ML18043A821

Integrated Inspection Report 05000443/2017004

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

05000443/2017004

Dear Mr. Nazar:

On December 31, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an

inspection at Seabrook Station, Unit No. 1 (Seabrook). On January 23, 2018, the NRC

inspectors discussed the results of this inspection with Mr. Eric McCartney, Regional Vice

President, and other members of his staff. The results of this inspection are documented in the

enclosed report.

NRC inspectors documented three findings of very low safety significance (Green) in this report.

Two of these findings involved a violation of NRC requirements. The NRC is treating these

violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement

Policy.

If you contest the violations or significance of these NCVs, you should provide a response within

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

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

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

NRC Resident Inspector at Seabrook. In addition, if you disagree with a cross-cutting aspect

assignment or a finding not associated with a regulatory requirement in this report, you should

provide a response within 30 days of the date of this inspection report, with the basis for your

disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,

Washington, DC, 20555-0001; with copies to the Regional Administrator, Region I, and the NRC

Resident Inspector at Seabrook.

M. Nazar 2

This letter, its enclosure, and your response (if any) will be made available for public inspection

and copying at http://www.nrc.gov/reading-rm/adams.html and the NRC Public Document Room

in accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390, Public

Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Fred Bower, Chief

Reactor Projects Branch 3

Division of Reactor Projects

Docket No. 50-443

License No. NPF-86

Enclosure:

Inspection Report 05000443/2017004

w/Attachment: Supplementary Information

cc w/encl: Distribution via ListServ

ML18043A821

SUNSI Review Non-Sensitive Publicly Available

Sensitive Non-Publicly Available

OFFICE RI/DRP RI/DRP RI/DRP

NAME R. Barkley P. Cataldo/RB F. Bower

DATE 2/7/2018 2/12/2018 2/12/2018

1

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No: 50-443

License No: NPF-86

Report No.: 05000443/2017004

Licensee: NextEra Energy Seabrook, LLC (NextEra)

Facility: Seabrook Station, Unit No. 1 (Seabrook)

Location: Seabrook, NH 03874

Dates: October 1, 2017 through December 31, 2017

Inspectors: P. Cataldo, Senior Resident Inspector

T. Daun, Acting Senior Resident Inspector

P. Meier, Resident Inspector

N. Perry, Senior Resident Inspector

B. Dionne, Health Physicist

B. Cook, Senior Reactor Analyst

N. Floyd, Reactor Inspector

A. Buford, Structural Engineer, NRR

D. Silk, Senior Operations Engineer

Approved By: Fred Bower, Chief

Reactor Projects Branch 3

Division of Reactor Projects

Enclosure

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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

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SUMMARY

IR 05000443/2017004; 10/01/2017 to 12/31/2017; Seabrook; Licensed Operator Requalification

Program, Emergency Preparedness Drill Observation, and Follow-Up of Events and Notices of

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

violations (NCVs) and one finding, all of which were of very low safety significance (Green).

The significance of most findings is indicated by their color (i.e., greater than Green, or Green,

White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, Significance

Determination Process, dated October 28, 2016. Cross-cutting aspects are determined using

IMC 0310, Aspects Within the Cross-Cutting Areas, dated December 4, 2014. All violations of

NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated

August 1, 2016. The NRCs program for overseeing the safe operation of commercial nuclear

power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6.

Cornerstone: Initiating Events

• Green. A self-revealing Green finding was identified for inadequate implementation of

procedure MA 4.5, Configuration Control, Revision 18. Specifically, maintenance

technicians failed to properly implement MA 4.5 while backfilling steam generator

instrumentation, and inadvertently left an instrumentation valve partially open instead of fully

open. This resulted in slow response of the instrument, and ultimately a high steam

generator level, a feedwater isolation signal and a manual reactor trip. NextEra promptly

rechecked other similar valves, then performed a root cause evaluation that eventually led

to additional technician training and improved configuration controls during such evolutions.

