ML18043A821
| ML18043A821 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 02/12/2018 |
| From: | Fred Bower Division Reactor Projects I |
| To: | Nazar M NextEra Energy Seabrook |
| References | |
| IR 2017004 | |
| Download: ML18043A821 (42) | |
See also: IR 05000443/2017004
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
2100 RENAISSANCE BOULEVARD, SUITE 100
KING OF PRUSSIA, PA 19406-2713
February 12, 2018
Mr. Mano Nazar
President and Chief Nuclear Officer
Nuclear Division
NextEra Energy Seabrook, LLC
Mail Stop: EX/JB
700 Universe Blvd.
Juno Beach, FL 33408
SUBJECT: SEABROOK STATION, UNIT NO. 1 - INTEGRATED INSPECTION REPORT
Dear Mr. Nazar:
On December 31, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an
inspection at Seabrook Station, Unit No. 1 (Seabrook). On January 23, 2018, the NRC
inspectors discussed the results of this inspection with Mr. Eric McCartney, Regional Vice
President, and other members of his staff. The results of this inspection are documented in the
enclosed report.
NRC inspectors documented three findings of very low safety significance (Green) in this report.
Two of these findings involved a violation of NRC requirements. The NRC is treating these
violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement
Policy.
If you contest the violations or significance of these NCVs, 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; and the
NRC Resident Inspector at Seabrook. In addition, if you disagree with a cross-cutting aspect
assignment or a finding not associated with a regulatory requirement in this report, you should
provide a response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,
Washington, DC, 20555-0001; with copies to the Regional Administrator, Region I, and the NRC
Resident Inspector at Seabrook.
M. Nazar 2
This letter, its enclosure, and your response (if any) will be made available for public inspection
and copying at http://www.nrc.gov/reading-rm/adams.html and the NRC Public Document Room
in accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Fred Bower, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Docket No. 50-443
License No. NPF-86
Enclosure:
Inspection Report 05000443/2017004
w/Attachment: Supplementary Information
cc w/encl: Distribution via ListServ
SUNSI Review Non-Sensitive Publicly Available
Sensitive Non-Publicly Available
OFFICE RI/DRP RI/DRP RI/DRP
NAME R. Barkley P. Cataldo/RB F. Bower
DATE 2/7/2018 2/12/2018 2/12/2018
1
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No: 50-443
License No: NPF-86
Report No.: 05000443/2017004
Licensee: NextEra Energy Seabrook, LLC (NextEra)
Facility: Seabrook Station, Unit No. 1 (Seabrook)
Location: Seabrook, NH 03874
Dates: October 1, 2017 through December 31, 2017
Inspectors: P. Cataldo, Senior Resident Inspector
T. Daun, Acting Senior Resident Inspector
P. Meier, Resident Inspector
N. Perry, Senior Resident Inspector
B. Dionne, Health Physicist
B. Cook, Senior Reactor Analyst
N. Floyd, Reactor Inspector
A. Buford, Structural Engineer, NRR
D. Silk, Senior Operations Engineer
Approved By: Fred Bower, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Enclosure
2
TABLE OF CONTENTS
SUMMARY ................................................................................................................................ 3
1. REACTOR SAFETY ........................................................................................................... 6
1R01 Adverse Weather Protection ..................................................................................... 6
1R04 Equipment Alignment ............................................................................................... 6
1R05 Fire Protection .......................................................................................................... 7
1R06 Flood Protection Measures ....................................................................................... 8
1R07 Heat Sink Performance ............................................................................................ 8
1R11 Licensed Operator Requalification Program and Licensed Operator Performance ... 9
1R12 Maintenance Effectiveness ......................................................................................13
1R13 Maintenance Risk Assessments and Emergent Work Control .................................13
1R15 Operability Determinations and Functionality Assessments .....................................14
1R19 Post-Maintenance Testing .......................................................................................14
1R22 Surveillance Testing ................................................................................................15
1EP6 Drill Evaluation ........................................................................................................16
2. RADIATION SAFETY.........................................................................................................20
2RS2 Occupational As Low As Is Reasonably Achievable Planning and Controls ............20
2RS3 In-Plant Airborne Radioactivity Control and Mitigation .............................................20
4. OTHER ACTIVITIES ..........................................................................................................21
4OA1 Performance Indicator Verification ...........................................................................21
4OA2 Problem Identification and Resolution .....................................................................22
4OA3 Follow-Up of Events and Notices of Enforcement Discretion ...................................27
4OA6 Meetings, Including Exit...........................................................................................28
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-1
LIST OF ACRONYMS ........................................................................................................... A-10
3
SUMMARY
IR 05000443/2017004; 10/01/2017 to 12/31/2017; Seabrook; Licensed Operator Requalification
Program, Emergency Preparedness Drill Observation, and Follow-Up of Events and Notices of
This report covered a three-month period of inspection by resident inspectors and announced
baseline inspections performed by regional inspectors. The inspectors identified two non-cited
violations (NCVs) and one finding, all of which were of very low safety significance (Green).
The significance of most findings is indicated by their color (i.e., greater than Green, or Green,
White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance
Determination Process, dated October 28, 2016. Cross-cutting aspects are determined using
IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014. All violations of
NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated
August 1, 2016. The NRCs program for overseeing the safe operation of commercial nuclear
power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6.
Cornerstone: Initiating Events
- Green. A self-revealing Green finding was identified for inadequate implementation of
procedure MA 4.5, Configuration Control, Revision 18. Specifically, maintenance
technicians failed to properly implement MA 4.5 while backfilling steam generator
instrumentation, and inadvertently left an instrumentation valve partially open instead of fully
open. This resulted in slow response of the instrument, and ultimately a high steam
generator level, a feedwater isolation signal and a manual reactor trip. NextEra promptly
rechecked other similar valves, then performed a root cause evaluation that eventually led
to additional technician training and improved configuration controls during such evolutions.
This finding is more than minor because it is associated with the configuration control
attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit
the likelihood of events that upset plant stability and challenge critical safety functions
during shutdown as well as power operations. Specifically, the failure to effectively
implement MA 4.5 resulted in a valve being left out of its required position, a subsequent
lack of steam generator water level control during low power operations, and ultimately
required a manual reactor trip. In accordance with IMC 0609.04, Initial Characterization of
Findings, issued June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The Significance
Determination Process for Findings At-Power, issued June 19, 2012, the inspectors
determined that this finding is of very low safety significance (Green), because the finding
did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the
plant from the onset of a trip to a stable shutdown condition. Additionally, the finding has a
cross-cutting aspect in the area of Human Performance, Work Management, because the
organization did not implement a process of planning, controlling, and executing the work
activity such that nuclear safety was the overriding priority. Specifically, NextEra did not
ensure that a steam generator backfilling activity was properly executed, which resulted in
the slow response of a steam generator level indication, the overfeeding of the steam
generator, a feedwater isolation signal, and the ultimate requirement to trip the
reactor. [H.5] (Section 4OA3)
4
Cornerstone: Mitigating System
- Green. The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code
of Federal Regulations (10 CFR) 55.49, Integrity of Examinations and Tests, for the failure
of the licensee to ensure that the integrity of the written examinations administered to
licensed operators was maintained. During the planning of the biennial written
examinations, two written examinations would have exceeded the 50 percent overlap
criteria limit of questions administered in the previous four weeks of this examination cycle.
This failure resulted in a compromise of examination integrity because it exceeded the
NextEra Fleet Procedure TR-AA-220-1004, Licensed Operator Continuing Training Annual
Operating and Biennial Written Exams, Revision 2, requirement to repeat less than or
equal to 50 percent of the questions used during the exam cycle. However, this
compromise did not lead to an actual effect on the equitable and consistent administration
of the examination because of detection of this issue by the NRC prior to examination
administration. This issue was entered into NextEras Corrective Action Program (CAP) as
The failure of NextEras training staff to maintain the integrity of examinations administered
to licensed operations personnel was a performance deficiency. The performance
deficiency was more than minor, and therefore a finding, because if left uncorrected, the
performance deficiency could have become more significant in that allowing licensed
operators to return to the control room without valid demonstration of appropriate
knowledge on the biennial examinations could be a precursor to a more significant event.
Using IMC 0609, Significance Determination Process, and the corresponding Appendix I,
Licensed Operator Requalification Significance Determination Process, the finding was
determined to have very low safety significance (Green) because although the finding
resulted in a compromise of the integrity of written examination, the equitable and
consistent administration of the test was not actually impacted by this compromise. This
finding had a cross-cutting aspect in the area of Human Performance, Resources, in that
leaders ensure procedures are available and adequate to support nuclear safety.
Specifically, NextEra established and implemented a procedure that contained instructions
to licensed operator biennial exam writers that were unclear regarding regulatory guidance
to limit written examination questions overlap. [H.1] (Section 1R11.3)
Cornerstone: Emergency Preparedness
- Green. The inspectors identified a Green non-cited violation (NCV) of Title 10 of the
Code of Federal Reglations (10 CFR) 50.47(b)(14) and 10 CFR Part 50, Appendix E,
Emergency Planning and Preparedness for Production and Utilization Facilities,
Section IV.F.2.g. Specifically, Seabrook did not identify and critique a weakness associated
with a risk significant planning standard (RSPS) during their critique following the
August 30, 2017, emergency preparedness drill. The weakness involved the licensees
declaration of a general emergency (GE) that was based on insufficient information.
NextEra entered the issue into the corrective action program (CAP) as AR2242073.
