.
b. Findings
No findings were identified.
6
1R05 Fire Protection
Resident Inspector Quarterly Walkdowns (71111.05Q - 5 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
NextEra controlled combustible materials and ignition sources in accordance with
administrative procedures. The inspectors verified that fire protection and suppression
equipment was available for use as specified in the area pre-fire plan, and passive fire
barriers were maintained in good material condition. The inspectors also verified that
station personnel implemented compensatory measures for out of service, degraded, or
inoperable fire protection equipment, as applicable, in accordance with procedures.
Containment 0' (C-F-2-Z) on October 14
Containment -26' (C-F-1-Z) on October 20
Containment 25' (C-F-3-Z) on October 24
Supplement emergency power system (SEPS-F-1-0) on December 3
Essential switchgear 'A' (CB-F-1A-A) on December 17
b. Findings
No findings were identified.
1R07 Heat Sink Performance (711111.07A - 1 sample)
a. Inspection Scope
The inspectors reviewed the B emergency diesel generator (EDG) heat exchanger to
determine its readiness and availability to perform its safety functions. The inspectors
reviewed the design basis for the component and verified NextEras commitments to
NRC Generic Letter 89-13. The inspectors reviewed results of previous inspections of
the B EDG heat exchanger, reviewed performance data obtained during recent
operation of the B EDG heat exchanger, observed the most recent inspection results of
the B EDG heat exchanger, and discussed the results with NextEra staff. The
inspectors also verified that NextEra initiated appropriate corrective actions for identified
deficiencies.
b. Findings
No findings were identified.
1R08 Inservice Inspection (711111.08 - 1 sample)
a. Inspection Scope
The inspectors conducted a review of NextEras implementation of inservice inspection
(ISI) program activities for monitoring degradation of the reactor coolant system (RCS)
boundary, risk significant piping and components, and containment systems during the
7
Seabrook Unit 1, OR17. Inspection sample selection was based on the inspection
procedure objectives and risk priority of those pressure retaining components in systems
where degradation would result in a significant increase in risk. The inspectors observed
in-process non-destructive examinations (NDE), reviewed documentation, and
interviewed NextEra personnel to verify that the NDE examination activities performed
as part of the Seabrook Unit 1 ISI program were being conducted during the Seabrook
Interval 3, Period 2, 2nd Outage, ISI in accordance with the requirements of American
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI,
2004 Edition, no Addenda.
Non-destructive Examination and Welding Activities (Inspection Manual Chapter (IMC)
Section 02.01)
The inspectors performed direct observations of select NDE activities, and reviewed
records of the NDEs listed below. The inspectors reviewed certifications of the NDE
technicians performing the examinations. The inspectors verified that the examinations
were performed in accordance with approved procedures and that the results were
reviewed and evaluated by certified Level III NDE personnel.
ASME Code Required Examinations
- The inspectors observed the remote, bare-metal visual (VT-2) examination of the
reactor vessel upper closure head and control rod drive mechanism (CRDM) nozzle
penetrations.
- The inspectors reviewed the VT-3 direct visual examination data from NextEras
examination of the CRDM structural supports to determine whether all results were
satisfactory.
- The inspectors reviewed the examination data sheets for the post mechanical stress
improvement process phased array, outside diameter ultrasonic test inspections and
the eddy current (EC), inside weld surface inspections of seven reactor vessel hot
leg and cold leg nozzle dissimilar metal safe-end welds. The inspectors verified the
examinations were completed in accordance with the applicable Seabrook TSs,
ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition, Title 10 of the
Code of Federal Regulations (10 CFR) 50.55a(b)(2)(xiv, xv and xiv), Code Case N-
770-1 and the Examination Program Plan (Scan Plan) WDI-PJF-1312142-EPP-001,
Revision 0, as submitted to, and reviewed by, the NextEra staff. The inspectors
determined whether the examination results were dispositioned in accordance with
the applicable requirements for the weld records identified in the list of documents
reviewed section of this report.
- The inspectors reviewed the data records from the visual examinations of the
containment liner concrete pad and floor plugs. The inspectors determined whether
NextEra personnel documented the conditions for investigation, repair, and
monitoring in accordance with their program requirements.
- The inspectors observed the ultrasonic test examination of two ASME Class 2
feedwater nozzle-to-pipe welds (FW 4607-04-19 and FW 4607-04-20) to determine
whether the examinations were performed to check for indications of thermal fatigue
8
in the welds and whether the results were properly dispositioned. No new indications
were detected, and previously detected indications were confirmed to show that no
changes had occurred.
- The inspectors reviewed the documentation for a containment liner anomaly on
the inside surface of the containment dome. NextEra staff completed a VT-1 remote,
visual examination on October 2, 2015. This indication, identified as
IWE-15-159-176, was first documented in October 2009, and re-examined in
October 2012 and April 2014. NextEra engineering staff evaluated the results and
did not identify measureable changes in the length, width, contour or color of the
anomaly from 2009 thru 2015. The inspectors noted that NextEra engineers were
not able to accurately measure the height of the anomaly due to limitations of their
imaging equipment. NextEra staff initiated plans to perform an examination during
the next refueling outage. The inspectors determined NextEra staff tracked and
dispositioned this indication in accordance with their program requirements.
Review of Originally Rejectable Indications Accepted by Evaluation
The inspectors determined there were no samples available for review during this
inspection that involved examinations with recordable indications that had been
accepted for continued service after evaluation.
Repair/Replacement Consisting of Welding Activities
The inspectors reviewed WO 4036801501, which provided direction to fabricate a diesel
fuel system connection to implement design change EC282581. The inspectors
reviewed the WO to determine whether welding was completed in accordance with
ASME Code welding requirements. Specifically, the inspectors determined whether
approved plant welding procedures and parameters were specified and that post-weld
quality control examinations were identified. The inspectors verified that the WO
contained completed weld traveler information, documentation of completed
examinations, and appropriate documentation of the work steps completed by NextEra
personnel.
Pressurized Water Reactor Vessel Lower Head Penetration Inspection Activities (IMC
Section 02.02)
The inspectors reviewed visual inspection records of the bare-metal visual examination
(VT-2) of the exterior surface of the Unit 1 lower reactor vessel head surface and
instrument penetrations for indications of boric acid leakage in order to confirm
appropriate inspection coverage was achieved and to verify that no boric acid leakage or
wastage had been observed. The inspectors further verified that NextEra staff took
actions, tracked in their CAP, to ensure the lower head surface was sufficiently clean to
accurately observe the annulus between the penetrations and the openings in the lower
head. The inspectors reviewed the examination records to verify that there was no
evidence of boric acid.
9
Steam Generator (SG) Tube Inspection Activities (IMC Section 02.04)
The inspectors reviewed the NextEra degradation assessment report for OR17, and
reviewed records associated with a sample of SG tube examinations completed from all
four SGs. The inspectors determined the following activities were planned:
1. Base scope of 100 percent full length bobbin coil examinations in all SGs to inspect
general tube conditions;
2. Hot and cold leg tube support plate to tube intersections examinations in high stress
tubes using a + Point probe to interrogate the regions with the potential for outside
diameter stress corrosion cracking due to tube bulges and tube expansion transitions
near the top of the tube sheet;
3. Examinations of all SGs utilizing a + Point probe, which cannot be inspected with
bobbin probes, such as small radius U-bends and local profile anomalies;
4. Inspection of the tube sheet region with +Point probe in each SG to detect tube
cracking near the top of the tube sheet; and
5. Examination of large residual noise tube support plate locations on high-stress tubes
(hot leg tube support plates only) using a + Point probe in regions with the potential
for outside diameter stress corrosion cracking and primary water stress corrosion
cracking, as well as regions not inspectable with bobbin probes, such as small radius
U-bends and local profile anomalies.
The inspectors reviewed the scope of the EC examinations to verify that known and
potential areas of tube degradation were inspected. The inspectors also verified that
examination scope expansion criteria were implemented based upon inspection results,
as directed by the Electric Power Research Institute (EPRI) Pressurized Water Reactor
Steam Generator Examination Guidelines. The inspectors verified compliance of these
examinations with the Seabrook TSs, the ASME Code,Section XI, and Nuclear Energy
Institute (NEI) 97-06, Steam Generator Program Guidelines. The inspectors verified
tubes with identified wall wear of 40 percent or greater were plugged.
The inspectors reviewed EC examination records and determined NextEra staff
identified a single axial indication in SG A, row 20, column 80 that was subsequently
plugged in accordance with program requirements. NextEra staff review of EC
examination results also identified a defect, due to anti-vibration bar wear in SG A, row
50, column 79, which was plugged during the refueling outage to remove it from service.
EC examination results further showed a single axial indication in SG C, row 39,
column 46, below the top of the tube sheet in the hot leg, which was plugged in
accordance with program requirements. The inspectors determined NextEra staff
developed and completed an expansion plan for all four SGs without discovery of further
indications. The inspectors verified that NextEra staff completed condition monitoring
evaluations for these indications to properly determine, in accordance with program
requirements, that in-situ pressure testing was not required in these tubes.
10
Boric Acid Corrosion Control Inspection Activities (IMC Section 02.03)
The inspectors reviewed the boric acid corrosion control program to determine whether
activities were completed in accordance with NextEra procedures. The inspectors
discussed the program with the boric acid program owner, and sampled photographic
examination records of boric acid found on safety significant piping and components
inside the containment structure during Mode 3 walkdowns conducted by NextEra
personnel and directly observed by the NRC Resident Inspectors on October 2, 2015.
The inspectors reviewed records related to the identification and documentation of non-
conforming conditions of boric acid leaks in the CAP with a focus on areas that could
cause degradation of safety significant components.
The inspectors verified the adequacy of VT-2 visual examination results of the bare-
metal examination of the Unit 1 reactor pressure vessel lower head penetration nozzle
welds that was performed by NextEra personnel during OR17. The inspectors reviewed
a sample of photographs and visual inspection documentation records to verify whether
indications of boric acid leakage were identified.
The inspectors verified that potential deficiencies identified were entered into the CAP
and reviewed evaluations of selected deficiencies documented in CRs to verify that the
corrective actions were consistent with the requirements of the ASME Code and 10 CFR
50, Appendix B, Criterion XVI. The inspectors also reviewed a sample of engineering
evaluations for the above CRs to verify that equipment or components that were wetted
or impinged upon by boric acid solutions were properly analyzed for degradation that
might impact their function.
Identification and Resolution of Problems (IMC Section 02.05)
The inspectors reviewed a sample of CRs since the previous refueling outage, and
verified that non-conforming conditions associated with ISI and SG inspections were
properly identified, evaluated, dispositioned, and appropriately entered into the CAP.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
(71111.11Q - 2 samples; 71111.11B - 1 sample)
.1 Quarterly Review of Licensed Operator Requalification Testing and Training
a. Inspection Scope
The inspectors reviewed the implementation of just-in-time training for licensed
operators on September 30 through October 1, 2015. This training included the
verification of plant shutdown activities consistent with the training plan and associated
objectives for the shutdown and entry into OR17.
b. Findings
No findings were identified.