This finding is more than minor because it is associated with the configuration control

attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit

the likelihood of events that upset plant stability and challenge critical safety functions

during shutdown as well as power operations. Specifically, the failure to effectively

implement MA 4.5 resulted in a valve being left out of its required position, a subsequent

lack of steam generator water level control during low power operations, and ultimately

required a manual reactor trip. In accordance with IMC 0609.04, Initial Characterization of

Findings, issued June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The Significance

Determination Process for Findings At-Power, issued June 19, 2012, the inspectors

determined that this finding is of very low safety significance (Green), because the finding

did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the

plant from the onset of a trip to a stable shutdown condition. Additionally, the finding has a

cross-cutting aspect in the area of Human Performance, Work Management, because the

organization did not implement a process of planning, controlling, and executing the work

activity such that nuclear safety was the overriding priority. Specifically, NextEra did not

ensure that a steam generator backfilling activity was properly executed, which resulted in

the slow response of a steam generator level indication, the overfeeding of the steam

generator, a feedwater isolation signal, and the ultimate requirement to trip the

reactor. [H.5] (Section 4OA3)

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Cornerstone: Mitigating System

• Green. The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code

of Federal Regulations (10 CFR) 55.49, Integrity of Examinations and Tests, for the failure

of the licensee to ensure that the integrity of the written examinations administered to

licensed operators was maintained. During the planning of the biennial written

examinations, two written examinations would have exceeded the 50 percent overlap

criteria limit of questions administered in the previous four weeks of this examination cycle.

This failure resulted in a compromise of examination integrity because it exceeded the

NextEra Fleet Procedure TR-AA-220-1004, Licensed Operator Continuing Training Annual

Operating and Biennial Written Exams, Revision 2, requirement to repeat less than or

equal to 50 percent of the questions used during the exam cycle. However, this

compromise did not lead to an actual effect on the equitable and consistent administration

of the examination because of detection of this issue by the NRC prior to examination

administration. This issue was entered into NextEras Corrective Action Program (CAP) as

AR 2239906.

The failure of NextEras training staff to maintain the integrity of examinations administered

to licensed operations personnel was a performance deficiency. The performance

deficiency was more than minor, and therefore a finding, because if left uncorrected, the

performance deficiency could have become more significant in that allowing licensed

operators to return to the control room without valid demonstration of appropriate

knowledge on the biennial examinations could be a precursor to a more significant event.

Using IMC 0609, Significance Determination Process, and the corresponding Appendix I,

Licensed Operator Requalification Significance Determination Process, the finding was

determined to have very low safety significance (Green) because although the finding

resulted in a compromise of the integrity of written examination, the equitable and

consistent administration of the test was not actually impacted by this compromise. This

finding had a cross-cutting aspect in the area of Human Performance, Resources, in that

leaders ensure procedures are available and adequate to support nuclear safety.

Specifically, NextEra established and implemented a procedure that contained instructions

to licensed operator biennial exam writers that were unclear regarding regulatory guidance

to limit written examination questions overlap. [H.1] (Section 1R11.3)

Cornerstone: Emergency Preparedness

• Green. The inspectors identified a Green non-cited violation (NCV) of Title 10 of the

Code of Federal Reglations (10 CFR) 50.47(b)(14) and 10 CFR Part 50, Appendix E,

Emergency Planning and Preparedness for Production and Utilization Facilities,

Section IV.F.2.g. Specifically, Seabrook did not identify and critique a weakness associated

with a risk significant planning standard (RSPS) during their critique following the

August 30, 2017, emergency preparedness drill. The weakness involved the licensees

declaration of a general emergency (GE) that was based on insufficient information.

NextEra entered the issue into the corrective action program (CAP) as AR2242073.