The inspectors determined that not identifying an exercise weakness related to a GE
classification based on insufficient information during the exercise critique was a
performance deficiency that was reasonably within the ability of Seabrook to foresee and
prevent. The finding is more than minor because it is associated with the Emergency
Response Organization attribute of the Emergency Preparedness Cornerstone and affected
the cornerstone objective to ensure that the licensee is capable of implementing adequate
measures to protect the health and safety of the public in the event of a radiological
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emergency. Specifically, Seabrook personnel did not identify an exercise weakness
associated with a RSPS when the incorrect basis for a GE declaration was used by the Site
Emergency Director (SED). The finding was assessed using IMC 0609, Attachment 4,
Initial Characterization of Findings, issued October 7, 2016. This attachment directs
inspectors to utilize IMC 0609, Appendix B, Emergency Preparedness Significance
Determination Process, issued September 22, 2015, because the finding and
the associated weakness is in the licensees emergency preparedness cornerstone. The
inspectors determined the finding was a critique finding, the drill scope was full scale, the
planning standard was risk-significant, and the performance opportunity was a success
utilizing figure 5.14-1, Significance Determination for Critique Findings, and thus
determined this finding was of very low safety significance (Green). The finding was
determined to have a cross-cutting aspect in the area of Human Performance, Change
Management, in that leaders use a systematic process for evaluating and implementing
change so that nuclear safety remains the overriding priority. Specifically, although recent
changes to the sites emergency classification and action level standard scheme were
effective on July 2017, the new EAL procedure and training regarding the changes lacked
sufficient specificity to ensure the users understood the new scheme with respect to the
status of the containment integrity. [H.3] (Section 1EP6)
6
REPORT DETAILS
Summary of Plant Status
Seabrook began the inspection period at full power, and there were no plant status changes of
regulatory significance during the remainder of the inspection period. Documents reviewed for
each section of this inspection report are listed in the Attachment.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01 - 1 samples)
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) pump house, the cooling
water tower (CWT) pump area, and portions of the turbine building that contains risk
important systems. 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 readiness
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.
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 emergency core cooling system (ECCS) during maintenance on the B charging
pump and safety injection pump on November 6
- Boric acid flow paths during maintenance on the boric acid control station on
November 8-9
- B fire pump during A fire pump maintenance on December 14
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
7
(WOs), condition reports (CRs), and the impact of ongoing work activities on redundant
trains of equipment in order to identify conditions that could have impacted 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 the period of November 27 through December 1, the inspectors performed a
complete system walkdown of accessible portions of the SW system to verify the
existing equipment lineup was correct. The inspectors reviewed operating procedures,
system diagrams, 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 as-built system configuration matched plant
documentation, and that system components and support equipment remained
operable. The inspectors confirmed that systems and components were aligned
correctly, free from interference from temporary services or isolation boundaries,
environmentally qualified, and protected from external threats. The inspectors also
examined the material condition of the components for degradation and observed
operating parameters of equipment to verify that there were no deficiencies.
Additionally, the inspectors reviewed a sample of related CRs and WOs to ensure
NextEra appropriately evaluated and resolved any deficiencies.
b. Findings
No findings were identified.
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
8
station personnel implemented compensatory measures for out of service, degraded, or
inoperable fire protection equipment, as applicable, in accordance with procedures.
- Primary auxiliary building (PAB) southeast corner (PAB-F-2A-Z) on December 20
- PAB boric acid tanks and sample sink rooms (PAB-F-2B-Z) on December 20
- PAB primary component cooling water (PCCW) pump area (PAB-F-2C-Z) on
December 20
- PAB PCCW heat exchangers (PAB-F-3A-Z) on December 20
- PAB SW pipe slot (PAB-F-1K-Z) on December 20
b. Findings
No findings were identified.
1R06 Flood Protection Measures (71111.06 - 1 sample)
Internal Flooding Review
a. Inspection Scope
The inspectors reviewed the UFSAR, site flooding analysis, and plant procedures to
identify internal flooding susceptibilities for the site. The inspectors review focused on
the B residual heat removal (RHR) vault to verify the adequacy of equipment seals
located below the flood line, floor and wall penetration seals, watertight door seals,
common drain lines and sumps, sump pumps, level alarms, control circuits, and
temporary or removable flood barriers. The inspectors assessed the adequacy of
operator actions that NextEra had identified as necessary to cope with flooding in this
area and also reviewed the CAP to determine if NextEra was identifying and correcting
problems associated with both flood mitigation features and site procedures for
responding to flooding.
b. Findings
No findings were identified.
1R07 Heat Sink Performance (711111.07A - 1 sample)
a. Inspection Scope
The inspectors reviewed the A and B RHR heat exchanger to ensure readiness and
availability. The inspectors conducted a walkdown of the heat exchangers and reviewed
the results of the most recent performance test. The inspectors verified that NextEra
initiated appropriate corrective actions for identified deficiencies.
b. Findings
No findings were identified.
9
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
.1 Quarterly Review of Licensed Operator Requalification Testing and Training
(71111.11Q - 1 sample)
a. Inspection Scope
The inspectors observed licensed operator simulator annual requalification exams on
November 7, 2017, which included various failures, a transient resulting in an anticipated
transient without a scram, and a faulted steam generator requiring safety injection.
Another scenario included losing a feedwater pump, requiring a reactor scram, followed
by a loss of offsite power/loss-of-coolant accident. The inspectors evaluated operator
performance during the simulated event and verified completion of risk significant
operator actions, including the use of abnormal and emergency operating procedures.
The inspectors assessed the clarity and effectiveness of communications,
implementation of actions in response to alarms and degrading plant conditions, and the
oversight and direction provided by the control room supervisor. Additionally, the
inspectors assessed the ability of the crew and training staff to identify and document
crew performance problems.
b. Findings
No findings were identified.
.2 Quarterly Review of Licensed Operator Performance in the Main Control Room
(71111.11Q - 1 sample)
a. Inspection Scope
On October 19, 2017, the inspectors observed and reviewed routine activities in the
main control room. The inspectors observed operators respond to alarms, complete a
reactor coolant system (RCS) dilution, conduct a pre-job briefing for a surveillance test,
and perform the surveillance test. Additionally, the inspectors verified that procedure
use, crew communications, and coordination of activities between work groups met
established expectations and standards.
b. Findings
No findings were identified.
.3 Licensed Operator Requalification (71111.11A - 1 sample, 71111.11B - 1 sample)
a. Inspection Scope
The following inspection activities were performed using NUREG-1021, Operator
Licensing Examination Standards for Power Reactors, Revision 11, and Inspection
Procedure 71111.11, Licensed Operator Requalification Program.
10
Examination Results
On December 26, 2017, the results of the annual operating tests and biennial written
examinations were reviewed to determine if pass/fail rates were consistent with the
guidance of NUREG-1021, and NRC IMC 0609, Appendix I, Operator Requalification
Human Performance Significance Determination Process. The review verified that the
failure rate (individual or crew) did not exceed 20 percent.
- Five out of 42 operators failed at least one portion of requalification examination
(written, job performance measures (JPMs) or individual scenario failures). The
overall individual failure rate was 11.9 percent.
- One out of eight crews failed the simulator test. The crew failure rate was
12.5 percent
Written Examination Quality
The inspectors reviewed the written examinations administered to reactor operators
(ROs) and senior reactor operators (SRO) during the weeks 2, 4, and 5 of this cycle
(November-December 2017) for qualitative and quantitative attributes as specified in
Appendix B of Attachment 71111.11,
Operating Test Quality
Ten JPMs and five scenarios were reviewed for qualitative and quantitative attributes as
specified in Appendix C of 71111.11.
Licensee Administration of Operating Tests
Observations were made of the dynamic simulator exams and JPMs administered during
the week of December 4, 2017. These observations included facility evaluations of crew
and individual performance during the dynamic simulator exams and individual
performance of JPMs.
Examination Security
The inspectors assessed whether facility staff properly safeguarded exam material. The
JPMs, scenarios, and written examinations were checked for excessive overlap of test
items.
Remedial Training and Re-Examinations
The inspectors reviewed remediation plans and examinations for one crew failure during
the first quarter of 2016.
Conformance with Operator License Conditions
Medical records for six SRO licenses and four RO licenses were reviewed to assess
conformance with license conditions. All records reviewed were satisfactory.
11
Proficiency watch standing records for licensed operators were reviewed for the first
three quarters of 2017. All active licensed operators met the watch standing
requirements to maintain an active license.
The reactivation plan for licensed operators (three ROs and 13 SROs) were reviewed to
assess the effectiveness of the reactivation process. The reactivation was successfully
processed in accordance with site procedures.
Records for the participation of licensed operators in the requalification program for the
first three quarters in 2017 were reviewed.
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, the licensees CAP, and the most recent NRC plant issues matrix. The
inspectors also reviewed specific events from the licensees CAP which indicated
possible training deficiencies, to verify that they had been appropriately addressed.
These reviews did not detect any operational events that were indicative of possible
training deficiencies.
b. Findings
Introduction. The inspectors identified a Green NCV of 10 CFR 55.49, Integrity of
Examinations and Tests, for NextEras failure to ensure the integrity of the biennial
written examinations that were to be administered to licensed operators. This would
have resulted in examining Seabrook licensed operators with questions that had been
administered to other crews during the exam cycle that were in excess of the limits
established for question overlap.
Description. On December 6, 2017, while performing a biennial inspection in
accordance with IP 71111.11, Licensed Operator Requalification Program, the
inspectors determined that the written examination that was planned to be administered
that day for Crew E (and for Crew F in the following week) contained more than
50 percent of questions that had been used cumulatively to the licensed operators in the
previous 4 weeks of the same exam cycle.