11
.2 Quarterly Review of Licensed Operator Performance in the Main Control Room
a. Inspection Scope
The inspectors observed infrequently performed test or evolution briefings, pre-shift
briefings, and reactivity control briefings to verify that the briefings met the criteria
specified in NextEras Administrative Procedure OP-AA-100-1000, Conduct of
Operations, Revision 16. In particular, the inspectors observed turbine overspeed
testing, power reduction, reactivity management, and reactor shutdown and entry into
OR17 on October 1, 2015. In addition, inspectors observed transition to shutdown
cooling and implementation of alarm response procedures on October 1, 2015, and RCS
drain-down to mid-loop activities on November 1, 2015. The inspectors observed test
performance to verify that procedure use, crew communications, and coordination of
activities between work groups similarly met established expectations and standards.
b. Findings
No findings were identified.
.3 Licensed Operator Requalification Biennial Review
a. Inspection Scope
The following inspection activities were performed using NUREG-1021, "Operator
Licensing Examination Standards for Power Reactors," Revision 10, and Inspection
Procedure Attachment 71111.11B, Licensed Operator Requalification Program.
Examination Results
On December 29, 2015, the results of the annual operating tests were reviewed in-office
to determine if pass/fail rates were consistent with the guidance of NUREG-1021,
"Operator Licensing Examination Standards for Power Reactors," Revision 10, and NRC
Manual Chapter 0609, Appendix I, Operator Requalification Human Performance
Significance Determination Process (SDP). The review verified that the failure rates
(individual or crew) did not exceed 20 percent.
Five out of 52 operators failed at least one section of the annual exam. The overall
individual failure rate was 9.6 percent.
One out of 11 crews failed the simulator test. The crew failure rate was 9.1 percent.
Written Examination Quality
The inspectors reviewed two written examinations administered during the 2015
examination cycle for qualitative and quantitative attributes as specified in Appendix B of
Attachment 71111.11B, Licensed Operator Requalification Program.
Operating Test Quality
Ten job performance measures (JPMs) and four dynamic scenarios were reviewed for
qualitative and quantitative attributes as specified in Appendix C of 71111.11B,
Licensed Operator Requalification Program.
12
Licensee Administration of Operating Tests
The dynamic simulator exams and JPMs administered during the week of November 30,
2015, were observed. These observations included facility evaluations of Shift Crew B
and Staff Crew Four during two dynamic simulator exams and individual performance of
five JPMs.
Examination Security
The inspectors assessed the facility staffs handling of exam material. The inspectors
also checked JPMs, scenarios, and written examinations for excessive overlap of test
items from week to week.
Remedial Training and Re-Examinations
The remediation plans for two individual JPM failures from the 2014 segment 14F annual
requalification exam, and one biennial comprehensive written exam failure from
November 2013, were reviewed to assess the effectiveness of the remedial training.
Conformance with Operator License Conditions
Medical records for ten license holders were reviewed to assess conformance with
license conditions. Proficiency watch standing records were reviewed for the second
and third quarters of 2015. The reactivation plans for license holders in 2015 were
reviewed to assess the effectiveness of the reactivation process.
Simulator Performance
Simulator performance and fidelity was reviewed for conformance to the reference plant
control room. A sample of simulator deficiency reports was also reviewed to ensure
facility staff addressed identified modeling problems. Simulator test documentation was
also reviewed.
Problem Identification and Resolution
A review was conducted of recent operating history documentation found in inspection
reports, NextEras CAP, and the most recent NRC plant issues matrix. The inspectors
also reviewed specific events from NextEras CAP which, indicated possible training
deficiencies, to verify that they had been appropriately addressed. The NRC resident
inspectors were also consulted for insights regarding licensed operators performance.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12Q - 3 samples)
a. Inspection Scope
The inspectors reviewed the samples listed below to assess the effectiveness of
maintenance activities on structure, system, and component (SSC) performance and
13
reliability. The inspectors reviewed system health reports, CAP documents,
maintenance WOs, and maintenance rule (MR) basis documents to ensure that NextEra
was identifying and properly evaluating performance problems within the scope of the
MR. For each sample selected, the inspectors verified that the SSC was properly
scoped into the MR in accordance with 10 CFR 50.65 and verified that the (a)(2)
performance criteria established by NextEra staff were 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 NextEra
staff was identifying and addressing common cause failures that occurred within and
across MR system boundaries.
Radiation monitors
Various safety-related snubber maintenance and leak resolutions
SW CT fan electrical and mechanical maintenance activities
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 3 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 NextEra 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 NextEra
personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the
assessments were accurate and complete. When NextEra 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.
RCS drain-down to the reactor vessel flange on October 5
A SW system outage during core offload period (i.e., out-of-service) on October 8
RCS decreased inventory and mid-loop operations on October 30
b. Findings
No findings were identified.
14
1R15 Operability Determinations and Functionality Assessments (71111.15 - 4 samples)
a. Inspection Scope
The inspectors reviewed operability determinations for the following degraded or non-
conforming conditions based on the risk significance of the associated components and
systems:
Feed regulating valve localized flow erosion on October 14
Foreign material (i.e., ball bearings) identified in the reactor vessel and system piping
from an ultrasonic inspection tool on October 23
Containment liner leakage rate floor plugs and channel inspections on October 28
Through wall leak on pipe threads of MS-V-413 drain line on December 21
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 NextEras evaluations to determine whether the components or
systems were operable. The inspectors confirmed, where appropriate, compliance with
bounding limitations associated with the evaluations.
b. Findings
No findings were identified.
1R18 Plant Modifications (71111.18 - 3 samples)
.1 Permanent Modifications
a. Inspection Scope
The inspectors reviewed the permanent modifications listed below. This included two
modifications associated with NextEras response to a Japan Lessons Learned Order.
The inspection did not address whether the modification satisfactorily addressed the
objectives of Order EA-12-049. The inspection scope for the modification was restricted
to those elements necessary to satisfy the stated objectives of Inspection Procedure
71111.18, specifically: 1) to verify that the modifications have not affected the safety
functions of important safety systems; 2) to verify that the design bases, licensing bases,
and performance capability of risk significant SSCs have not been degraded through
modifications; and, 3) to verify that modifications performed during increased risk-
significant configurations did not place the plant in an unsafe condition. The inspectors
reviewed modification documents that included applicable 10 CFR 50.59 documentation,
post-modification testing, and design interface activities, such as procedure changes and
applicable training activities.
EC282453, replacement of MS-V-393
EC282474, installation of reactor coolant pump low leakage shutdown seals
EC282580, Fukushima FLEX strategy RCS makeup
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b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19 - 5 samples)
a. Inspection Scope
The inspectors reviewed the post-maintenance tests for the maintenance activities listed
below to verify that procedures and test activities adequately tested the safety functions
that may have been affected by the maintenance activity, that the acceptance criteria in
the procedure were 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, and 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.
Reactor trip breaker inspection and testing on October 1 and November 4
B RHR pump retest following outage maintenance activities on October 26
Emergency feedwater pump P-37A retest following outage maintenance activities on
November 12
10 CFR Part 21, primary/secondary trip latch replacement on ABB Model 7 K-Line
breaker for CT fan, 2-SW-FN-51B, on December 1
'D' vital battery cell #18 replacement on December 30
b. Findings
No findings were identified.
1R20 Refueling and Other Outage Activities (71111.20 - 1 sample)
a. Inspection Scope
The inspectors reviewed the stations work schedule and outage risk plan for the
maintenance and refueling outage (OR17), conducted October 1 through November 13.
The inspectors reviewed NextEras 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
operator performance and other attributes associated with portions of the shutdown and
cooldown processes, and compliance with cooldown rates associated with TS. Further,
the inspectors reviewed multiple surveillance and other critical evolutions in the control
room, and monitored controls associated with the following outage activities:
Configuration, risk, and outage management, including monitoring of key shutdown
safety functions, and compliance with the applicable TSs when taking equipment out
of service;
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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;
Configuration and appropriate monitoring of reactor coolant level and temperature
instruments, particularly during activities associated with the highest level of
shutdown risk, which occurred during the RCS drain-down and reactor vessel head
de-tensioning activities, as well as the drain-down to support RCS evacuation and fill;
Status and configuration of electrical systems and switchyard activities to ensure that
TSs were met; activities associated with the installation and testing of the new
generator step-up transformers, including monitoring systems for combustible gas
generation and induced currents from solar magnetic disturbances;
Monitoring of decay heat removal operations, during initial onset into shutdown
cooling, as well as activities associated with spent fuel pool cooling following full core
offload;
Observed various stages of NextEras management and design implementation of
post-Fukushima beyond design basis mitigating strategies and to verify no impacts
on systems important to safety for both operating and shutdown conditions;
Reactor water makeup and inventory controls, including appropriate flow paths,
configurations, alternative means for inventory additions, and controls to prevent
inventory loss;
Activities that could affect reactivity;
Industrial safety activities, including equipment heavy lifts consistent with the
Memorandum of Understanding with the Occupational Safety and Health
Administration;
Refueling activities, including fuel handling during core offload and reload inside
containment, fuel handling in the spent fuel pool, as well as pre-outage fuel receipt
inspections;
Fatigue management that involved covered workers, and review of work hour
controls and waivers;
Prioritization and completion of mode hold condition reports and work orders, and
review of operating mode transition checklists;
Performed a final containment closeout/walk-down to verify that debris or equipment
had not been left inside, particularly in areas that could impact operability of the
containment recirculation sumps; additionally, performed condition assessment of
debris interceptor screens, scuppers and doorways that contribute to overall
operability of the containment sumps and emergency core cooling systems and
component;
Reactor start-up, plant heat-up, and power ascension activities; and
Problem identification and resolution actions related to OR17 refueling outage
activities.
b. Findings
No findings were identified.
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1R22 Surveillance Testing (71111.22 - 2 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 NextEra 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:
Refueling water storage tank to charging pump suction isolation valve, 1-CS-LCV-
112E, inservice test surveillance on October 8 (inservice test)
CBS-V-8 containment isolation valve equipment qualification and operability testing
on October 11 (containment isolation valve test)
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP2 Alert and Notification System Evaluation (71114.02 - 1 sample)
a. Inspection Scope
An onsite review was conducted to assess the maintenance and testing of the Seabrook
Station Alert and Notification System (ANS). During this inspection, the inspectors
conducted a review of the ANS testing and maintenance programs. The inspectors
reviewed the associated ANS procedures and the Federal Emergency Management
Agency approved ANS Design Report to ensure compliance with design report
commitments for system maintenance and testing. The inspectors toured the facility
used by, and interviewed, the dedicated Seabrook staff responsible for the maintenance
and testing of the Seabrook ANS. Title 10 CFR 50.47(b)(5) and the related
requirements of 10 CFR Part 50, Appendix E, were used as reference criteria.
b. Findings
No findings were identified.
