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 05000443/201
7 004 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. Eri c 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 finding s of very low safety significance (Green) in this report
. Two of these finding s involved a violation of NRC requirements. The NRC is treating
these violation s as non-cited violation
s (NCV s) 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
-rm/adams.html
and the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations
(C FR) 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
0500 0 443/2017004 w/Attachment: Supplementary Information
cc w/encl:
Distribution via ListServ
SUNSI Review
Non-Sensitive Sensitive Publicly Available
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 Enclosure U.S. NUCLEAR REGULATORY COMMISSION
REGION I Docket No:
50-443 License No:
NPF-86 Report No.:
05000 443/2017004 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. D aun, 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
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 05000 443/20 17004; 10/01/2017 to 12/31/2017; Seabrook; Licensed Operator Requalification Program, Emergency Preparedness Drill Observation
, and Follow-Up of Events and Notices of Enforcement Discretion
. 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 violation s (NCV s) 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 NRC's Enforcement Policy, dated August 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG
-1649, "Reactor Oversight Process," Revision 6.
Cornerstone:
Green. A se lf-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 NextEra's
Corrective Action Program (CAP) as AR 2239906.
The failure of NextEra's 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 question
s 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 licensee's 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
5 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
andthe associated weakness is in the licensee's 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 site's 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
- 1 samples) Readiness for Seasonal Extreme Weather Conditions
a. Inspection Scope
The inspectors reviewed NextEra's readiness for the onset of seasonal cold temperatures. The review focused on the service water
(SW) pump house, the cooling water tower (C W T) 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 NextEra's 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
- 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 system's 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
- 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
- 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 verif y 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 respondin g 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
- 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
-o f-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
- 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
- 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
I MC 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 inspector
s 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 inspector
s 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
fou r 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 RO s and 13 SRO
s) 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 licensee's CAP, and the most recent NRC plant issues matrix. The inspectors also reviewed specific events from the licensee's 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 NextEra's failure to ensure the integrity of the biennial
written examination
s 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 version
s 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 week's examination with any other week's examination questions. In other words
, the licensee was applying the question overlap criteria from
12 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 use d 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 NextEra's 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 year's comprehensive written examination was developed by a different individual who, along with other fleet personnel, misapplied the fleet procedure's overlap criteria.
The licensee entered this issue into their CAP as AR 2239906. Analysis. The failure of NextEra's 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 question
s 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 NextEra's CAP as AR 2239906, this violation is being treated as an NCV consistent with Section 2.3.2.a of the NRC Enforcement Policy.
-01 Licensed Operator Examination Integrity Not Ensured) 1R12 Maintenance Effectiveness
- 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
- 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 station's probabilistic risk analyst to verify plant conditions were consistent with the risk assessment. The inspectors also reviewed the
TS requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met.
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 M
ode 1 actuation logic test on November 27
b. Findings No findings were identified.
1R15 Operability Determinations and Functionality Assessments
- 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 NextEra's evaluations to determine whether the components or systems were operable. The inspectors confirmed, where appropriate, compliance with bounding limitations associated with the evaluations. Where compensatory measures were required to maintain operability , the inspectors determined whether the measures in place would function as intended and were properly controlled by
NextEra. b. Findings No findings were identified.
1R19 Post-Maintenance Testing
- 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
- 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.
16 Cornerstone:
1EP6 Drill Evaluation
(71114.06)
.1 Emergency Preparedness Drill Observation
(2 samples)
a. Inspection Scope
The inspectors evaluated the conduct of routine NextEra emergency drill
s 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 NextEra's 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 a n RCS leak of approximately 300 gallons per minute and resulted in the actuation of safety injection. The SE D, 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 SED's 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 Agency's (EPA's) protective action guide (PAG) levels
that are incorporated in the emergency plan
. Seabrook's emergency plan directs the comparison of radiological release data with EPA's 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, Seabrook's 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 location
s 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, Seabrook's 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 escalat e 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 site's 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
A ppendix 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 nee
d 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 NextEra's declaration of a GE that was based on insufficient information. Because this violation was of very low safety significance and was entered into Seabrook's CAP
as AR 2242073, this finding is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy.
-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 crew's performance and ensure that NextEra evaluators noted the same issues and entered them into the CAP. b. Findings No findings were identified.
20 2. RADIATION SAFETY
Cornerstone:
2RS2 Occupational As Low As Is Reasonably Achievable Planning and Controls
(71124.02 - 1 sample) a. Inspection Scope
The inspectors assessed NextEra's 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
C FR Part 20, applicable Regulatory Guides
(RGs) 8.8 and 8.1
0 , 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 evaluate d 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
- 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 inspector
s reviewed the
following: the
status and surveillance records for three Self-Contained Breathing Apparatus
(SCBAs) staged in-plant for use during emergencies; Next Era's 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.
21 4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification
(71151) .1 Mitigating Systems Performance Index (3 samples) a. Inspection Scope
The inspectors reviewed NextEra's 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 NextEra's 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 inspector
s reviewed licensee submittals for the occupational radiological occurrences
PI for the fourth quarter 2016
through the first, second, and third quarters 2017. The inspector
s used PI definitions and guidance contained in the N
EI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the PI data reported. The inspector
s reviewed electroni
c 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 inspector
s 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 inspector
s
22 used PI definitions and guidance contained in the N
EI Document 99
-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine if the PI data was reported properly.
The inspector
s reviewed the public dose assessments for the PI for public radiation safety to determine if related data was accurately calculated and reported. The inspector
s 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 inspector
s 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 NextEra's 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 NRC
's 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 Observation
s 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 is
sues following a program assessment documented under CR 221
9903. .3 Annual Sample:
a. Inspection Scope
The inspectors performed an in
-depth review of NextEra's 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 NextEra's problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of NextEra's 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 NextEra's 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 Reactio
n a. Inspection Scope
The purpose of periodic site visits to Seabrook Station over the past few
years has been to review the adequacy of NextEra's monitoring of alkali
-silica reaction (ASR) on affected reinforced concrete structures, per their 10 CFR 50.65 "Maintenance Rule" Structures Monitoring Program (SMP), and NextEra's 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 document
s. 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 NextEra's actions to verify compliance with the SMP, the CAP , and 10 CFR
Part 50, Appendix B requirements.
b. Findings and Observation
s 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 NextEra's 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.
Inspector s' 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 NextEra's 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 discussion
s 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 plan ned 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 NextEra's 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 review ed 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)
05000 443/201 7-001-00: 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 full y 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
ful ly 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 NextEra's failure
to properly implement
MA 4.5 was a performance deficiency within NextEra's 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 t
he 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 05000 443 /2017004-0 3 , Inadequate Procedure Implementation Results in a Manual Reactor Trip)
4OA6 Meetings, Including Exit
On January 23, 201 8, 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 Attachment
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
E. McCartney , Regional Vice President, Northern Region
C. Domingos , Site Director
K. Boehl, Senior Radiation Protection Analy
st 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
-0 1 NCV Licensed Operator Examination Integrity
Not Ensured (Section 1R11.3) 05000443/2017004
-0 2 NCV Failure of Exercise Critique to Identify a RSPS
Weakness (Section 1EP6)
05000 443/20 17004-0 3 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
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 User's 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
EC 290088 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, Revisi
on 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
EC s 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, Revisio
n 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 201
6 Annual Radioactive Release Report, April
28, 201 7
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, Revision
s 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 Seismi
c 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
AS R 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 RCS reactor coolant system
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