NextEra Fleet Procedure TR-AA-220-1004, Licensed Operator Continuing Training
Annual Operating and Biennial Written Exams, Revision 2, requires that, Each biennial
comprehensive written exam version shall consist of at least 50 percent new, different, or
significantly modified test items compared to all previously administered versions of the
same exam. Since the procedure was not clear regarding the intent of this requirement,
the licensee incorrectly applied this to mean that there could be no more than 50 percent
overlap of questions in any one weeks examination with any other weeks examination
questions. In other words, the licensee was applying the question overlap criteria from
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examination to examination instead of applying it to the cumulative usage of questions in
the entire cycle. By applying their overlap criteria as they did, in conjunction with how
they selected the questions to be used on each examination, the examinations for Crews
E and F would have had 30 of 33 questions that had been previously used in this cycle.
According to 10 CFR 55.49, the integrity of a test or examination is considered
compromised if any activity, regardless of intent, affected or, but for detection, could
have affected the equitable and consistent administration of the test or examination. The
inspectors concluded that exceeding the 50 percent overlap was a failure to fulfill the
requirements of NextEras procedure and constituted a compromise of examination
integrity required by 10 CFR 55.49.
The inspectors informed the licensee of this overlap issue prior to the administration of
the written examination to Crew E. The licensee then postponed this written
examination until they could develop a written examination that did not violate the
overlap requirement. The first four written examinations of this 2017 cycle did use
common questions, but did not exceed the 50 percent overlap limit. Thus, there was no
actual effect on the equitable and consistent administration of the written examinations.
(Furthermore, the licensee has operators sign a security agreement to not reveal any
information about the requalification examinations with other operators who have not yet
taken their examinations.) During the previous comprehensive written examination in
2015, the examination developer used unique questions for each of the examinations in
that cycle. This years comprehensive written examination was developed by a different
individual who, along with other fleet personnel, misapplied the fleet procedures overlap
criteria. The licensee entered this issue into their CAP as AR 2239906.
Analysis. The failure of NextEras training staff to ensure the integrity of examinations
administered to licensed operations personnel was a performance deficiency. The
performance deficiency was a finding that was more than minor because, if left
uncorrected, the performance deficiency had the potential to lead to a more significant
safety concern. Specifically, the potential to allow operators to return to the control room
without valid demonstration of appropriate knowledge on the biennial written
examinations could result in having less than adequately qualified operators
manipulating plant controls in response to events. Using IMC 0609, Significance
Determination Process, and the corresponding Appendix I, Licensed Operator
Requalification Significance Determination Process, the finding was determined to have
very low safety significance (Green) because, although the examinations were not
administered, the integrity of an examination is considered to be compromised if any
activity affected, or but for detection, would have affected the equitable and consistent
administration of the examination. This finding had a cross-cutting aspect in the area of
Human Performance, Resources, in that leaders ensure procedures are available and
adequate to support nuclear safety. Specifically, NextEra established and implemented
a procedure that contained instructions to licensed operator biennial exam writers that
were unclear regarding regulatory guidance to limit written examination questions
overlap. [H.1]
Enforcement. Title 10 CFR 55.49, Integrity of Examinations, requires, in part, that
facility licensees shall not engage in any activity that compromises the integrity of any
test or examination required by this part. The integrity of a test or examination is
considered compromised if any activity, regardless of intent, affected or, but for
detection, could have affected the equitable and consistent administration of the test or
examination. This includes activities related to the preparation, administration, and
13
grading of the tests and examinations required by this part. Contrary to the above,
during the 2017 annual examination cycle (November through mid-December), NextEra
engaged in an activity at Seabrook that compromised the integrity of a test required by
10 CFR Part 55. Specifically, two scheduled written examinations would have contained
more than 50 percent of questions previously used in the cycle but for detection by the
NRC. Administering a written examination with greater than 50 percent cumulative
overlap from previously administered questions during a cycle is considered a
compromise of the integrity in that it is a practice that, but for detection, could affect the
equitable and consistent administration of the examination. The inspectors determined
that this overlap issue did not result in an actual effect on the equitable and consistent
administration of the written examinations. Because this finding was of very low safety
significance (Green) and has been entered into NextEras CAP as AR 2239906, this
violation is being treated as an NCV consistent with Section 2.3.2.a of the NRC
Enforcement Policy. (NCV 05000443/2017004-01 Licensed Operator Examination
Integrity Not Ensured)
1R12 Maintenance Effectiveness (71111.12Q - 1 sample)
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
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 was reasonable. As applicable, for
SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective
actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that NextEra
staff was identifying and addressing common cause failures that occurred within and
across MR system boundaries.
- Boric acid control station
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 4 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
14
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.
- Switchyard work, startup feed pump testing, and A emergency diesel generator
(EDG) maintenance and testing on October 17
- CS-FCV-111B fail to open during the period November 1-4
- Switchyard work and supplemental emergency power system maintenance on
November 20
- B solid state protection system Mode 1 actuation logic test on November 27
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments (71111.15 - 2 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:
- A EDG fuel oil return line leaks on October 16
- D vital DC battery abnormal ammeter reading on November 15
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. Where compensatory measures
were required to maintain operability, the inspectors determined whether the measures
in place would function as intended and were properly controlled by NextEra.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19 - 6 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
15
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 point were performed and checked, and
that results adequately demonstrated restoration of the affected safety functions.
- B CWT bistable card replacement on October 2
- C SW pump instantaneous overcurrent relay set point adjustment on October 16
- RC-V-2832, RCS sample valve relay replacement on November 2
- CS-FCV-111-B repairs on November 4
- Limitorque maintenance for CC-V-266 on November 28
- A fire pump annual maintenance on December 14
b. Findings
No findings were identified.
1R22 Surveillance Testing (71111.22 - 3 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:
- Power range channel 44 resealing calibration on October 4
- A PCCW pump on October 19 (in-service test)
- B charging pump surveillance on October 20
b. Findings
No findings were identified.
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Cornerstone: Emergency Preparedness
1EP6 Drill Evaluation (71114.06)
.1 Emergency Preparedness Drill Observation (2 samples)
a. Inspection Scope
The inspectors evaluated the conduct of routine NextEra emergency drills on August 30,
2017, and November 29, 2017, to identify any weaknesses and deficiencies in the
classification, notification, and protective action recommendation development activities.
The inspectors observed emergency response operations in the simulator, technical
support center, and emergency operations facility (EOF) to determine whether the event
classification, notifications, and protective action recommendations were performed in
accordance with procedures. The inspectors also attended the station drill critique to
compare inspector observations with those identified by NextEra staff in order to
evaluate NextEras critique and to verify whether the NextEra staff was properly
identifying weaknesses and entering them into the CAP.
b. Findings
Introduction. The inspectors identified a Green NCV of 10 CFR 50.47(b)(14) and
10 CFR Part 50, Appendix E, Emergency Planning and Preparedness for Production
and Utilization Facilities,Section IV.F.2.g. Specifically, Seabrook did not identify and
critique a weakness associated with a RSPS during their critique following the
August 30, 2017, emergency preparedness drill.
Description. On August 30, 2017, Seabrook conducted an emergency preparedness
exercise, which included activating the simulator control room, the technical support
center (TSC), the operational support center, and the EOF. Consistent with the exercise
scenario script, a seismic event caused an RCS leak of approximately 300 gallons per
minute and resulted in the actuation of safety injection. The SED, located in the
simulator control room responded appropriately by declaring an emergency action level
(EAL) of Alert at 8:29 a.m. because the loss of the RCS barrier (loss of single fission
product barrier) threshold criterion was met. At 10:15 a.m., a containment release was
prematurely introduced by the simulator operator. This release was indicated on the
plant stack wide range gas monitor (WRGM) and the containment enclosure ventilation
area (CEVA) radiation monitor. The drill controllers recognized the error but did not
interject and allowed the events surrounding the premature simulated release to play
out.
At 10:19 a.m., consistent with the exercise scenario script, a second seismic event
occurred that resulted in a large break loss of coolant accident (LOCA). Plant conditions
deteriorated to the point that all ECCS necessary to inject for subsequent core cooling
had failed. This plant condition met the threshold for a potential loss of the fuel clad
barrier from a valid core cooling orange entry condition. The combination of the prior
loss of RCS barrier and the potential loss of the fuel clad barrier met the criteria for
classifying the event as a site area emergency (SAE). However, the SED, located in the
TSC, did not declare a SAE, but declared a GE at 10:28 a.m. The typical threshold for
declaring a GE is the loss of two barriers and the potential loss of the third. The SEDs
basis for concluding that the GE classification threshold criteria were met was the loss of
17
the RCS barrier, the potential loss of the fuel clad barrier, and the loss of the
containment barrier. The SED determined that a loss of containment barrier occurred
based on an unisolable pathway; however, no open pathway was scripted in the
exercise scenario and there were no valid indications that this was the case. The SED
concluded that the containment barrier was unisolable even though the radiological
release data at the time was well below the Environmental Protection Agencys (EPAs)
protective action guide (PAG) levels that are incorporated in the emergency plan.
Seabrooks emergency plan directs the comparison of radiological release data with
EPAs PAGs to inform decision making regarding whether a loss of containment barrier
exists.