1EP3 Emergency Response Organization Staffing and Augmentation System
(71114.03 - 1 sample)
a. Inspection Scope
The inspectors conducted a review of the Seabrook Station Emergency Response
Organization (ERO) augmentation staffing requirements and the process for notifying
and augmenting the ERO. The review was performed to verify the readiness of key
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licensee staff to respond to an emergency event and to verify NextEras ability to
activate their emergency response facilities (ERFs) in a timely manner. The inspectors
reviewed the Seabrook Station Emergency Plan for ERF activation and ERO staffing
requirements, the ERO duty roster, applicable station procedures, augmentation test
reports, the most recent drive-in drill report, and corrective action reports related to this
inspection area. The inspectors also reviewed a sample of ERO responder training
records to verify training and qualifications were up to date. Title 10 CFR 50.47(b)(2)
and related requirements of 10 CFR Part 50, Appendix E, were used as reference
criteria.
b. Findings
No findings were identified.
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04 - 1 sample)
a. Inspection Scope
NextEra implemented various changes to the Seabrook Emergency Action Levels
(EALs), Emergency Plan, and Implementing Procedures. NextEra had determined that,
in accordance with 10 CFR 50.54(q)(3), any change made to the EALs, emergency plan,
and its lower-tier implementing procedures had not resulted in any reduction in
effectiveness of the Plan, and that the revised Plan continued to meet the standards in
50.47(b) and the requirements of 10 CFR Part 50 Appendix E.
The inspectors performed an in-office review of all EAL and emergency plan changes
submitted by NextEra as required by 10 CFR 50.54(q)(5), including the changes to
lower-tier emergency plan implementing procedures, to evaluate for any potential
reductions in effectiveness of the emergency plan. This review by the inspectors was
not documented in an NRC Safety Evaluation Report and does not constitute formal
NRC approval of the changes. Therefore, these changes remain subject to future NRC
inspection in their entirety. The requirements in 10 CFR 50.54(q) were used as
reference criteria.
b. Findings
No findings were identified.
1EP5 Maintaining Emergency Preparedness (71114.05 - 1 sample)
a. Inspection Scope
The inspectors reviewed a number of activities to evaluate the efficacy of NextEras
efforts to maintain the Seabrook emergency preparedness (EP) program. The
inspectors reviewed: Memorandums of Understanding with offsite agencies; the 10 CFR
50.54(q) emergency plan change process and practices; NextEra maintenance of
equipment important to EP; records of evacuation time estimate population evaluation;
and provisions for, and implementation of, primary, backup, and alternate emergency
response facility maintenance. The inspectors also verified NextEras compliance at
Seabrook with new NRC EP regulations regarding: EALs for hostile action events;
protective actions for on-site personnel during events; emergency declaration timeliness;
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ERO augmentation and alternate facility capability; evacuation time estimate updates;
on-shift ERO staffing analysis; and, ANS back-up means.
The inspectors further evaluated NextEras ability to maintain their EP program through
their identification and correction of EP weaknesses, by reviewing a sample of drill
reports, actual event reports, self-assessments, 10 CFR 50.54(t) review reports, and
EP-related CRs. The inspectors reviewed a sample of EP-related CRs initiated at
Seabrook from July 2013 through December 2015. The inspection was conducted in
accordance with NRC Inspection Procedure 71114.05. Title 10 CFR 50.47(b) and the
related requirements of 10 CFR Part 50, Appendix E, were used as reference criteria.
b. Findings
No findings were identified.
2. RADIATION SAFETY
Cornerstone: Public Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01 - 1 sample)
a. Inspection Scope
The inspectors reviewed NextEras performance in assessing and controlling radiological
hazards in the workplace. The inspectors used the requirements contained in 10 CFR
20, TSs, applicable Regulatory Guides (RGs), and the procedures required by TSs as
criteria for determining compliance.
Inspection Planning
The inspectors reviewed the performance indicators (PIs) for the occupational exposure
cornerstone, radiation protection (RP) program audits, and reports of operational
occurrences in occupational radiation safety since the last inspection.
Radiological Hazard Assessment
The inspectors reviewed recent plant radiation surveys and any changes to plant
operations since the last inspection to identify any new radiological hazards for onsite
workers or members of the public.
Instructions to Workers
The inspectors observed several containers of radioactive materials and assessed
whether the containers were labeled and controlled in accordance with requirements.
The inspectors reviewed several occurrences where a workers electronic personal
dosimeter alarmed. The inspectors reviewed NextEras evaluation of the incidents,
documentation in the CAP, and whether compensatory dose evaluations were
conducted when appropriate.
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Contamination and Radioactive Material Control
The inspectors observed the monitoring of potentially contaminated material leaving the
radiological control area and inspected the methods and radiation monitoring
instrumentation used for control, survey, and release of that material. The inspectors
selected several sealed sources from inventory records and assessed whether the
sources were accounted for and were tested for loose surface contamination. The
inspectors evaluated whether any recent transactions involving nationally tracked
sources were reported in accordance with requirements.
Radiological Hazards Control and Work Coverage
The inspectors assessed whether posted surveys, radiation work permits, worker
radiological briefings, the use of continuous air monitoring, and dosimetry monitoring
were consistent with the present plant conditions. The inspectors examined the control
of highly activated or contaminated materials stored within the spent fuel pool and the
posting and physical controls for selected high radiation areas, locked high radiation
areas, and very high radiation areas to verify conformance with the occupational PI.
b. Findings
No findings were identified.
2RS2 Occupational ALARA Planning and Controls (71124.02)
a. Inspection Scope
The inspectors assessed performance with respect to maintaining occupational
individual and collective radiation exposures as low as is reasonably achievable
(ALARA). The inspectors used the requirements in 10 CFR Part 20, RG 8.8, RG 8.10,
TSs, and NextEras procedures required by TSs as criteria for determining compliance.
Radiological Work Planning
The inspectors compared the results achieved (dose rate reductions, actual dose) with
the intended dose established in NextEras ALARA planning for selected high-dose work
activities. The inspectors compared the person-hour estimates provided by maintenance
planning and other groups to the RP group actual person-hours for the work activity, and
evaluated the accuracy of these time estimates. The inspectors assessed the reasons
for any inconsistencies between intended and actual work activity doses.
The inspectors determined whether post-job reviews were conducted to identify lessons
learned. If problems were identified, the inspectors verified that worker suggestions for
improving dose and contamination reduction techniques were entered into NextEras
CAP.
b. Findings
No findings were identified.
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4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
.1 Mitigating Systems Performance Index (3 samples)
a. Inspection Scope
The inspectors reviewed NextEras submittal of the Mitigating Systems Performance
Index for the following systems for the period of October 1, 2014, through September 30,
2015:
Safety System Functional Failures (MS05)
RHR System (MS09)
Cooling Water System (MS10)
To determine the accuracy of the PI data reported during those periods, the inspectors
used definitions and guidance contained in NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 7. The inspectors also
reviewed NextEras operator narrative logs, mitigating systems performance index
derivation reports and basis documents, event reports, and NRC integrated inspection
reports to validate the accuracy of the submittals.
b. Findings
No findings were identified.
.2 Occupational Exposure Control Effectiveness (1 sample)
a. Inspection Scope
The inspectors reviewed NextEra submittals for the occupational radiological
occurrences PI for the fourth quarter 2014 through the third quarter 2015. The
inspectors used PI definitions and guidance contained in NEI Document 99-02, Revision
7, to determine the accuracy of the PI data reported. The inspectors reviewed NextEras
electronic personal dosimetry accumulated dose alarms, dose reports, and dose
assignments for any intakes that occurred during the time period reviewed, to determine
if there were potentially unrecognized PI occurrences. The inspectors conducted
walkdowns of various locked high radiation area entrances to determine the adequacy of
the controls in place for these areas.
b. Findings
No findings were identified.
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.3 Radiological Effluent TS/Offsite Dose Calculation Manual Radiological Effluent
Occurrences (1 sample)
a. Inspection Scope
The inspectors reviewed licensee submittals for the radiological effluent TS/Offsite Dose
Calculation Manual radiological effluent occurrences PI for the fourth quarter 2014
through the third quarter 2015. The inspectors used PI definitions and guidance
contained in NEI Document 99-02, Revision 7, to determine if the PI data was reported
properly. The inspectors reviewed NextEras public dose assessments for the PI for
public radiation safety to determine if related data was accurately calculated and
reported.
The inspectors reviewed the CAP database to identify any potential occurrences such as
unmonitored, uncontrolled, or improperly calculated effluent releases that may have
impacted offsite dose. The inspectors reviewed gaseous and liquid effluent summary
data and the results of associated offsite dose calculations to determine if indicator
results were accurately reported.
b. Findings
No findings were identified.
.4 Emergency Preparedness Performance Indicators (3 samples)
a. Inspection Scope
The inspectors reviewed data for the following three EP PIs: (1) Drill and Exercise
Performance; (2) ERO Drill Participation; and, (3) ANS Reliability. The last NRC EP
inspection at Seabrook was conducted in the second calendar quarter of 2012.
Therefore, the inspectors reviewed supporting documentation from EP drills and
equipment tests from the fourth calendar quarter of 2014 through the third calendar
quarter of 2015 to verify the accuracy of the reported PI data. The review of the PIs was
conducted in accordance with NRC Inspection Procedure 71151. The acceptance
criteria documented in NEI 99-02, Regulatory Assessment Performance Indicator
Guidelines, Revision 7, was used as reference criteria.
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 NextEra entered issues into the CAP at an appropriate threshold,
gave adequate attention to timely corrective actions, and identified and addressed
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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, NextEra performed an evaluation in
accordance with 10 CFR Part 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 NextEra
outside of the CAP, such as trend reports, PIs, major equipment problem lists, system
health reports, MR assessments, and maintenance or CAP backlogs. The inspectors
also reviewed NextEras CAP database for the third and fourth quarters of 2015, 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 Seabrook Stations Self-Evaluation and
Trending Analysis report for the second quarter of 2015, conducted under PI-AA-207-
1000, Station Self-Evaluation and Trending Analysis, Revision 5, to verify that NextEra
personnel were appropriately evaluating and trending adverse conditions in accordance
with applicable procedures.
The inspectors evaluated a sample of departments that are required to provide input into
the quarterly trend reports, which included the operations and maintenance
departments. This review included a sample of issues and events that occurred over the
course of the past two quarters to objectively determine whether issues were
appropriately considered or determined to be emerging or adverse trends, and in some
cases, verified the appropriate disposition of resolved trends. The inspectors verified
that these issues were addressed within the scope of the CAP, or through department
review and documentation in the quarterly trend report for overall assessment.
b. Findings and Observations
No findings were identified.
The inspectors noted that on occasion, potential adverse trends were identified through:
(1) the use of statistical tools available to staff and utilized throughout the quarter to
identify statistically significant issues that reach a predetermined threshold, or (2)
cognitive trends by staff or collectively during review by the Management Review
Committee (MRC) while screening action requests (ARs). In general, the inspectors
noted that new and existing adverse trends, as well as management awareness areas,
were consistent with those identified by the NRC through daily CR reviews, including
those trends identified as cognitive trends during MRC reviews. For example, the
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inspectors had identified several issues associated with Maintenance Department
rework, and noted that two quick hit self-assessments were independently initiated by
maintenance personnel to evaluate potential causes of the rework.
.3 Annual Sample: Transmitter Failures
a. Inspection Scope
The inspectors performed an in-depth review of NextEras evaluations and corrective
actions associated with multiple transmitter spurious indications and failures occurring on
site. The inspectors assessed NextEras problem identification threshold, problem
analysis, extent of condition reviews, compensatory actions, and the prioritization and
timeliness of corrective actions to determine whether NextEra was appropriately
identifying, characterizing, and correcting problems associated with these transmitter
issues and whether the planned or completed corrective actions were appropriate. The
inspectors compared the actions taken to the requirements of NextEras CAP, licensing
basis, and other requirements. The inspectors interviewed engineering and operations
personnel to assess the effectiveness of the implemented corrective actions,
reasonableness of the planned corrective actions, and to evaluate the extent of any on-
going problems.
The inspectors evaluated seven ARs (see below) documenting various transmitter
malfunctions and failures. The transmitter issues spanned multiple systems and
occurred over a timeframe of approximately 18 months. The review focused on various
aspects of the corrective actions and operability process to assess the effectiveness and
scope of corrective actions, including trending, and the bases and supporting information
for immediate operability evaluations, as applicable. Associated documentation and
bases, as well as the MRCs review of operability statements and actions to address
weaknesses were assessed as well. The ARs that were reviewed included:
Volume control tank level transmitter spurious indication (AR 01972065)
Containment building sump level transmitter spiking (AR 01999679)
B EDG SW jacket water cooling flow transmitter out of tolerance low (AR 01995588)
Thermal barrier heat exchanger head pipe level transmitter erratic (AR 02002668)
A EDG SW jacket water cooling flow transmitter discolored water during calibration
Volume control tank level deviations (AR 01991967)
B reactor coolant pump hi-range leakage flow transmitter noisy condition
b. Findings and Observations
No findings were identified.
Based on the documents reviewed and discussions with NextEra personnel, the
inspectors determined that NextEras responses to the identified transmitter issues were
commensurate with their safety significance, and that the actions completed and
planned were reasonable to address the issues identified.
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.4 Annual Sample: Motor Control Center Breaker Failure to Trip
a. Inspection Scope
The inspectors performed an in-depth review of NextEras corrective actions associated
with CR 1998220, Breaker at 1-EDE-MCC-512 for 1-AS-V-175 Would Not Close.
Specifically, a breaker failed to reset several times following manual trip actions during
the performance of breaker preventive maintenance.
The inspectors assessed NextEras problem identification threshold, and applicable
cause analyses, extent of condition reviews, and compensatory actions, as well as the
prioritization and timeliness of corrective actions to determine whether NextEra 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 NextEras CAP and 10
CFR 50, Appendix B. In addition, the inspectors performed field walkdowns and
interviewed appropriate personnel to assess the effectiveness of the implemented
corrective actions.
b. Findings and Observations
No findings were identified.
NextEra originally suspected a well-known industry phenomenon of hardened grease as
the cause of the breaker to not operate as expected during implementation of breaker
preventive maintenance in October 2014, under WO 40274287. The inspectors noted
that the preventive maintenance included an activity to assess the quality of the grease
consistent with the grade characteristics of the National Lubricating Grease Institute, and
LS 0569.20, Lubrication PM and Starter Inspection for Motor Operated Valve
Actuators, Revision 10. The as-found grease condition of Grade 2 was consistent with
grease that is new, and therefore, was not the probable cause of the breaker failure to
operate. Subsequent inspections resulted in the replacement of a unitized starter
associated with the breaker circuitry. Following successful replacement of the starter,
the post-maintenance testing was performed and the breaker operated as designed.
.5 Annual Sample: Review of Corrective Actions for Alkali-Silica Reaction (ASR) Affected
Structures
a. Inspection Scope
During this inspection period, a region-based inspector accompanied Office of Nuclear
Reactor Regulation, Division of Engineering and Division of License Renewal staff
members on an audit of NextEras large scale test activities at the University of Texas -
Austin, Ferguson Structural Engineering Laboratory (FSEL). The audit was conducted
to support the NRC staffs ongoing review of the Seabrook license renewal application.
The audit team observed the progress and implementation of NextEras large specimen
testing program.
The inspectors also conducted an in-office review of NextEras root cause evaluation
(RCE) and corrective actions related to the identification of differential movement, or
deformation, of the containment enclosure building (CEB). This condition is described in
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an NRC inspection report, dated August 25, 2015 (Agencywide Documents Access and
Management System (ADAMS) Ascension No. ML15217A256). In addition, inspectors
reviewed evaluations related to discrete cracks identified in below grade, internal walls in
the RHR and containment spray (CS) equipment vault structure. This condition was
previously described in an NRC inspection report dated August 5, 2014 (ADAMS
Accession No. ML14212A458).
The inspectors assessed the problem identification threshold, operability and
functionality assessments, extent of condition reviews, and the prioritization and
timeliness of corrective actions to determine whether NextEra personnel were
appropriately identifying, characterizing, and correcting problems associated with the
ASR-affected CEB and RHR vault. The inspectors evaluated NextEras actions to verify
compliance with the CAP procedure and 10 CFR 50, Appendix B requirements.
The resident inspectors continued to conduct routine walk downs of Seabrook Station
ASR-affected structures, focusing on the identification of any changes that have the
potential for further degrading the functional performance of these structures or
associated systems and components.
b. Findings and Observations
During the week of October 26, 2015, the NRC audit team observed the last planned
large-scale specimen testing at FSEL, reviewed test program results and analyses
completed, to date, and interviewed NextEra staff and their consultants. Audit team
activities and conclusions are documented in an NRC audit report (ADAMS Accession
No. ML15307A019) dated December 17, 2015. No significant observations or concerns
were identified related to the conduct of testing to appropriate quality assurance criteria.
NextEra staff planned to have the test data analyses completed by the end of 2015, in
support of submitting to the NRC a license amendment request in 2016, to address an
ASR-related non-conforming condition with the current licensing basis.
NRC inspectors conducted in-office reviews of the root cause evaluation for the
Containment Enclosure Building Local Deformation, Event Dated December 19, 2014,
completed per AR 02014325, and the Condition Assessment of Cracking in RHR and
CS Equipment Vault, documented in Foreign Print (FP) 100903, dated March 17, 2015.
NextEras CEB RCE described two root causes. First, regarding the physical causes of
CEB deformation, NextEra staff concluded that internal expansion (strain) produced by
ASR in the CEB concrete (in-plane direction of the CEB shell) and ASR expansion in the
backfill concrete, coincident with a unique building configuration, resulted in CEB
deformation. Second, regarding NRC identification of this issue, NextEra concluded this
was not identified by plant staff due to an organizational mindset that viewed conditions
such as concrete cracks, water infiltration, and misalignment issues as acceptable and
inconsequential. Additionally, the RCE identified that NextEra staff did not perform and
document comprehensive evaluations of building conditions that could have potentially
revealed more significant underlying conditions, such as localized deformation of the
CEB. These NRC-identified performance deficiencies were previously dispositioned as
non-cited violation (NCV) 2015002-01 and NCV 2015002-02 in NRC inspection report
05000443/2015002 (ADAMS Accession No. ML15217A256).
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The inspectors determined NextEra corrective actions to address these problems
included: 1) a revision to their design control procedures to require pozzolanic materials
like fly ash or slag cement to be added to concrete mixes to prevent ASR in any new
concrete structures; and, 2) the implementation of multiple training and program
changes to correct the organizational mindset issues and strengthen individual
responsibilities and accountability for implementation of the Seabrook Structures
Monitoring Program. The inspectors concluded the RCE was reasonably thorough and
utilized a cause and effect methodology that was appropriate to the problem statement.
However, NextEras RCE, dated December 19, 2014, concluded that the reason that
NRC inspectors, and not plant staff, identified the presence of localized ASR-induced
deformation in Seabrooks concrete structures was due to an organizational mindset that
viewed conditions such as concrete cracks, water infiltration, and misalignment issues
as acceptable and inconsequential. The inspectors concluded the RCE was reasonably
thorough and utilized a cause and effect methodology that was appropriate to the
problem statement. The planned and in-process organizational and program related
corrective actions appeared appropriately focused on the identified causes of the
problem. However, the inspectors concluded that the corrective actions, taken to date,
to implement multiple training, program and oversight changes to correct the
organizational mindset issues and strengthen individual responsibilities and
accountability for implementation of the Seabrook Structures Monitoring Program have
either not been implemented or are not yet effective and thus require additional
management attention near-term.
The inspectors noted that the CEB RCE referenced the results of a Finite Element
Analysis (FEA) model of the CEB. The FEA and results were documented in FP
100985. The inspectors review of FP 100985 identified that the FEA model simulated
ASR expansion to assess the impact of expansion induced deformation on the structural
performance of the CEB. The FEA evaluated CEB design capacity against assumed
loading, based on ACI 318-71 criteria, including simulated loads associated with the as-
deformed condition. As documented in FP 100985, the FEA model also simulated the
impact of the external structural loading due to ASR expansion of the backfill concrete.
Based upon the review of the CEB structural assessment described in FP 100985 and
the limited structural analysis of the RHR/CS vault documented in FP100903, the
inspectors had multiple follow-up questions regarding the CEB and RHR/CS vault
structural assessments and the potential impact of these evaluation results on NextEras
open Immediate Operability Determinations and Prompt Operability Determinations
(PODs) for these ASR-affected structures (reference AR 01664399, AR 01929460, AR
01977456, AR 02004749, and AR 02044627). The follow-up questions were developed
via a collegial review by the Region 1 Senior Reactor Analyst and structural engineers
from NRC Offices of Nuclear Reactor Regulation and Region 2, under the auspices of
the Seabrook ASR Technical Team (ADAMS Accession No. ML14014A378). The
questions were documented and shared with the NextEra staff on December 23, 2015
(ADAMS Accession No. ML15357A326). NextEras responses and the inspectors
review are planned for the first quarter of 2016 and will be documented in an NRC
inspection report.
Following the guidance of IMC 0612, Power Reactor Inspection Reports, the inspectors
identified an issue of concern regarding NextEras implementation of their Seabrook
Structures Monitoring Program. Specifically, the structural evaluations, performed by
contractors and accepted by NextEra staff via the FP process, included discussions that
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identified the potential to exceed limits in the applicable design code (ACI 318-71) for
specific locations in the CEB and RHR/CS vault walls. The evaluations further
recommended actions to determine whether this was the case. The inspectors noted
that the Seabrook staff screened or reviewed these evaluations without documenting a
justification in a revision or update to the open PODs for these structures. The additional
information requested on December 23, 2015 is required for the inspectors to determine
whether this issue involves a performance deficiency. As a result, the NRC opened an
unresolved item (URI).
Introduction. The inspectors identified an issue of concern regarding NextEras
implementation of the Seabrook structures monitoring program and acceptance of
evaluations via the FP and CAP.