Seabrook completed their formal drill critique on September 19, 2017. During the
critique, Seabrook did not identify that the declaration by the SED of a GE with protective
action recommendations (PARs) was based on insufficient information. Specifically,
Seabrooks EAL for the loss of the containment barrier is driven by a containment
isolation being required and either of the following: 1) Containment integrity has been
lost based on Short-Term Emergency Director (STED)/SED judgment, or 2) an
unisolable pathway from the containment to the environment exists. ER 1.1,
Classification of Emergencies, Revision 58, defines unisolable, as an open or breached
system line that cannot be isolated, remotely or locally.
Following the formal drill critique on September 19, 2017, the inspectors questioned the
basis for considering containment integrity lost, which resulted in characterizing the
circumstances present as a loss of the containment barrier. NextEra indicated that the
SED considered the loss of containment integrity was due to containment isolation being
required and the existence of an unisolable pathway from the containment to the
environment. NextEra noted that a containment isolation signal was received as
expected and all available remote indications showed the containment isolation valves
were closed. There were no other confirmed pathways open from containment to the
environment. The licensee also confirmed that containment pressures and pressure
trends were indeterminate with respect to the status of containment integrity. The
licensee validated after the exercise that the higher than normal WRGM readings
indicated noble gases that could only come from damaged nuclear fuel inside
containment; however, the containment post-LOCA radiation monitors were reading
relatively normal with no indication of damaged fuel.
As planned by the exercise scenario script, a containment recirculation sump isolation
valve CBS-V-8, was not opening when required, to place the containment on
recirculation cooling, which led the SED to suspect the penetration and its encapsulated
valve were the possible locations of an unisolable pathway. This determination by the
SED is noteworthy, because the control room operators had confirmed that the valve
was closed based on remote indication. The status of this valve, encapsulation tank,
and penetration line were not validated locally. Taking into account the seismic events
that caused the large break LOCA, the suspect encapsulated valve, higher than normal
WRGM readings and CEVA radiation levels, the SED concluded that a loss of
containment integrity, as defined in their EAL scheme and basis, existed. A GE
declaration was made due to the loss of containment conclusion and the previously
determined potential loss of fuel clad and the loss of the RCS barrier (versus the
originally scripted SAE). Due to the lack of any other valid indications that containment
integrity was jeopardized, the SED relied upon the radiological releases seen on the
WRGM and CEVA radiation monitor as positive indication of the loss of the containment
18
barrier. The fission product barrier EAL (FG1) allows the SED to use judgment to make
a determination of containment barrier integrity based on less discrete information.
Specifically, 4.A.1 states that containment integrity has been lost when the actual
containment atmospheric leak rate likely exceeds that associated allowable leakage.
However, Seabrooks procedure, ER 1.1 states, it is expected that the SED will assess
the threshold using judgment, and with due consideration given current plant conditions,
and available operational and radiological data.
The inspectors determined that, as a result of the deviation from the preplanned
scenario script and due to the actual condition experienced during the exercise, the GE
declaration would have been an appropriate event classification if it had been based on
SED judgment instead of an unisolable pathway. The conditions presented at the time
could have warranted the use of judgement to escalate from an SAE to a GE based on
imminent fuel melt and the uncertainty recognized by the SED, regarding the fuel
condition based on radiation monitors indicating a release outside the containment.
Therefore, the GE threshold criteria (loss of two and the potential loss of the third fission
product barriers) would have been met by the loss of the RCS barrier, the potential loss
of the containment barrier and judgement that the loss of the fuel clad barrier was
imminent.
As a result, in accordance with IMC-0609, Appendix B, Emergency Preparedness
Significance Determination Process, the performance demonstrated by NextEra
participants in the drill, provided specific opportunities that could preclude effective
implementation of the emergency plan that the inspectors concluded was a weakness.
In addition, the inspectors also identified deficiencies associated with the Emergency
Classification system RSPS under 10 CFR 50.47(b)(4). These deficiencies involved the
less than adequate translation of specific guidance incorporated into the Seabrook EAL
basis document during implementation of a recent upgrade to the Seabrook emergency
plan to incorporate a revision (5 to 6) to Nuclear Energy Institute (NEI) Document 99-02,
Development of Emergency Action Levels for Non-Passive Reactors. Moreover, the
inspectors determined that the requisite training for decision-makers for the most
relevant portion of the revised guidance, was also developed and provided in a less than
adequate manner. More importantly, the germane sections of the revised guidance
associated with the Containment Barrier portion of the Fission Product Matrix EALs were
directly exercised during the August 2017 drill.
Analysis. The inspectors determined that not identifying an exercise weakness related
to a GE classification based on insufficient information during the exercise critique was a
performance deficiency that was reasonably within the ability of Seabrook to foresee and
prevent. The finding is more than minor because it is associated with the ERO attribute
of the Emergency Preparedness Cornerstone and affected the cornerstone objective to
ensure that the licensee is capable of implementing adequate measures to protect the
health and safety of the public in the event of a radiological emergency. Specifically,
Seabrook personnel did not identify an exercise weakness associated with a RSPS
when the incorrect basis for a GE declaration was used by the SED.
The inspectors assessed the finding using IMC 0609, Attachment 4, Initial
Characterization of Findings, issued October 7, 2016. This attachment directs
inspectors to use IMC 0609, Appendix B, Emergency Preparedness Significance
Determination Process, issued September 22, 2015, because the finding and the
19
associated weakness are in the emergency preparedness cornerstone. Inspectors
determined the finding was a critique finding, the drill scope was full scale, the planning
standard was risk-significant, and the performance opportunity was a success. As a
result, and using figure 5.14-1, Significance Determination for Critique Findings, the
inspectors determined this finding was of very low safety significance (Green).
The finding is related to the cross-cutting area of Human Performance, Change
Management in that leaders use a systematic process for evaluating and implementing
change so that nuclear safety remains the overriding priority. Specifically, although
recent changes to the sites emergency classification and action level standard scheme
were effective on July 2017, the new EAL procedure and training regarding the changes
lacked sufficient specificity to ensure the users understood the new scheme with respect
to the status of the containment integrity [H.3].
Enforcement. Title 10 CFR 50.54(q)(2) requires, in part, that a licensee shall follow and
maintain the effectiveness of an emergency plan that meets the requirements in
Appendix E to this part and, for nuclear power reactor licensees, the planning standards
of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(14) requires, in part, that periodic exercises
be conducted to evaluate major portions of emergency response capabilities and that
deficiencies identified as a result of exercises are corrected. Section lV.F.2.g of
Appendix E to 10 CFR Part 50 requires that all training, including exercises, shall
provide for formal critiques in order to identify weak or deficient areas that need
correction. Any weaknesses or deficiencies that are identified shall be corrected.
Contrary to the above, during a formal critique on September 19, 2017, Seabrook did not
identify a weakness needing correction that was demonstrated during a full participation
exercise on August 30, 2017. The weakness needing correction involved NextEras
declaration of a GE that was based on insufficient information. Because this violation
was of very low safety significance and was entered into Seabrooks CAP as
AR 2242073, this finding is being treated as an NCV consistent with Section 2.3.2 of the
NRC Enforcement Policy. (NCV,05000443/2017004-02, Failure of Exercise Critique
to Identify a Risk Significant Planning Standard Weakness)
.2 Training Observations (1 sample)
a. Inspection Scope
The inspectors observed a simulator training evolution for licensed operators on
November 7, 2017, which required emergency plan implementation by an operations
crew. NextEra planned for this evolution to be evaluated and included in the drill and
exercise performance indicator (PI) data. The inspectors observed event classification
and notification activities performed by the crew. The focus of the inspectors activities
was to note any weaknesses and deficiencies in the crews performance and ensure that
NextEra evaluators noted the same issues and entered them into the CAP.
b. Findings
No findings were identified.
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2. RADIATION SAFETY
Cornerstone: Public Radiation Safety
2RS2 Occupational As Low As Is Reasonably Achievable Planning and Controls
(71124.02 - 1 sample)
a. Inspection Scope
The inspectors assessed NextEras performance with respect to maintaining
occupational individual and collective radiation exposures as low as is reasonably
achievable (ALARA). The inspectors used the requirements contained in 10 CFR
Part 20, applicable Regulatory Guides (RGs) 8.8 and 8.10, TSs, and procedures
required by TSs as criteria for determining compliance.
Verification of Dose Estimates and Exposure Tracking Systems
The inspectors reviewed the current annual collective dose estimate; basis methodology;
and measures to track, trend, and reduce occupational doses for ongoing work activities.
The inspectors evaluated the adjustment of exposure estimates, or re-planning of work.
The inspector reviewed post-job ALARA evaluations of excessive exposure.
b. Findings
No findings were identified.
2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03 - 1 sample)
a. Inspection Scope
The inspectors reviewed the control of in-plant airborne radioactivity and the use of
respiratory protection devices in these areas. The inspectors used the requirements in
10 CFR Part 20, RG 8.15, RG 8.25, NUREG/CR-0041, TS, and procedures required by
TS as criteria for determining compliance.
Self-Contained Breathing Apparatus for Emergency Use
The inspectors reviewed the following: the status and surveillance records for three
Self-Contained Breathing Apparatus (SCBAs) staged in-plant for use during
emergencies; Next Eras SCBA procedures and maintenance and test records; the
refilling and transporting of SCBA air bottles; SCBA mask size availability; and the
qualifications of personnel performing service and repair of this equipment.
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 July 1, 2017, through June 30, 2018:
- Safety System Functional Failures
- RHR System
- Cooling Water System
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, 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 licensee submittals for the occupational radiological
occurrences PI for the fourth quarter 2016 through the first, second, and third quarters
2017. The inspectors used PI definitions and guidance contained in the NEI Document
99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to
determine the accuracy of the PI data reported. The inspectors reviewed 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.
b. Findings
No findings were identified.
.3 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual
Radiological Effluent Occurrences (1 sample)
a. Inspection Scope
The inspectors reviewed licensee submittals for the radiological effluent technical
specifications/offsite dose calculation manual radiological effluent occurrences PI for the
fourth quarter 2016 through the first, second, and third quarters of 2017. The inspectors
22
used PI definitions and guidance contained in the NEI Document 99-02, Regulatory
Assessment Performance Indicator Guideline, Revision 7, to determine if the PI data
was reported properly. The inspectors reviewed the 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.
4OA2 Problem Identification and Resolution (71152 - 3 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
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 to identify trends that
might indicate the existence of more significant safety concerns. As part of this review,
the inspectors included repetitive or closely-related issues documented by NextEra in
quarterly trend reports, site 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 2017 to assess CRs written
in various subject areas (equipment problems, human performance issues, etc.), as well
as individual issues identified during the NRCs daily CR review (Section 4OA2.1). The
inspectors reviewed the NextEra trend reports for the previous six months of 2017,
conducted under PI-AA-207-1000, Station Self-Evaluation and Trend Analysis,
23
Revision 8, to verify that NextEra personnel were appropriately evaluating and trending
adverse conditions in accordance with applicable procedures.
b. Findings and Observations
No findings were identified.
Overall, the inspectors noted that the system health reports for the safety related
systems and systems important to safety to be up to date and reflective of current plant
status. The health reports were reflective of issues that were trending on the daily plant
status report and discussed on a regular basis by plant management for timely
resolution. The inspectors evaluated a sample of CRs generated over the course of the
past two quarters by departments that provide input to the quarterly trend reports. The
inspectors determined that, in most cases, the issues were appropriately evaluated by
Seabrook staff for potential trends and resolved within the scope of the CAP. Moreover,
the inspectors identified instances where potential adverse trends were identified by
department staff during the course of the assessment period, which were consistent with
similar station-level trends, and confirmed that station personnel were utilizing statistical
and trending tools to identify potential emerging trends. Additionally, the inspectors
verified that discussions between department and performance improvement staff were
occurring to ensure emerging trends were appropriately captured either in the CAP or
the quarterly trend report, as applicable. One such example was an issue with the
overall health of the preventive maintenance program, which included implementation
and knowledge issues following a program assessment documented under CR 2219903.
.3 Annual Sample: Ultimate Heat Sink
a. Inspection Scope
The inspectors performed an in-depth review of NextEras evaluations and corrective
actions associated with the ultimate heat sink over the last year, which includes the
ocean SW system, CWT, and PCCW system. This included degraded piping and leaks,
PCCW pump motor issues, and increasing SW pump motor winding temperatures.
The inspectors assessed NextEras problem identification threshold, cause analyses,
extent of condition reviews, compensatory actions, and the prioritization and timeliness
of NextEras 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 Part 50,
Appendix B.
b. Findings and Observations
No findings were identified.
NextEra was timely in documenting issues once they were identified and screened
appropriately for immediate operability concerns. For example, control room operators
noted an increased trend in SW pump motor winding temperatures. It did not
immediately impact the safe operation of the plant, but the issue was captured in the
CAP and the motors were systematically replaced in a timely manner.
24
An outstanding issue continues to be degraded SW piping associated with the ocean
SW and the cooling water systems. NextEra has a systematic program, reflected in
PEG-94, Service Water Inspection and Repair Trending, to ensure that long term
corrective actions are implemented to minimize unexpected leaks and challenges to the
safe operation of the plant. The inspectors verified that PEG-94 is continuously updated,
and pipe inspections and replacements are completed as scheduled. When unexpected
leaks did occur, the station demonstrated timely assessment and appropriate
compensatory measures until final corrective actions to restoration were feasible.
The inspectors noted that NextEra implemented industry initiatives to improve the
effectiveness of issue resolution, also known as CAP-002, in August 2017. The changes
are reflected in PI-AA-104-1000, Condition Reporting. The inspectors have been
closely monitoring the impact to ensure issues important to nuclear safety are addressed
appropriately. No concerns have been noted by the inspectors to date.
.4 Annual Sample: Alkali-Silica Reaction
a. Inspection Scope
The purpose of periodic site visits to Seabrook Station over the past few years has been
to review the adequacy of NextEras monitoring of alkali-silica reaction (ASR) on affected
reinforced concrete structures, per their 10 CFR 50.65 Maintenance Rule Structures
Monitoring Program (SMP), and NextEras corrective action process. In addition, the
inspectors verify on a sampling basis that significant changes or different manifestations
of ASR on the affected structures are appropriately considered for impact on the
Seabrook prompt operability determinations for the affected structure(s). Two NRC
region-based inspectors and a structural engineer from the Office of Nuclear Reactor
Regulation were on site from October 10-13, 2017, to conduct an inspection of ongoing
ASR related activities. The inspectors also conducted in-office reviews of ASR-related
documentation made available before and after the on-site inspection via an electronic
server (Certrec Inspection Management System). Although available for review, the
inspectors did not receive or take possession of these documents.
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 structures. The inspectors evaluated NextEras actions to verify
compliance with the SMP, the CAP, and 10 CFR Part 50, Appendix B requirements.
b. Findings and Observations
No findings were identified.
The inspectors performed a review of the CEVA north wall operability determination,
including a field walkdown of the structure. The North wall is laterally deformed below
the CEVA heating, ventilation, and air conditioning (HVAC) room floor slab as measured
by the plumbness. NextEra has preliminarily concluded the movement at this location is
the result of ASR expansion of the concrete backfill confined between the wall and the
adjacent bedrock, which is a load that was not considered in the original design of the
25
wall in accordance with American Concrete Institute (ACI) 318-71. The out-of-plumb
wall section is located between the +3 and +19 foot elevation and exhibits visual
horizontal flexure cracks with evidence of delamination (identified via hammer testing) in
the vicinity of the cracks. The cracks are spaced at approximately 1 foot intervals, which
is the same spacing as the horizontal reinforcing bars. The detected delaminations were
found around the horizontal cracks where the largest displacement is occurring on the
order of approximately 1.5 inches. An initial SMP structural evaluation by NextEra staff
(simple beam finite element analysis) was performed, and with the estimated
compressive strains in the concrete in some areas and the opposing tensile strains in
the rebar in other sections, the analysis concluded that delamination is predicted.
Subsequently, a nonlinear finite element analysis based on the deformed shape of the
wall was performed by NextEra to determine the maximum allowable lateral
displacement before a modification is necessary. The inspectors reviewed this analysis
as part of the operability determination and determined that NextEras conclusions that
the structure is capable of performing its intended functions was technically supported.
The inspectors further verified that SMP Appendix C was updated with additional
qualitative monitoring requirements for the CEVA building. Discussions with the
responsible NextEra engineering staff identified that remediation methods are being
evaluated to ensure long-term continued stabilization and structural performance of the
wall. The inspectors noted that this lower portion of the north wall was identified as a
non-structural member for the CEVA structure (i.e., not part of the structural load
resisting system for the CEVA) and is not part of the boundary that establishes the
safety-related CEVA air envelope. However, the wall is required to maintain its
structural stability because it supports attached equipment.
Inspectors walkdown of the RHR/containment spray (CS) Vault confirmed the presence
of several small areas of delamination. Review of FP101055, Condition Assessment of
Cracking in RHR and CS Equipment Vault - Second Visit, dated February 4, 2016,
summarizes the results of a detailed examination of the RHR/CS Vault by NextEras staff
contractors following an earlier examination in December 2014. One of the
recommendations in FP101055 was to remove cores from areas exhibiting delamination
to better understand the extent of concrete degradation. At the request of the
inspectors, NextEra posted the results of concrete coring and associated petrographic
examination (FP101034) on their electronic server (Certrec Inspection Management
System) for review. FP101034 summarizes the petrographic examination of 19 core
samples and their associated bore holes. The examination results identified that all of
the cores taken from the external walls exhibited signs of ASR, whereas the cores taken
from the interior walls did not. The large cracks observed in the interior walls were likely
a result of upward expansion due to ASR in the exterior walls, which transferred the
resulting tension to the interior walls of the Equipment Vault. The inspectors noted that
there were no discussions on the surface delamination areas or confirmation of the
depth of delamination as was recommended in earlier reports.
The identification of delamination as either a primary (caused by internal ASR expansion
in the wall) or secondary (caused by ASR expansion of concrete backfill and associated
loading) effect of ASR is preliminarily being reviewed by the NRC inspectors as a
phenomenon associated with ASR based on plant operating experience. At the
conclusion of the on-site inspection, NextEra staff had not drawn conclusions regarding
the implications of delamination associated with ASR expansion and loading. Based
upon the inspectors initial assessment, NextEra decided to develop criteria for
identifying and monitoring delamination of ASR-affected structures and how best to use
26
hammer testing or other non-destructive examination methods (e.g., impact-echo
testing), which was captured as an action in their Change Management Plan for the
SMP. The SMP currently does not describe hammer testing or include delamination
monitoring guidance, and NextEra had not specifically identified this ASR phenomenon
in the structures Aging Management Program for their license renewal application.