Description. Additional inspection is warranted to determine whether a performance
deficiency exists related to NextEras disposition of FP 100985 for the CEB condition and
FP 100903 for the RHR/CS vault condition. Specifically, further inspection is warranted
to determine whether NextEra staff properly implemented the Seabrook structures
monitoring program for the acceptance and review of structural evaluations potentially
impacting the functionality of the CEB and RHR/CS vault, as currently documented in
open PODs. (URI 05000443/2015-01, Issue of Concern Regarding Implementation
of the Seabrook Structures Monitoring Program and Structural Evaluations of the
CEB and RHR/CS Vault)
4OA5 Other Activities
Operation of an ISFSI at Operating Plants (60855 & 60855.01)
a. Inspection Scope
The inspectors evaluated Seabrooks activities related to long-term operation and
monitoring of their ISFSI, and verified that activities were being performed in accordance
with the Certificate of Compliance (CoC), TSs, regulations, and site procedures.
The inspectors reviewed any changes to the10 CFR Part 72.212 evaluation as well as
associated Part 72.48 and 50.59 screenings related to dry cask storage.
The inspectors performed tours of the ISFSI pad to assess the material condition of
the pad. The inspectors also verified that transient combustibles were not being stored
on the ISFSI pad or in the vicinity of the HI-STORMS. The inspectors verified that
Seabrook was performing daily HI-STORM surveillances in accordance with TS
requirements, in addition to the periodic monitoring of the condition of the pad HI-
STORM exterior surfaces.
The inspectors interviewed spent fuel group personnel and reviewed Seabrooks
program associated with fuel characterization and selection for storage from the last
ISFSI loading campaign in 2013. The inspectors verified that the criteria meets the
conditions for cask and canister use as specified in the CoC. The inspectors also
confirmed that physical inventories were conducted annually and were maintained as
required by the regulations.
29
The inspectors reviewed radiological records from the last ISFSI loading campaign to
confirm that radiation and contamination levels measured on the casks were within limits
specified by the TS and consistent with values specified in the final safety analysis
report. The inspectors reviewed RP procedures associated with ISFSI operations. The
inspectors also reviewed annual environmental reports to verify that areas around the
ISFSI pad and the ISFSI site boundary were within limits specified in 10 CFR Part 20
and 10 CFR Part 72.104. The inspectors verified that radiation surveys of the ISFSI pad
were performed in accordance with site procedures.
The inspectors reviewed CAP ARs, and the associated follow-up actions associated with
ISFSI operations to ensure that issues were entered into the CAP, prioritized, and
evaluated commensurate with their safety significance.
b. Findings
No findings were identified.
4OA6 Meetings, Including Exit
On January 21, 2016, the inspectors presented the inspection results to Mr. Dean
Curtland, Site Vice President, and other members of the Seabrook Station staff. The
inspectors verified that no proprietary information was retained by the inspectors or
documented in this report.
4OA7 Licensee-Identified Violation
The following violation of very low safety significance (Green) was identified by NextEra
and is a violation of NRC requirements. The violation meets the requirements of the
NRC Enforcement Policy for being dispositioned as an NCV.
Title 10 CFR 50.55a (g)(4), Inservice Inspection Requirements, requires, in part,
that throughout the service life of a boiling or pressurized water-cooled nuclear
power facility, components (including supports) that are classified as ASME Code
Class 1, must meet the requirements set forth in Section XI of editions and addenda
of the ASME Boiler Pressure and Vessel Code.Section XI of the ASME Boiler
Pressure and Vessel Code, 2001 Edition with 2003 Addenda, Table IWF-2500-1,
Examination Category F-A Supports, requires VT-3 examination of 100 percent of
the ASME Class 1 supports, other than piping supports, every ISI Interval
(examination item F1.40). Contrary to this requirement, from initial commercial
operation until October 14, 2015, (when NextEra staff completed the initial required
VT-3 examinations) NextEra did not perform the required ASME Section XI VT-3
examination of ASME Class 1 supports (i.e. seismic support plates and associated
load path components) on the CRDM assemblies of Seabrook Unit 1. NextEra staff
entered the issue into their CAP as AR 01991880 and completed the VT-3
30
examinations during the October 2015 refueling outage. The finding is more than
minor because it was associated with the protection against external factors attribute
of the Mitigating Systems cornerstone and adversely affected the objective to ensure
reliability and capability of systems that respond to initiating events to prevent
undesirable consequences. The inspectors determined that this finding is of very low
safety significance (Green) in accordance with IMC 0609, Appendix A, Exhibit 2,
Mitigating Systems, because the finding did not involve the loss or degradation of
equipment or function specifically designed to mitigate a seismic initiating event.
NextEra completed the required examinations on October 14, 2015.
ATTACHMENT: SUPPLEMENTARY INFORMATION
A-1
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Curtland, Site Vice President
R. Dodds, Plant General Manager
C. Adams, Operations Training Manager (Acting)
K. Boehl, Senior Rad Protection Analyst
V. Brown, Senior Licensing Engineer
D. Currier, Emergency Planning Manager
K. Douglas, Maintenance Director
D. Flahardy, Radiation Protection Manager
A. Giotas, Chemistry Analyst
S. Hamel, NDE Level III
D. Hickey, Radiation Protection Supervisor
J. Kennish, Instructor
D. Loch, Reactor Vessel Upper Head Inspection Lead
R. Mauer, SG Data Analyst
G. Mikos, Project Manager
D. Merrill, Simulator Support Supervisor
M. Nadeau, RP Analyst
W. Neal, ANII
M. Ossing, Licensing Manager
J. Packer, MSIP Project Manager
K. Randall, Reactor Engineering
D. Ritter, Site Operations Director
D. Robinson, Chemistry Manager
D. Slavon, ISI Program Manager
T. Smith, Radiation Protection Supervisor
D. Strand, RP Manager
C. Thomas, Licensing Engineer
K. Thompson, SG Engineer
T. Vassalo, MPR staff
S. Wellhoffer, Nurse Manager RN
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened
05000443/2015004-01 URI Issue of Concern Regarding Implementation of
the Seabrook Structures Monitoring Program
and structural evaluations of the CEB and
RHR/CS Vault (Section 4OA2.5)
Attachment
A-2
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
OP-AA-102-1002, Seasonal Readiness, Revision 9
OP-AA-102-1002-F01, Cold Weather Data Sheets, Revision 0
OS1023.75, Operation of Feedwater Isolation Valves Temporary Heating, Revision 04
Condition Reports
02071737 02075786 02093279 02096190 02098538 02099933
Maintenance Orders/Work Orders
40266470 40334413 40334716 40353490 40410503
Miscellaneous
BM-Winter-Ready, Cold Weather Preventive Maintenance Activity
Memo to Chief Operating Officer, Seasonal Readiness - Winter 2015, September 24, 2015
Section 1R04: Equipment Alignment
Procedures
OS1002.02, Operation of Letdown, Charging, and Seal Injection, Revision 48
OS1012.03, Primary Component Cooling Water Loop A Operation, Revision 23
OS1012.04, Primary Component Cooling Water Loop B Operation, Revision 26
OS1013.04, Residual Heat Removal Train B Startup and Operation, Revision 28
OS1016.01, Service Water System Fill and Vent, Revision 18
OS1212.01, PCCW System Malfunction, Revision 13
Condition Reports
01943964 01955559 01961777 02038084 02038084 02043141
02055074 02079841 02083248 02083834 02088009 02092165
02092171
Maintenance Orders/Work Orders
40311084 40418154 40419867
Maintenance Requests/Work Requests
94096827 94117626 94117629 94127549
Miscellaneous
Seabrook Component Cooling System Detailed Systems Text, Revision 9
Section 1R05: Fire Protection
Procedures
FP2.2, Control of Combustible Materials, Revision 20
A-3
Miscellaneous
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, C-F-1-Z
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, C-F-2-Z
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, C-F-3-Z
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, CB-F-1A-A
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, CB-F-1G-A
Seabrook Station Fire Protection Pre-Fire Strategies, Volume II, SEPS-F-1-0
Section 1R07: Heat Sink Performance
Procedures
PEG-268, Heat Exchanger and NRC GL 89-13 Program, Revision 0
Condition Reports
01957744 02037230 02050993 02075755 02080499 02081299
02083447 02094198
Miscellaneous
Evaluation of DG-E-42B Tube Plugging
Section 1R08: Inservice Inspection
Procedures
ES15-01-03, Seabrook Unit 1 Reactor Vessel Nozzle MSIP Dissimilar Metal Weld Centerline
Locating and Marking Field Services Procedure, Revision 0, 7/31/2017
Seabrook Station Engineering Procedure ES15-01-04, Revision 0, 7/24/2017; Title:
Seabrook Unit 1 Reactor Vessel Nozzle Mechanical Stress Improvement Process Field
Service Procedure, July 24, 2017
Seabrook Station Engineering Procedure ES10-01-32, Revision 2, 7/24/2017; Title: Remote
Inservice Examination of Reactor Vessel Nozzle to Safe End, Nozzle to Pipe and Safety
End to Pipe Welds Using the Nozzle Scanner, June 30, 2017
Seabrook Unit 1 Reactor Vessel Nozzle Mechanical Stress Improvement Process Field Service
Procedure, July 24, 2017
Seabrook Station Engineering Procedure ES1807.050, Revision 1, Visual Examination (VE) For
CRDM Inspections, ASME Code Case N-729-1
Seabrook Station Engineering Procedure ES10-01-32, Revision 2, Remote Inservice
Examination of Reactor Vessel Nozzle to Safe End, Nozzle to Pipe and Safe End to Pipe
Welds Using the Nozzle Scanner, June 30, 2017
Seabrook Station Engineering Procedure ES10-01-31, Revision 3, ET Examination of Reactor
Vessel Pipe Welds Inside Surface, August 31, 2017
Seabrook Station Engineering Procedure ES1807.050, Revision 1A; Visual Examination (VE)
Procedure for Reactor Vessel Upper Head (RVUH) Penetration Inspections (N-729-1)
Seabrook Station Administrative Procedure MA 10.3, Revision 15, Boric Acid Corrosion Control
Program
Seabrook Engineering Department Instructions (EDI) BORIC ACID EVALUATIONS, EDI No.
30560, Revision 0
Seabrook Chemistry Procedure CS0905.08, Revision 8; Response To A Primary-to-Secondary
Leak
Seabrook Station Engineering Department Procedure ES1807.050, Revision 1;
Visual Examination (VE) Procedure for CRDM Inspections (N-729-1)
A-4
Seabrook Station Engineering Department Procedure ES1807.037, Revision 4;
Visual Examination (VE) Procedure for BMI Inspection (N-722-1)
Seabrook Station Engineering Department Instruction CDI-018, Revision 4; Primary-To-
Secondary Leak Rate Monitoring Below 1 GPD
Seabrook Station Engineering Procedure ES10-01-31, Revision 3; ET Examination of Reactor
Vessel Pipe Welds Inside Surface, August 31, 2017
Condition Reports
00207921 01934094 01954739 01956858 01956974 01958294
01958658 01959231 01960060 01960192 01960662 01962130
01967763 01968281 01977393 01979332 01980333 01981302
01981663 01985049 01985507 01995292 02005943 02010624
02013631 02017775 02020023 02023008 02025166 02030144
02032984 02038340 02038887 02039668 02042461 02044332
02044452 02044899 02045365 02049171 02051868 02058610
02059112 02065121 02065815 02069165 02070738 02072120
02073698 02073706 02073711 02073712 02073714 02073720
02073720 02073740 02079633 02079633 02079633 02079633
02079690 02081530 02081530 20133359
Documented Boric Acid Leaks and Evaluations
01751940 01751948 01954739 01954894 01956887 01958294
01958658 01960060 01960192 01967763 01968281 01969662
01979332 01981302 01981663 01985507 02005943 02013359
02013631 02017775 02020023 02023008 02025166 02039668
02043261 02043789 02044332 02044452 02044772 02044772
02044899 02045365 02073706 02073711 02073712 02073714
02073720 02073720 02078055 02078097 02078112 02078117
02078120 02078121 02078124 02078128 02078130 02078133
02078134 02078135 02078139 02078140 02078143 02078145
02078146 02078268 02078274 02078472 02078700 02078839
02079208 02079236 02079241 02079247 02079278 02079287
02079290 02079294 02079297 02079303 02079318 02079325
02079330 02079333 02079336 02079341 02079342 02079343
02079349 02079353 02079360 02079361 02079363 02079366
02079367 02079372 02079375 02079377 02079380 02079381
02079385 02079389 02079393 02079394 02079398
20278063
Self-Assessments
Quick Hit Assessment Report, AR 02004668, Seabrook: Engineering Programs
Quick Hit / Department Assessment Report, AR 2024069, Fleet Steam Generator Program Self-
Assessment; 7/30/15
Focused Self-Assessment AR 1773622; Chemistry Sampling and Analysis Program, 8/15/12
Focused Self-Assessment AR 02050220; Maintenance Implementation of Welding Program,
9/15/15
Quick Hit Assessment Report, AR 01950056, Seabrook/Engineering Programs - Materials,
OR16 Inservice Inspection (ISI) NRC Inspection, 3/24/14
Quick Hit Assessment Report, AR 01725016, Seabrook Engineering Programs - Materials,
Title: Integration of the Alloy 600 Program Into Inservice Inspection
A-5
Engineering Evaluations Design Changes and SG Procedures:
Engineering Evaluation EE-09-016 Revision 0; Engineering Evaluation Flaw Indications In
Reactor Vessel Safe End Welds. By H.W. Mentel-Seabrook Station, October 22, 2009
Engineering Change (EC) - 281842, Revision 0 Design Change Package Form MSIP,
15 Pages Document/PCR/EC#281724, Revision 0, 10 CFR Applicability Determination
Form, 5 Pages
Steam Generator Degradation Assessment for Seabrook OR17 Refueling Outage, Revision 1,
SG-SGMP-15-10, October 2015
Steam Generator Degradation Assessment for Seabrook OR17 Refueling Outage, Revision 2,
SG-SGMP-15-10, October 2015
Seabrook OR16 Condition Monitoring Assessment and Final Operational Assessment, April
2014, SG-SGMP-15-10
Seabrook Appendix H & I Techniques, Fall 2015 Inspection; MRS-TRC-2286, October 2015
EC-284861, Revision 0; Evaluation #EE-14-013, Revision 1; Seabrook Unit 1, Title:
Reactor Vessel Head Effective Degradation Years (EDY) & Re-Inspection Years (RIY)
for Cycle 17
ACR # 96-1258; Adverse Condition Report, 69034975, Cause Determination and corrective
actions for 1996 leaks from SG B and SG C, 11/25/96
Seabrook OR17 Condition Monitoring Assessment and Final Operational Assessment,
November 2015, SG-SGMP-15-20, Revision 0
NDE Data Reports:
Examination data sheets related to seven reactor vessel nozzle dissimilar metal welds as
follows: RC-RPV-SE-301-121-338-deg-D; RC-RPV-SE-302-121-293 deg-C;
RC-RPV-SE 302-121-247-deg-B; RC-RPV-SE-301-121-202-deg-A;
RC-RPV-SE-302-121-113 deg-G; RC-RPV-SE-301-121-67-deg-F; RC-RPV-SE-301-
121-22 deg-E
Data Sheet 15-VT-041, CRDM structural support, VT3 examination10-15-15
ES1807.037 Form A: Visual Examination of BMI Penetrations and RPV Bottom Head -
UNSAT due to Residue/Deposits or Wastage Detected (CR written to sample residue,
clean residue and reinspect penetrations); VT Inspector Notes for each penetration (16
pages) and pictures of each penetration (105 pages)(11/3/15)
ES1807.037 Form A: Visual Examination of BMI Penetrations and RPV Bottom Head after
cleaning and re-inspection- SAT; (16 pages) and pictures of each penetration (11/6/15)
Data Sheet 15-UT-001(5 pages), FW Elbow to pipe, Class 2, Thermal Fatigue Exam
Data Sheet 15-UT-002(5 pages), FW Elbow to pipe, Class 2, Thermal Fatigue Exam
Data Sheet 15-VT3-041(6 pages), CRDM STRUCTURAL SUPPORT Inspection, Class 1
IWE VT-1 Re-Examination of a previously identified Containment Dome Anomaly/Indication
Seabrook Nuclear Power Plant Unit 1
Refueling Outage OR17, Post MSIP Reactor Vessel Nozzle DM Weld Examination Summary,
October 12 through October 22, 2015
Repair Replacement Activity
WO 4036801501; Fabricate B Diesel Fuel Connection for B Train, EC282581
Miscellaneous
NUREG-0313, Revision 2, Nuclear Regulatory Commission, Office of Nuclear Reactor
Regulation, W.S. Hazelton; W.H. Koo, Title: Technical Report on Material Selection and
Processing Guidelines for BWR Coolant Pressure Boundary Piping
NRC Bulletin 2003-02: Leakage from Reactor Pressure Vessel Lower Head Penetrations and
Reactor Coolant Pressure Boundary Integrity, August 21, 2003
A-6
Westinghouse NSAL-14-6, 9/24/2014, Subject: Reactor Coolant Pump Turning Vane Bolt
Cracking and Failures
Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds In Class 1
Components Fabricated with Alloy 600/82/182 MaterialsSection XI, Division 1 Code
Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper
Heads With Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI,
Division 1 Code Case N-770-1, Alternative Examination Requirements and Acceptance
Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS
N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed
Mitigation ActivitiesSection XI, Division 1
EPRI Technical Report 2011, Steam Generator Management Program: Primary-to-Secondary
Leak Guidelines, Revision 4
Seabrook Station Reference Manual; Steam Generator Maintenance Reference, SGRE
Revision 20, 3/10/14
Seabrook Station, Steam Generator Eddy Current Data Analysis Guideline Manual, Revision 7,
4/4/2014
NextEra Energy Program Description, ER-AP-121, Revision 2; Steam Generator Integrity,
9/2/15
Apparent Cause Evaluation for AR 00207921
Section 1R11: Licensed Operator Requalification Program
Procedures
TR-AA-104, NEXTERA Energy Fleet Licensed Operator Continuing Training Program,
Revision 7
TR-AA-220-1004, Licensed Operator Continuing Training Annual Operating and Biennial
Written Exams, Revision 1
TR-AA-230-1004, SAT Implementation, Revision 5
TR-AA-230-1005, SAT Evaluation, Revision 2
TR-AA-230-1007, Crew-Specific Simulator Plan, Revision 3
Condition Reports
01917736 01923551 01954983 01966222 01987175 01988101
01994628 02012298 02014014 02015273 02022491 02022755
02024157 02027235 02038184 02040722 02042410 02046889
02050058 02054056
Miscellaneous
ACAD 10-001, February 2010, Revision 0
ANS 3.4-1983
Focused Area Self-Assessment Report 2058898
LORT Training Program Description, December 2012
NT 3744, Simulator Transient Tests, Revision 17
NRC Information Notice IN 2004-20
NRC Information Notice IN 94-14
NRC Information Notice IN 91-08
Requalification Training Program Annual Examination Sample Plan 2015
Seabrook Transient Tests 1 through 11
A-7
Section 1R12: Maintenance Effectiveness
Procedures
ES1807.025, Inservice Inspection (ISI) Visual Examination Procedure, Revision 6
EX1805.03, Snubber Functional Test, Revision 12
LS0557.19, 480 Vac K600 Electrically Operated Breaker Refurbishment, Revision 15
MS0515.08, Paul-Munroe 2300, 2400 and 2500 Hydraulic Snubber Maintenance, Revision 4
MS0515.12, Cooling Tower Primary Fan and Gear Reducer Maintenance, Revision 5
MS0517.05, Installation of Pipe Supports, Revision 3
PEG-24, Maintenance Rule Goal Setting and Monitoring, Revision 8
PEG-45, Maintenance Rule Program Monitoring Activities, Revision 17
Condition Reports
02065121 02071704 02081775 02083296 02093663 02094577
02099774
Maintenance Orders/Work Orders
40289589 40353703 40379186 40410443 40418627
Miscellaneous
EE-10-010, Maintenance Rule PRA Basis Document PRA Risk Ranking and Performance
Criteria Based on SSPSS-2009, dated March 2011
Maintenance Rule Functional Failure Evaluation, A main Steam Line (RM-6481-1) in High
Alarm - Not Valid
NUMARC 93-01, Industry Guidelines for monitoring the Effectiveness of Maintenance at
Nuclear Power Plants, Revision 2
Radiation Monitoring System Health Report
SBK-L-15183, Seabrook Station Technical Specification Special Report - Radiation Monitor
Inoperable Greater Than 7 Days
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
ODI101, Guarded Equipment Recommendations for Refueling Outages, Revision 16
OP-AA-102-1003, Guarded Equipment, Revision 9
Condition Reports
02079027 02087297 02087302
Miscellaneous
LCOTR Log on October 8, 2015
Seabrook Station Shutdown Condition Status Report for 10/8/15 OCC Turnover at 0500
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
PI-AA-104-1000, Corrective Action, Revision 3
A-8
Condition Reports
02044332 02080495 02080584 02081530 02082377 02082510
02083255 02083422 02084748 02084825 02085161 02085433
02098607
Maintenance Orders/Work Orders
94131854
Miscellaneous
2004 ASME Boiler and Pressure Vessel Code,Section XVI