On November 22, 2017, NextEra provided the inspectors with an assessment of
ASR-related delamination, to date, that concluded the delamination areas were a result
of loading on the wall and were limited to the cover concrete layer (near surface), and
therefore, not relevant to structural performance. NextEra staff planned to perform
impact-echo testing, a non-destructive test method that uses sound waves to detect
flaws within the concrete, to verify that delamination is only occurring in the cover
concrete. If delaminations deeper than the cover are identified, then NextEra staff
indicated that cores would be taken to verify the condition of the concrete. The
inspectors determined that this proposed validation plan was technically adequate to
assess the implications of delamination.
Consistent with the current SMP, the B Electrical Tunnel Stage 1 structural evaluation
was recently completed. NextEra staff concluded that by including an assumed ASR
loading from the concrete backfill in the building design shear capacity calculations, the
calculated electrical tunnel wall loading (assumed demand) exceeds the design capacity
and would not conform to established standards in the ACI 318-71 structural design
code. To address this non-conforming condition, NextEra wrote a separate operability
determination and initiated further engineering evaluations to review the ASR backfill
loading assumptions and to consider potential remediation methods for the B Electrical
Building, including support struts and/or bolted plates. The inspectors noted that there
are no visual indications of loading distress or other structural integrity issues as evident
by the absence of structural cracks. The inspectors conducted a conference call with
NextEra staff and their principle ASR engineering contractor (SG&H) on October 18,
2017, to better understand the assumed backfill loading profiles used by NextEra staff in
the structural evaluations. The inspectors were informed that the concrete backfill
loading profiles differ for each Seabrook structure and that these profiles were
developed by a seven step iterative process. Based upon this conference call, the
inspectors understand that NextEra staff used as-built drawings with backfill details to
develop the initial ASR load profiles, taking into consideration whether or not the
concrete backfill was confined or unrestrained by any overburden or adjoining excavated
surfaces. If appropriate, the backfill load profile adjustments were made utilizing field
observations. Examination of NextEras methodology for assessing concrete backfill
loading is currently under review by the NRC staff, as an element of the August 1, 2016,
License Amendment Request (16-03).
Based upon discussions with the responsible engineering staff and inspector review of
the Structures Monitoring Program Manual (SMPM), the inspectors understand that as
Stage 1, 2, and 3 structural susceptibility evaluations are completed, NextEra staff intend
to update SMPM, Appendix C, Building Deformation Monitoring Tables, with critical
structural monitoring points (qualitative and/or quantitative) that are deemed appropriate
to effectively monitor ASR impacts and progression for each affected structure. The
inspectors also reviewed the current Change Management Plan for the SMP (AR No.
02148021, dated October 11, 2017), which identified numerous pending changes that
were being tracked for the next revision to the SMPM. Revision 03, dated November 17,
2017, was approved after the end of the inspection. The inspectors verified that the
27
monitoring points for the recently completed structural evaluations were added to
Appendix C.
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 1 sample)
(Closed) Licensee Event Report (LER) 05000443/2017-001-00: Manual Reactor Trip in
Response to a Feedwater Isolation due to High Level in Steam Generator B
a. Inspection Scope
The inspectors reviewed the LER, root cause analysis, and event analysis, following the
April 29, 2017, plant trip, due to steam generator water level perturbations. Additionally,
the inspectors reviewed follow-up actions related to the event to assure that NextEra
staff implemented appropriate corrective actions commensurate with their safety
significance. The enforcement actions associated with this LER are discussed below.
This LER is closed.
b. Findings
Introduction. A self-revealing Green finding was identified for inadequate
implementation of procedure MA 4.5, Configuration Control, Revision 18. Specifically,
maintenance technicians failed to properly implement MA 4.5 while backfilling steam
generator instrumentation, and inadvertently left an instrumentation valve partially open
instead of fully open. This resulted in slow response of the instrument, and ultimately a
high steam generator level, a feedwater isolation signal and a manual reactor trip.
Description. On April 29, control room operators manually tripped the reactor when the
B steam generator level reached the feedwater isolation signal setpoint. The plant was
at approximately 12 percent power, and operators were raising power in preparation for
main generator synchronization. At the time, feedwater was being manually controlled
by the operators, and the wide range steam generator level indication was being used to
determine the required feedwater flow. The wide range level indication was responding
slowly to level changes which resulted in overfeeding the steam generator. This caused
the steam generator level to increase to the feedwater isolation signal setpoint.
NextEra personnel determined that the slow response of the steam generator level
indication was due to an instrumentation valve left partially open instead of fully open as
required. On April 26, instrumentation and control technicians had performed a
backfilling of the steam generator reference legs. The technicians used procedure
MA 4.5, including Form MA 4.5A, Configuration Change, to track the valve
manipulations to maintain configuration control. MA 4.5 requires that all component
manipulations and changes to component and plant configuration are performed only to
a detailed procedure or written instruction, and shall be documented on form MA 4.5A or
in an operating procedure WO, or job plan. The technicians did not properly use
place-keeping and concurrent verification during the performance of the backfilling
activity, and one instrumentation valve was left in a nearly full closed position instead of
the full open position. NextEra promptly rechecked other similar valves, then performed
a root cause evaluation that eventually led to additional technician training and improved
configuration controls during such evolutions.
28
Analysis. The inspectors determined that NextEras failure to properly implement
MA 4.5 was a performance deficiency within NextEras ability to foresee and correct, and
should have been prevented. Specifically, instrumentation and control technicians failed
to open an instrumentation valve at the end of a steam generator level indicating system
backfill maintenance activity. This resulted in operators unable to properly control steam
generator water level during startup operations, and ultimately led to a required plant trip
due to high steam generator level and a feedwater isolation signal.
This finding is more than minor because it is associated with the configuration control
attribute of the Initiating Events cornerstone and affected the cornerstone objective to
limit the likelihood of events that upset plant stability and challenge critical safety
functions during shutdown as well as power operations. Specifically, the failure to
effectively implement MA 4.5 resulted in a valve being left out of its required position, a
subsequent lack of steam generator water level control during low power operations, and
ultimately required a manual reactor trip. Additionally, the finding is similar to
Example 4.b of IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples
of Minor Issues, issued August 11, 2009, in that the performance deficiency caused a
reactor trip. In accordance with IMC 0609.04, Initial Characterization of Findings,
issued June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The Significance
Determination Process for Findings At-Power, issued June 19, 2012, the inspectors
determined that this finding is of very low safety significance (Green), because the
finding did not cause a reactor trip and the loss of mitigation equipment relied upon to
transition the plant from the onset of a trip to a stable shutdown condition.
In accordance with IMC 0310, the finding has a cross-cutting aspect in the area of
Human Performance, Work Management, because the organization did not implement a
process of planning, controlling, and executing the work activity such that nuclear safety
was the overriding priority. Specifically, NextEra did not ensure that a steam generator
backfilling activity was properly executed, which resulted in the slow response of a
steam generator level indication, the overfeeding of the steam generator, a feedwater
isolation signal, and the ultimate requirement to trip the reactor [H-5].
Enforcement. This finding does not involve enforcement action because no violation of a
regulatory requirement was identified. Because this finding does not involve a violation
and is of very low safety significance, it is identified as a finding.
(FIN 05000443 /2017004-03, Inadequate Procedure Implementation Results in a
Manual Reactor Trip)
4OA6 Meetings, Including Exit
On January 23, 2018, the inspectors presented the inspection results to Mr. Eric
McCartney, Regional Vice President, Northern Region, 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.