Engineering Evaluation Documenting Integrity of the Containment Liner Plate dated
October 28, 2015
NRC Information Notice 2014-07: Degradation of Leak-Chase Channel Systems for Floor Welds
of Metal Containment Shell and Concrete Containment Metallic Liner
Seabrook Station UFSAR
Westinghouse Letter LTR-RC-15-47, Operability Assessment for Ball Bearings Primary Side
Loose Parts at Seabrook Unit 1
Drawings
1-MS-B20582, Main Steam System Emergency Feedwater Pump Supply Detail, Revision 21
9763-F-805139, Containment Liner Floor Leak Chase System Piping Arrangement, Revision 10
Section 1R18: Plant Modifications
Procedures
EN-AA-205-1100, Design Change Package, Revision 13
EN-AA-203-1201, 10 CFR Applicability of 10 CFR 50.59 Screening Reviews, Revision 4
ES1807.025, Form B, VT-2 Visual Examination Form, Revision 6
ES1850.012, Air Operated Valve Program Procedure, Revision 4
ES1810.206, Charging System - Charging Pump Discharge Pipe ISI System Leakage Test,
Revision 2
OX1408.06, Controlled Leakage Monthly Surveillance, Revision 7
OX1456.01, Charging Pump A & B Quarterly Flow and Valve Stroke Test And 18 Month
Remote Position Indication Verification, Revision 21
SM7.20, Control of Time Critical Actions and operator Response Times, Revision 8
Condition Reports
01973578 02041778 02041838 02042131 02079830 02087598
02090399
Maintenance Orders/Work Orders
40339187 40367789
Miscellaneous
50.59 Evaluation No. 15069, Revision 0
Calculation #CS145705, Revision 2
DBDEFW01, Emergency Feedwater System Design Basis Document, Revision 6
EC282453, Revision 0
EC282474, Revision 1
ES15-01-18, Radiographic Examination in Accordance with ASME Section V, Article 2, Rev. 0
A-9
Instruction Book 205, Reactor Coolant Pump Model W-11012-A1 (93-A1), Revision 5
MISTRAS Radiographic Inspection Report, 3INCH-SCH160-10132015, dated 10/14/15
PWROG-14006-P, Implementation Guide for the generation III Westinghouse Shutdown Seal,
Revision 0-B
Seabrook Emergency Feedwater System Detailed Systems Text, Revision 10
Westinghouse LTR-RES-14-104, Shutdown Seal Purge Volume, Coastdown Time and Material
Properties for Seabrook Station, dated October 13, 2014
Westinghouse LTR-RES-14-126, Impact on the Seismic Evaluation by Installing the SHIELD at
Seabrook Station, dated November 13, 2014
Westinghouse LTR-SEE-I-14-61, Downstream Effects Evaluation of an Inadvertent Actuation of
the RCP Shutdown Seal at Seabrook Unit 1, dated December 4, 2014
Drawings
1NHY250000, Rev. 81, Data Sheets for Motor and Air Operated Valves and Dampers
Section 1R19: Post-Maintenance Testing
Procedures
LS0557.19, 480 VAC K-600 Electrically Operated Breaker Refurbishment, Revision 15
LX0556.113, 1-EDE-B-1-D 18-Month Technical Specification Battery Surveillance, Revision 3
OX1410.04, Post Refueling Pre-Startup Rx Trip Breaker Surveillance, Revision 10
OX1413.03, B Train RHR Quarterly Flow and Valve Stroke Test, and 18 Month Valve Stroke
Observation, Revision 13
OX1413.05, RHR Cold Shutdown Valve Testing, Revision 7
OX1413.08, RH-P-8B Comprehensive Pump Test, Revision 10A
OX1416.03, Monthly Cooling Tower Fan Operability Test, Revision 10
OX1436.13, Turbine Driven Emergency Feedwater Pump Post Cold Shutdown or Post
Maintenance Surveillance and Comprehensive Pump Test, Revision 29
Condition Reports
01647943 01837698 02069435 02071668 02083213 02083832
02085868 02085922 02090725 02091049
Maintenance Orders/Work Orders
40103650 40289589 40310982 40310991 40310993 40311461
40311466 40315437 40369086 40421948 40436813 40436853
Miscellaneous
1-NHY-31017, Sheet EG2a, 125VDC Bus 1-SWG-11D Distr Pnl 1-PP-111D Schedule,
Revision 8
1-NHY-310042, Sheet 2, 125VDC Vital Distribution System One Line Diagram, Revision 4
1-NHY-310942, Sheet E04/10, Process Protection Cabinet Power Supply MM-CP-4 Cable
Schematic, Revision 0
1-NHY-310949, Sheet E01/2a, Solid State Protection System Cabinet MM-CP-12 Schematic
Diagram, Revision 5
1-NHY-310949, Sheet E02/4a, Solid State Protection System Cabinet MM-CP-13 Schematic
Diagram, Revision 5
A-10
Section 1R20: Refueling and Other Outage Activities
Procedures
AD-AA-101-1004, Work Hour Controls, Revision 16
EX1808.014, Containment Enclosure Building Integrity 18 Month Surveillance, Revision 10
IS1682.222, Spent Fuel Pool Level Calibration, Revision 7
ODI-101, Attachment D, Outage Status Sheet, Revision 16
ON1090.04, Containment Entry, Revision 30
OP-AA-102-1003, Guarded Equipment, Revision 9
OS1000.02, Plant Startup From Hot Standby to Minimum Load, Revision 31
OS1000.04, Plant Cooldown From Hot Standby to Cold Shutdown, Revision 51
OS1000.07, Approach to Criticality, Revision 13
OS1000.09, Refueling Operation, Revision 30
OS1000.10, Operation at Power, Revision 31
OS1000.14, Reactor Coolant System Evacuation and Fill, Revision 22
OS1000.14, Reactor Coolant System Evacuation and Fill, Revision 23
OS1001.02, Draining the Reactor Coolant System for Vessel Head Removal, Revision 19
OS1013.03, Residual Heat Removal Train A Startup and Operation, Revision 31
OS1013.04, Residual Heat Removal Train B Startup and Operation, Revision 28
OS1015.18, Setting Containment Integrity for Mode IV Entry, Revision 10
OX1406.12, 18 Month Containment and Containment Spray Recirculation Sump Surveillance,
Revision 11
OS1436.19, 18 Month ATWS Mitigation System Auto Actuation Surveillance, Revision 9
OX1426.34, Diesel Generator 1A 36 Month Operability Surveillance, Revision 17
RS1737, Post Refueling Low Power Physics Testing, Revision 7
Condition Reports
02005438 02074905 02080584 02080921 02080921 02082510
02084164 02088138 02097832
Maintenance Orders/Work Orders
40273392 40312116 40339169 40339170 40367789 40416945
Miscellaneous
Cycle 18 Core Map, dated 10/17/15
Drawing 9763-F-805139, Containment Liner Floor Leak Chase System Piping Arrangement,
Revision 10
Engineering Change 281724, Revision 0
Engineering Change 284624, Revision 0
LTR-PAFM-15-104, Seabrook SG & RCP Support Shim Gap Measurement Process Flowchart
LTR-PAFM-15-017, Seabrook SG & RCP Support Shim Gap Re-Shim Guidance
PRA Evaluation PRAE-15-011, AR 2057299, OR17 Outage Schedule Shutdown Risk Review
AT-01.13 Mode Hold / Milestone Reports
Report SBK 15-013, Seabrook Nuclear Oversight Report, dated 11/19/15
Westinghouse Letter, LTR-RVHP-15-56, Seabrook SG & RCP Support Option 2 Shim
Adjustments Post MSIP, dated October 20, 2015
Westinghouse Letter, LTR-ISE-15-039, Power Ascension Shim Gap Monitoring Plan for
Seabrook Station Unit 1 During OR17, dated October 22, 2015
A-11
Section 1R22: Surveillance Testing
Procedures
ES1850.003, Motor Operated Valve Performance Monitoring, Revision 10
LS0569.16, Diagnostic Testing of Rising Stem MOVs, Revision 23
LS0569.17, Diagnostic Testing of Rising Stem MOVs Using the Torque Thrust Cell, Revision 17
OX1402, CVCS Cold Shutdown and Refueling Interval Valve Test, Revision 13
OX1456.81, Operability Testing of IST Valves, Revision 23
Condition Reports
02004690 02083866
Maintenance Orders/Work Orders
40312073 40312078 40321086 40414003 40414014
Section 1EP2: Alert and Notification System Evaluation
Procedures
EP-AA-101-1000, Nuclear Division Drill and Exercise Procedure, Revision 13
EP-AA-102, NextEra Energy Alert and Notification System, Revision 0
SIR10, WPS-3000 and WPS-4008 Siren Bi-Weekly Functional Test, Revision 4
SIR11, WPS-3000 and WPS-4008 Siren Annual Maintenance, Revision 2
SIR45, State Siren Activation Control System Annual Maintenance and Testing, Revision 4
SIR76, Local Town Siren Activation Control System Annual Maintenance and Testing,
Revision 4
Miscellaneous
Seabrook Station ANS Maintenance and Testing records, August 2013 - December 2015
Seabrook Station Nuclear Power Plant Alert and Notification System Design Report,
Addendum 7, June 2013
Seabrook Station Radiological Emergency Plan, Revision 69
Section 1EP3: Emergency Response Organization Staffing and Augmentation System
Procedures
NM11700, Emergency Preparedness Responsibilities of Primary, Subject-to-call, and
Secondary, Emergency Response Organization members, Revision 35
Miscellaneous
2014 3rd Quarter Emergency Response Organization Notification Test Results
2014 4th Quarter Emergency Response Organization Notification Test Results
2015 3rd Quarter Emergency Response Organization Notification Test Results
2015 4th Quarter Emergency Response Organization Notification Test Results
E-Plan Personnel Expiration Report - 14 day look ahead dated December 4, 2015
ERO Staffing Succession Planning dated December 8, 2015
Radiological Emergency Preparedness Drill and Exercise Manual, Revision 41
Seabrook Station Emergency Response Manual, Revision 122
Seabrook Station On-Shift Staffing Analysis Report, Revision 1
Seabrook Station Radiological Emergency Plan, Revision 69
Weekly ERO Pager Test, and Quarterly ERO Call-In Drill, results, January 2012 - June 2013
A-12
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Procedures
ER1.1, Classification of Emergencies, Revision 55
ER1.2, Emergency Plan Activation, Revision 64
ER5.3, Operation of the Raddose-V, Revision 33
ER5.4, Protective Action Recommendations, Revision 34
Miscellaneous
Radiological Emergency Plan (SSREP), Revision 69
Section 1EP5: Maintaining Emergency Preparedness
Procedures
EN-AA-203-1201, 10 CFR Applicability of 10 CFR 50.59 Screening Reviews, Revision 4
EP-AA-100-1002, Emergency Preparedness Change Review Committee Guideline, Revision 2
EP-AA-100-1007, Evaluation of the Change to the Emergency Plan, Supporting Documents and
Equipment (10CFR 50.59(a)), Revision 3
SM7.28, Seabrook Equipment Important to Emergency Response, Revision 2
Condition Reports
01890396 01937133 02005383 02005487 02023560 02025317
02030373 02031036 02032420 02052946 02052947 02052947
02068065
Miscellaneous
Benchmark Report on Use of Teams in the ERO, AR 02062157
EP Readiness Assessment for INPO Evaluation, AR 02014862
Formal Benchmark Report on VC Summer Emergency Preparedness, AR 01928502
Quick Hit Assessment Report on EP NRC Inspection Readiness, AR 02004700
Quick Hit Assessment Report on EAL Classification Timing, AR 01898102
Quick Hit Assessment Report on ERO Self-Assessment, AR 02004700
Quick Hit Assessment Report on PARs for Large Bodies of Water in EPZ, AR 02022403
SEP 2014016, Seabrook Station CFE 14-05 Report
SEP 2015003, Seabrook Station CFD 15-01 Report
SEP 2015016, Seabrook Station CFD 15-02 Report
Seabrook Nuclear Oversight Report, SBK 13-012
Seabrook Nuclear Oversight Report, SBK 14-008
Seabrook Nuclear Oversight Report, SBK 15-007
Seabrook Station Development of Evacuation Time Estimates 2014, Final Report, Revision 0
Seabrook Station Development of Evacuation Time Estimates 2015, Final Report, Revision 0
Seabrook Station Radiological Emergency Plan, Appendix D, letters of Agreement with
Emergency Response Organization, Revision 60
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Procedures
HD095817, Performance of Routine Radiological Surveys, Revision 13
HD0955.50, Far West REM-500 Operation, Revision 6
HN0960.10, Radiological Requirements for Entry Beneath Reactor Vessel, Revision 31
A-13
HN0960.17, Radiological Controls for Transfer of Spent Fuel between the Containment and the
Spent Fuel Pool, Revision 4
OA 13-007 Pre-Planned Posting and Survey Instructions, Revision 0
RP-AA-102-1001, Area Radiological Surveys, Revision 1
RP-AA-103-1001, Posting Requirements for Radiological Hazards, Revision 2
RP-AA-103-1002, High Radiation Area Controls, Revision 3
RP-AA-107-1001, Radioactive Material Receipt, Revision 2
RP-AA-107-1003, Unconditional and Conditional Release of Material, Revision 1
Condition Reports
02075677 02075678 02078744 02078869 02081011 02092044
Miscellaneous
Air Sample Results15-491 Transfer Canal FSB, October 4, 2015
Air Sample Results15-515 CTMT Reactor Cavity, October 7, 2015
Air Sample Results15-683 CTMT Reactor Cavity, October 20, 2015
Air Sample Results15-708 CTMT 25 Reactor Head O Ring, October 22, 2015
Air Sample Results15-778 CTMT Reactor Cavity during Decontamination, October 30, 2015
HD0958.19 Form A Dosimetry Abnormality Occurrence Report, ED Dose Alarm Badge Number
6494, November 2, 2015
HD0958.19 Form A Dosimetry Abnormality Occurrence Report, ED Dose Alarm Badge Number
6478, November 5, 2015
HD0958.19 Form A Dosimetry Abnormality Occurrence Report, ED Dose Alarm Badge Number
6564, November 3, 2015
HD0958.19 Form A Dosimetry Abnormality Occurrence Report, ED Dose Alarm Badge Number
6045, November 5, 2015
HD0958.19 Form A Dosimetry Abnormality Occurrence Report, Lost TLD Badge Number 6584,
October 20, 2015
HD0958.03 Form A Personnel Contamination Report, Facial Contamination Badge Number
5937, October 19, 2015
HSYM-085 YRD-20-YD100, Radioactive Material Storage Building Rad Survey M-20151203-1,
December 3, 2015
HSYQ-001C Radiac Calibration Facility Survey M-20151021-34, October 21, 2015
HSYQ-001D C Vans & RM Trucks Survey M-20151102-8, November 2, 2015
HSYQ066A FSB-21-FB202 Fuel Storage Building Survey M-20151101-9, November 1, 2015
HX0958.23 Figure 1, Non Exempt Source Index by Location, September 10, 2015
SB Radioactive Source Leak Test Survey M-20150910-3, September 10, 2015
Section 2RS2: Occupational ALARA Planning and Controls
Procedures
RP15.4, Use and Control Temporary Shielding, Revision 12
RP-AA-104, ALARA Program, Revision 2
RP-AA-104-1000, ALARA Implementing Procedure, Revision 5
Condition Reports
02079857 02089191 02061478 02082603 02082386 02085254
Miscellaneous
Next Era, OR 17 ALARA and Station Dose Performance, December 8, 2015
Next-Era, ALARA Review Board Meeting ARB 15-06 Presentation, December 4, 2015
A-14
RP-AA-104-1000-F2 Pre Job ALARA Review ALARA Package No. 15-16 RCP Tie Rod and
Shim, October 25, 2015
Section 4OA1: Performance Indicator Verification
Procedures
CS0917.02, Gaseous Effluent Releases, Revision 14
CX0917.01, Liquid Effluent Releases, Revision 20
EPDP-03, Emergency Preparedness Performance Indicators, Revision 25
HD0958.33, Performance of Radiation Protection Supervisory Plant Walkdowns, Revision 6
JD0999.910, Reporting Key Performance Indicators per NEI 99-02, Revision 7
Condition Reports
02092877 02079246 02079779 02082355
Miscellaneous
SBK-PRAE-15-001, Seabrook MSPI Basis Document Update 2014, Revision 0
LIC-15004, Performance Indicator Data, 4th Quarter 2014
LIC-15015, Performance Indicator Data, 1st Quarter 2015
LIC-15026, Performance Indicator Data, 2nd Quarter 2015
LIC-15-032, Performance Indicator Data, 3rd Quarter 2015
Alert Notification System PI data, December 2014 - December 2015
DEP PI data, December 2014 - December 2015
ERO Drill Participation PI data, December 2014 - December 2015
NextEra - Seabrook Station 2014 Annual Radioactive Release Report, April 24, 2015
CP4.1B GEW Sample Collection Data, Permit Number 15-321 Steam Generator Blowdown,
August 13, 2015
CP4.1B GEW Sample Collection Data, Permit Number 15-316, Plant Vent, August 15, 2015
CS0908.01 Form F - Radioactive Effluent Dose Assessment Report 2015, December 9, 2015
CS0917.02 Form C GEW Containment Purge Release Permit 15-409 Containment Purge,
September 28, 2015
CX0917.01, Form C LEW Release Data, Permit Number 15-433, Waste Test Tank,
October 7, 2015
CX0917.01, Form C LEW Release Data, Permit Number 15-438, Turbine Building Sump,
October 6, 2015
CX0917.01, Form C LEW Release Data, Permit Number 15-319, Groundwater, July 31, 2015
JD0999.910 Figure 1 Occupational Exposure Occurrence, October, November and December
2014, January 12, 2015
JD0999.910 Figure 1 Occupational Exposure Occurrence, January, February and March 2015,
April 2, 2015
JD0999.910 Figure 1 Occupational Exposure Occurrence, April, May and June 2015,
December 2, 2015
JD0999.910 Figure 1 Occupational Exposure Occurrence, July, August and September, 2015,
October 15, 2015
Section 4OA2: Problem Identification and Resolution
Procedures
EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 21
PI-AA-104-1000, Corrective Actions, Revision 6
PI-AA-207-1000, Station Self-Evaluation and Trending Analysis, Revision 5
A-15
Condition Reports
01664399 01929460 01936340 01972065 01977373 01977456
01990185 01991967 01995588 01999679 02000523 02002668
02002750 02004749 02007975 02011136 02013241 02014325
02014325 02018118 02031149 02034198 02039025 02044627
02063563 02070871 02078215 02078215 02085144 02085147
Maintenance Orders/Work Orders
01191454 40274351 40279910 40314207 40320492 40325682
40345801
Work Requests
94098872 94106748
Miscellaneous
Foreign Print 100903, Condition Assessment of Cracking in RHR and CS Equipment Vault,
dated March 17, 2015
Foreign Print 100985, Investigate Apparent Movement of Containment Enclosure Building at
NextEra Seabrook Facility, Seabrook, NH, dated November 2015
Maintenance Department Rework Assessment dated January 9, 2015
Maintenance Department Rework Assessment dated May 29, 2015
Operations Department Station Self-Evaluation and Trending Analysis Report, 2nd quarter 2015
Operations Department Station Self-Evaluation and Trending Analysis Report, 3rd quarter 2015
Section 4OA5: Other Activities
Procedures
Seabrook Station Instrumentation & Controls Procedure, Dry Fuel Storage HSM Temperature
Monitoring System Calibration Check, IS3600.401, Revision 3
Seabrook Station Radiation Protection Procedure, Performance of Routine Radiological
Surveys, HD0958.17, Revision 13
Seabrook Station Radiation Protection Procedure, DFS Radiological Controls, HX3950.85,
Revision 7
Seabrook Station Security Procedure, Protected Area, DFS, and Vehicle Barrier System
Inspections, GN1322.00, Revision 39
Condition Reports
01898310 01949316 01970650 02036364 02040234 02042227
02059136 02060075 02096511
Calculations
Calc No: SBK-1FJF-13-014, Rev. 2, Seabrook Station- Irradiated Fuel Assembly Selection for
Second Dry Cask Loading Campaign
Miscellaneous
Action Request Report, AR Number 02050994, Concrete Inspection at DFS Area, 05/29/2015
Activity Documentation Sheet, HEN-Q6, Evaluate Quarterly Environmental and Protected Area
TLD Results Second Quarter 2015
DFS Lessons Learned 2013
Evaluation for Blind Spike TLD Test BS-13-08-28 Neutron Test Outside of Tolerance for
AR01949316
A-16
Health Physics Study/Technical Information Document 13-007, Determination of Dry Fuel
Storage Neutron Correction Factor for Global TLDs 09/2013
HSYQ-001M Dry Fuel Storage Surveys, 10/17/2013 through 11/03/2015
Nuclear Engineering Department Discipline Standard, STD-F-040, Revision 6, Nuclear Fuel, Dry
Cask Storage Fuel Assembly Assignments, August 2013
Personnel Contamination Report for AR01898310, 08/21/2013
Pre and Post-Job ALARA Review 13-01, Rev. 0, Dry Fuel Transfer Pool to Pad and Associated
Tasks for 8 Casks
Radiation Work Permit, 15-0023, Dry Fuel Storage Area Maintenance/Inspection Activities,
03/02/2015, Revision 01
Seabrook 2013 ISFSI Demonstration Exercise Plan, 06/2013
Tech Spec and Commitment Logs, DFS Daily Temperature or Visual Inspection July 14, 2015
through October 25, 2015
A-17
LIST OF ACRONYMS
ADAMS Agencywide Documents Access and Management System
ALARA as low as reasonably achievable
ANS Alert and Notification System
AR action request
ASME American Society of Mechanical Engineers
ASR alkali-silica reaction
CAP corrective action program
CEB containment enclosure building
CFR Code of Federal Regulations
CoC Certificate of Compliance
CR condition report
CRDM control rod drive mechanism
EAL emergency action level
EC eddy current
EDG emergency diesel generator
ERF Emergency Response Facility
ERO Emergency Response Organization
FP Foreign Print
FSEL Ferguson Structural Engineering Laboratory
IMC Inspection Manual Chapter
ISFSI independent spent fuel storage installation
ISI inservice inspection
MR maintenance rule
MRC Management Review Committee
NCV non-cited violation
NDE non-destructive examination
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
OR refueling outage
PI performance indicator
POD Prompt Operability Determination
RCE root cause evaluation
RG Regulatory Guide
RP radiation protection
SSC structure, system, and component
TS technical specification
URI unresolved item
VCT volume control tank
WO work order