ATTACHMENT: SUPPLEMENTARY INFORMATION
A-1
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
E. McCartney, Regional Vice President, Northern Region
C. Domingos, Site Director
K. Boehl, Senior Radiation Protection Analyst
K. Browne, Licensing Manager
E. Carley, License Renewal Supervisor
A. Giotos, Senior Analyst
J. Hulbert, Nuclear Engineer
D. Robinson, Chemistry Manager
D. Strand, Radiation Protection Manager
T. Smith, Radiation Protection Supervisor
C. Thomas, Licensing Engineer
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened/Closed
05000443/2017004-01 NCV Licensed Operator Examination Integrity Not
Ensured (Section 1R11.3)05000443/2017004-02 NCV Failure of Exercise Critique to Identify a RSPS
Weakness (Section 1EP6)05000443/2017004-03 FIN Inadequate Procedure Implementation Results in
a Manual Reactor Trip (Section 4OA3)
Closed
05000443/2017-001-00 LER Manual Reactor Trip in Response to a
Feedwater Isolation due to High Level in Steam
Generator B (Section 4OA3)
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
OP-AA-102-1002, Seasonal Readiness, Revision 20
Condition Reports
2225659 2227085 2227175
Maintenance Orders/Work Orders
40500528
Attachment
A-2
Miscellaneous
Seabrook Station certification of seasonal readiness, Winter 2017-2018, dated 9/22/17
Section 1R04: Equipment Alignment
Procedures
OS0443.36, Fire Pump House Weekly Valve Alignment, Revision 6
OS1016.03, A Service Water Operation, Revision 17
OS1016.04, B Service Water Operation, Revision 20
OS1016.05, Service Water Cooling Tower Operation, Revision 34
OX1416.01, Service Water Monthly Valve Verification, Revision 12
OX1416.05, Service Water Quarterly Operability Test Cooling Tower Pump, Revision 27
OX1416.03, Cooling Tower Fan Monthly Operability Test, Revision 10
OX1456.02, ECCS Monthly System Verification, Revision 20
Miscellaneous
UFSAR 9.2.1, Revision 18
Drawings
1-CS-B20725, Chemical & Volume Control Charging System Detail, Revision 32
1-CS-B20729, Chemical & Volume Control System Boric Acid Detail, Revision 20
1-FP-B20266, Fire Protection Fire Pump House Detail, Revision 25
1-SI-B20446, Safety Injection System Intermediate head Injection System Detail, Revision 18
1-SW-B20792, Service Water System Nuclear Overview, Revision 6
1-SW-B20794, Service Water System Nuclear Detail, Revision 39
1-SW-B20795, Service Water System Nuclear Detail, Revision 44
Section 1R05: Fire Protection
Miscellaneous
Seabrook Station Fire Protection Pre-Fire Strategies, Volume 1
Section 1R06: Flood Protection Measures
Procedures
OS1212.01, PCCW System Malfunction, Revision 13
OS1213.01, Loss of RHR During Shutdown Cooling Revision 19
OP-AA-109, Control of Time Critical Operator Actions and Time Sensitive Actions, Revision 2
Miscellaneous
Report TP-7, Seabrook Station Moderate Line Break Study, Revision 5
UFSAR Section 3.6B, Revision 8; Section 9.2, Revision 14
Section 1R07: Heat Sink Performance
Miscellaneous
A RHR Heat Exchanger Performance Monitoring Data from OR18
B RHR Heat Exchanger Performance Monitoring Data from OR18
EPRI NP-7552, Heat Exchanger Performance Monitoring Guidelines
A-3
Drawings
1-CC-B20204, Primary Component Cooling Loop A Overview, Revision 4
1-CC-B20205, Primary Component Cooling Loop A Detail, Revision 26
1-RH-B20660, Residual Heat Removal System Overview, Revision 3
1-RH-B20663, Residual Heat Removal System Train B Cross-tie Detail, Revision 21
1-RH-B20660, Residual Heat Removal System Overview, Revision 3
9763-F-805203, PAB Vaults Piping Zone 30D Plan at EL(-) 9-0, Revision 12
Section 1R11: Licensed Operator Requalification Program
Procedures
OP 9.2, Transient Response Procedure Users Guide, Revision 18
OP-AA-100-1001, License Maintenance and Activation, Revision 4
TR-AA-220-1004, Licensed Operator Continuing Training Annual Operating and Biennial
Written Exams, Revision 2
TR-AA-230-1007, Conduct of Simulator Training and Evaluation, Revision 5
Condition Reports
2114495 2117035 2202358
Miscellaneous
Seabrook 2016-2017 Requalification Training Program Annual Examination Sample Plan
Simulator-Related Test Documents
NT-3730-1, SBT Package for L15R11, Rev. 11, dated 9/23/16
NT-3730-1, Seabrook Transient No. 1, Manual Reactor Trip, Rev. 17, dated 3/25/17
NT-3730-1, Seabrook Transient No. 11, Large Break LOCA with Loss of Offsite Power, Rev. 17
dated 3/16/17
NT-3730-1, Seabrook Transient No. 2, Simultaneous Trip of Both Main Feedwater Pumps,
Rev. 17, dated 3/25/17
NT-3730-1, Seabrook Transient No. 3, Simultaneous Closure of All Main Steam Isolation Valves,
Rev. 17, dated 3/25/17
NT-3730-1, Seabrook Transient No. 6a, Main Turbine Trip Below the P-9 Permissive, Rev. 17,
dated 3/25/17
NT-3730-1, Seabrook Transient No. 8, Slow Primary Depressurization, Rev. 17, dated 3/15/17
NT-3730-1, Steady State Value Comparison Test - 100% Power, Rev. 17, dated 5/16/17
NT-3730-1, Steady State Value Comparison Test - 46% Power, Rev. 17, dated 9/13/16
NT-3730-1, Steady State Value Comparison Test - 79% Power, Rev. 17, dated 5/15/17
NT-3730-1, Steady State Value Comparison Test - Post Event Test A Water Box Isolated,
Rev. 17 dated 1/20/16
Section 1R12: Maintenance Effectiveness
Condition Reports
021307 222005 592531 1682547 2234042 2234311
Maintenance Orders/Work Orders
4052273 40125669 40568790
Miscellaneous
A-4
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
IXI680.032, Solid State Protection System (SSPS) Train B MODE 1 Actuation Logic Test,
Revision 08
OP-AA-105-1000, Operational Decision Making, Revision 10
OP-AA-103-1000, Reactivity Management, Revision 6
WM-AA-100, Risk Management Program, Revision 2
WM-AA-100-1000, Work Activity Risk Management, Revision 10
Condition Reports
0200122 0513191 0515294 0601265 2230707 2234042
2234311
Maintenance Orders/Work Orders
4054097 40437454 40490516 40513114 40513114 40516271
40516273 40568790 94167526
Miscellaneous
Just-in-Time Training, IX1680.932 SSPS B Actuation Logic Test Handout
Drawings
1-NHY-310949, Solid State Protection System Schematic Diagram
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
EN-AA-203-1001, Operability Determinations / Functionality Assessments, Revision 27
Condition Reports
2230707 2236247
Maintenance Orders/Work Orders
40565937
Section 1R19: Post-Maintenance Testing
Procedures
IS1672.315, SW-P-8282 Service Water Pump B/D Discharge Header Pressure Calibration,
Revision 6
IX1605.013, IST Solenoid Valve Time Response Testing, Revision 4
LS0563.23, Type IAC Overcurrent Relay Inspection, Testing and PM, Revision 13
LS0569.09, Diagnostic Testing of Butterfly MOVs, Revision 27
MA-AA-100-1011, Equipment Troubleshooting, Revision 3
MA3.5, Post Maintenance Testing, Revision 23
OX0443.01, Diesel Fire Pump Weekly Test, Revision 16
OX1456.81, Operability Testing of IST Valves, Revision 29
OX1456.86, Operability Testing of IST Pumps, Revision 15
OX1490.05, Miscellaneous Systems ASME Quarterly Testing, Revision 9
A-5
Condition Reports
0289856 2227780 2230622 2234042 2238019 2238020
2238038 2238053 2240790
Maintenance Orders/Work Orders
40189098 40496829 40497318 40516877 40531737 40563635
40568790 94170738
Miscellaneous
ECs 288964, 286645
Calculation 9763-3-ED-00-23-F, Medium Voltage Protective Relay Coordination, Revision 5
Drawings
1-NHY-250000, Revision 83
1-NHY-506839, Service Water Pumps P-41B & P41D Control Loop Diagram, Revision 9
Section 1R22: Surveillance Testing
Procedures
IX1656.938, NI-N-44 Power Range NI Rescaling Calibration, Revision 12
OPMM, Operations, Management Manual, Revision 107
OS1412.13, PCCW Train A Quarterly Operability, 18 Month Position Indication, and
Comprehensive Pump Testing, Revision 0
OX1456.86, Operability Testing of IST Pumps, Revision 15
Condition Reports
2227744
Maintenance Orders/Work Orders
40508512 40515051 40561271
Drawings
PID-1-CC-B20205, Revision 27
Section 1EP6: Drill Evaluation
Procedures
ER 1.1, Classification of Emergencies, Revision 58
ER 3.1, Technical Support Center Operations, Revision 64
EP-AA-100-1000, Conduct of Emergency Preparedness, Revision 6
EP-AA-101-1000, Nuclear Division Drill and Exercise Procedure, Revision 20
Condition Reports
2223189 2229621 2232420
Miscellaneous
CFD 17-03 Drill Scenario
Combined Functional Drill Report, CFD-17-03, dated October 11, 2017
NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6
Training Lesson Plan, E01090I, Emergency Classifications, Revision 6
Training Lesson Plan, L1809C, Nuclear Energy Institute (NEI), Emergency Action Levels,
Revision 6
A-6
Section 2RS2: Occupational ALARA Planning and Controls
Procedures
RP-AA-104-1000, ALARA Implementing Procedure, Revision 13
Condition Reports
02173460 02198478 02198480 02198735 02199920 02215940
02220919 02221333 02223171
Miscellaneous
2017 Department Exposure Goals and Year to Date Department Exposures, December 5, 2017
2017 Routine Operating Dose Report, December 3, 2017
ALARA Dose Estimate Report for Work Week 1749 (December 3-8, 2017), December 4, 2017
ALARA Review Board Meeting 17-04 Subcommittee, September 19, 2017
ALARA Review Board Meeting 17-03 Subcommittee, August 14, 2017
ALARA Review Board Meeting 17-02 Subcommittee, July 26, 2017
Level 1 Assessment for NRC ALARA RETS, REMP Inspections, AR 2233688, October 30, 2017
Post-Job ALARA Review No. 17-0031, Dry Fuel Storage Project Activities, December 6, 2017
Post-Job ALARA Review No. 17-0140, OR 18 Scaffolding, December 6, 2017
Post-Job ALARA Review No.17-002, Steam Generator Primary Eddy Current Test,
December 6, 2017
Post-Job ALARA Review No.17-001, Reactor Vessel Disassembly and Reassembly,
December 6, 2017
Section 2RS3: In-Plant Airborne Radioactivity Controls and Mitigation
Procedures
HD0965.01, Respiratory Protection Quality Assurance and Maintenance Program, Revision 22
HD0965.02, Repair, Inspection, Inventory and Maintenance of Respiratory Protection
Equipment, Revision 27
HD0965.08, Breathing Air Certification, Revision 17
HD0965.10, Respirator Fit Testing Using the TSI Portacount, Revision 19
HD0965.12, Respiratory Equipment Issue and Use, Revision 42
RP-AA-106, Respiratory Protection Program, Revision 0
Condition Reports
02122162 02149186 02168471 02178320
Miscellaneous
Annual Assessment of the 2016 Respiratory Protection Program, AR 2206817, June 7, 2017
FireHawk M7 SCBA Use: Inspection and Donning Instructions, Operator Aide, Revision 9
Fit Test Report for MSA Ultra Elite 1000 (medium) using Portacount # 8030142409,
December 7, 2017
Fit Test Report for MSA Ultra Rubber (medium) using Portacount # 8030142409,
December 7, 2017
HD0965.02, Figure 2: SCBA Inventory, November 30, 2017
HD0965.02, HRE-M1 SCBA Inspection and Inventory, November 30, 2017
HD0965.14, Form B: SCBA Face Piece Test for Ultra Elite 1000 (medium) FH-022,
December 4, 2017
HD0965.14, Form B: SCBA Face Piece Test for Ultra Elite 1000 (medium) FH-037,
September 7, 2017
HD0965.14, Form A: SCBA/ PremAire Cadet Regulator Flow Test ANAD063768,
December 4, 2017
A-7
HD0965.14, Form A: SCBA/ PremAire Cadet Regulator Flow Test APAB279701,
August 11, 2017
Honeywell Certificate of Calibration No. 56041717L02497 Serial No. L02497, April 1, 2017
MSA CARE Authorized Repair Center and MSA MMR Certified CARE Technician Certification,
March 3, 2015
Posi3 USB Test Results Serial No. L02497 for MSA Ultra Elite (medium) FH-022,
December 4, 2017
Posi3 USB Test Results Serial No. L02497 for MSA FireHawk M7 Air Mask (medium) PR 14,
December 4, 2017
SBK HPT HP0090J, RP Technician Respirator Training, June 2, 2014
SBK GET GT1074J, Firehawk M7 SCBA Training, July 11, 2013
Service History for Instrument Model SCBA Regulator (including maintenance/repair notes),
December 6, 2017
TRI Air Testing, Inc Laboratory Report Compressed Air/Gas Quality Test for Firefighting Annex
Breathing Air, September 15, 2017
TRI Air Testing, Inc Laboratory Report Compressed Air/Gas Quality Test for Firefighting Annex
Breathing Air, June 15, 2017
TRI Air Testing, Inc Laboratory Report Compressed Air/Gas Quality Test for Firefighting Annex
Breathing Air, June 23, 2017
TSI Certificate of Testing PortaCount 8030 Bench 2 Serial No. 8030134708,
September 20, 2017
TSI Certificate of Testing PortaCount 8030 Bench 2 Serial No. 8030134713, July 12, 2017
Section 4OA1: Performance Indicator Verification
Procedures
CS0917.02, Gaseous Effluent Releases, Revision 14
CX0917.01, Liquid Effluent Releases, Revision 20
HD0958.33, Performance of Radiation Protection Supervisory Plant Walkdowns, Revision 6
JD0999.910, Reporting Key Performance Indicators per NEI 99-02, Revision 8
Condition Reports
02093824 02162340 02195218
Miscellaneous
CP 4.1C, Release Index Log 2016, November 6, 2017
CX0917.01, Form C: LEW Release Data, Permit # 17-448, Waste Test Tank B,
October 29, 2017
CX0917.01, Form C: LEW Release Data, Permit # 17-458, Storm Drain/Groundwater Extraction
Wells, October 31, 2017
CX0917.01, Form C: LEW Release Data, Permit # 17-462, Steam Generator Blowdown Drain
Flash Tank, November 8, 2017
CX0917.01, GEW Sample Collection Data, Permit # 17-451, Plant Vent, October 31, 2017
JD0999.910, Figure 1, Occupational Exposure Occurrence, January, February and March 2017,
dated April 25, 2017
JD0999.910, Figure 1, Occupational Exposure Occurrence, April, May and June 2017,
dated July 7, 2017
JD0999.910 Figure 1 Occupational Exposure Occurrence, July, August and September 2017,
dated October 27, 2017
LIC-17010, Seabrook Station NRC Third Quarter 2017 Performance Indicator Submittal
NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7
NextEra - Seabrook Station 2016 Annual Radioactive Release Report, April 28, 2017
A-8
MSPI Derivation Reports for MSPI Systems Residual Heat Removal System and Cooling Water
System, November 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
November 2017, December 1, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
October 2017, November 1, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
September 2017, October 2, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
August 2017, September 5, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
July 2017, August 1, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
June 2017, July 5, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
May 2017, June 2, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
April 2017, May 2, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
March 2017, April 2, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
February 2017, March 1, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
January 2017, February 1, 2017
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
December 2016, January 3, 2016
Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,
November 2016, December 1, 2017
SBK-PRAE-15-001
Section 4OA2: Problem Identification and Resolution
Procedures
ER-AA-101, Equipment Reliability, Revision 7
ER-AA-201-2001, System Health Reporting, Revision 12
ER-AA-201-2002, System Performance Monitoring, Revision 4
OP-AA-108-1000, Operator Challenges Program Management, Revision 5
PI-AA-207-1000, Station Self-Evaluation and Trending Analysis, Revision 8
PI-AA-207, Trend Coding and Analysis, Revision 12
PI-AA-101, Assessment and Improvement Programs, Revision 23
SMPM, Structures Monitoring Program Manual, Revisions 2 and 3
Condition Reports
1637922 2053980 2144822 2151482 2153374 2157499
2162430 2162696 2162696 2164268 2164482 2168700
2175840 2178962 2178962 2181193 2205604 2207649
2214502 2215560 2215959 2216230 2216936 2217146
2217211 2219903 2222763 2222809 2223576 2224985
2227328 2232578 2235442 2236473 2237328 2237940
2238405 2111108 2148021 2240426
A-9
Maintenance Orders/Work Orders
01209317 01209321 40176613 40260904 40395367 40531735
40538714 40540846 40568543
Miscellaneous
160268-CA-05, Susceptibility Evaluation of Containment Enclosure Ventilation Area, Revision 0,
dated March 22, 2017
170400-SVR-04-RA, 2017 Tier 2 Inspections - ASR Inspections and Cracking Index
Measurements on Concrete Structures, dated October 10, 2017
170400-SVR-05-RA, 2017 Tier 2 Inspections - Measurements for ASR Expansion on Concrete
Surfaces, dated October 10, 2017
Evaluation - North Wall of Containment Enclosure Ventilation Area (CEVA) Near-Surface
Delamination (Cover Concrete Separation), dated October 30, 2017
FP 101034, Petrographic Examinations of Equipment Vaults, Revision 1
FP 101044, Identify and Measure Seismic Gaps Between the CEB and CB at 4 Missile Shields,
Revision 0
FP 101055, Condition Assessment of Cracking in RHR and CS Equipment Vault - Second Visit,
Revision 0
PEG-94, Revision 11
Prompt Operability Determination (POD) for AR 01664399, Consolidation of PODs for Reduced
Concrete Properties in Alkali Silica Reaction (ASR) Affected Seismic Category I
Structures, Revision 2, dated October 6, 2017
POD AR 02014325, Consolidation of Building Deformation Prompt Operability Determinations,
Revision 1, dated October 6, 2017
POD for AR 02193235, Alkali Silica Reaction (ASR) effects on CEVA Structure North Wall,
Revision 1, dated September 28, 2017
POD for AR 02215578, Evaluation of B Electrical Cable Tunnel as an Alkali Silica Reaction
(ASR) Affected Seismic Category I Structure, Revision 2, dated July 19, 2017
Drawings
9763-F-101620, Sheet 1, Containment Enclosure Ventilation Area Concrete Sections,
Revision 5
9763-F-113230, Sheet 1, Schedule of Required Backfill Concrete and Isolation Material for
Structures, Revision 5
Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion
Procedures
MA 4.5, Configuration Control, Revision 18
MA-AA-100, Conduct of Maintenance, Revision 16
MA-AA-203-1001, Work Order Planning, Revision 8
OP-AA-100-1000, Conduct of Operations, Revision 20
Condition Reports
2202358
Maintenance Orders/Work Orders
40532423
A-10
Miscellaneous
LER 2017-001-00, Manual Reactor Trip in Response to a Feedwater Isolation due to High Level
in Steam Generator B, June 27, 2017
Manual Reactor Trip in Response to a Feedwater Isolation due to High Level in Steam
Generator B, Event Date: 4/29/17, Root Cause Evaluation
P-14 Event Analysis
LIST OF ACRONYMS
ACI American Concrete Institute
ADAMS Agencywide Documents Access and Management System
ALARA As Low As is Reasonably Achievable
ASR alkali silica reaction
CAP corrective action program
CEVA containment enclosure ventilation area
CFR Code of Federal Regulations
CR condition report
CWT cooling water tower
FIN finding
EAL emergency action level
ECCS emergency core cooling system
EDG emergency diesel generator
EOP emergency operations facility
EPA Environmental Protection Agency
GE general emergency
HVAC heating, ventilation, and air conditioning
IMC Inspection Manual Chapter
LER Licensee Event Report
LOCA loss of coolant accident
MR Maintenance Rule
NCV non-cited violation
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
PAB primary auxiliary building
PAG protective action guide
PAR protective action recommendation
PCCW primary component cooling water
PI performance indicator
RG Regulatory Guide
RO reactor operator
RSPS risk significant planning standard
SAE site area emergency
SCBA self-contained breathing apparatus
SED Site Emergency Director
SMP Structures Monitoring Program
SMPM Structures Monitoring Program Manual
SRO senior reactor operator
SSC structure, system, and component
A-11
STED Short-Term Emergency Director
TS technical specification
UFSAR September 22, 2015, because the finding Updated Final Safety
Analysis Report
WO work order
WRGM wide range gas monitor