The inspectors determined that not identifying an exercise weakness related to a GE

classification based on insufficient information during the exercise critique was a

performance deficiency that was reasonably within the ability of Seabrook to foresee and

prevent. The finding is more than minor because it is associated with the Emergency

Response Organization attribute of the Emergency Preparedness Cornerstone and affected

the cornerstone objective to ensure that the licensee is capable of implementing adequate

measures to protect the health and safety of the public in the event of a radiological

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emergency. Specifically, Seabrook personnel did not identify an exercise weakness

associated with a RSPS when the incorrect basis for a GE declaration was used by the Site

Emergency Director (SED). The finding was assessed using IMC 0609, Attachment 4,

Initial Characterization of Findings, issued October 7, 2016. This attachment directs

inspectors to utilize IMC 0609, Appendix B, Emergency Preparedness Significance

Determination Process, issued September 22, 2015, because the finding and

the associated weakness is in the licensees emergency preparedness cornerstone. The

inspectors determined the finding was a critique finding, the drill scope was full scale, the

planning standard was risk-significant, and the performance opportunity was a success

utilizing figure 5.14-1, Significance Determination for Critique Findings, and thus

determined this finding was of very low safety significance (Green). The finding was

determined to have a cross-cutting aspect in the area of Human Performance, Change

Management, in that leaders use a systematic process for evaluating and implementing

change so that nuclear safety remains the overriding priority. Specifically, although recent

changes to the sites emergency classification and action level standard scheme were

effective on July 2017, the new EAL procedure and training regarding the changes lacked

sufficient specificity to ensure the users understood the new scheme with respect to the

status of the containment integrity. [H.3] (Section 1EP6)

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REPORT DETAILS

Summary of Plant Status

Seabrook began the inspection period at full power, and there were no plant status changes of

regulatory significance during the remainder of the inspection period. Documents reviewed for

each section of this inspection report are listed in the Attachment.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 - 1 samples)

Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspectors reviewed NextEras readiness for the onset of seasonal cold

temperatures. The review focused on the service water (SW) pump house, the cooling

water tower (CWT) pump area, and portions of the turbine building that contains risk

important systems. The inspectors reviewed the Updated Final Safety Analysis Report

(UFSAR), technical specifications (TSs), control room logs, and the CAP to determine

what temperatures or other seasonal weather could challenge these systems, and to

ensure NextEra personnel had adequately prepared for these challenges. The

inspectors reviewed station procedures, including NextEras seasonal readiness

procedure and applicable operating procedures. The inspectors performed walkdowns

of the selected systems to ensure station personnel identified issues that could

challenge the operability of the systems during cold weather conditions.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial System Walkdowns (71111.04 - 3 samples)

a. Inspection Scope

The inspectors performed partial walkdowns of the following systems:

• A emergency core cooling system (ECCS) during maintenance on the B charging

pump and safety injection pump on November 6

• Boric acid flow paths during maintenance on the boric acid control station on

November 8-9

• B fire pump during A fire pump maintenance on December 14

The inspectors selected these systems based on their risk-significance relative to the

reactor safety cornerstones at the time they were inspected. The inspectors reviewed

applicable operating procedures, system diagrams, the UFSAR, TSs, work orders

7

(WOs), condition reports (CRs), and the impact of ongoing work activities on redundant

trains of equipment in order to identify conditions that could have impacted the systems

performance of its intended safety functions. The inspectors also performed field

walkdowns of accessible portions of the systems to verify system components and

support equipment were aligned correctly and were operable. The inspectors examined

the material condition of the components and observed operating parameters of

equipment to verify that there were no deficiencies. The inspectors also reviewed

whether NextEra staff had properly identified equipment issues and entered them into

the CAP for resolution with the appropriate significance characterization.

b. Findings

No findings were identified.

.2 Full System Walkdown (71111.04S - 1 sample)

a. Inspection Scope

During the period of November 27 through December 1, the inspectors performed a

complete system walkdown of accessible portions of the SW system to verify the

existing equipment lineup was correct. The inspectors reviewed operating procedures,

system diagrams, TSs, and the UFSAR to verify the system was aligned to perform its

required safety functions. The inspectors also reviewed electrical power availability,

component lubrication and equipment cooling, hanger and support functionality, and

operability of support systems. The inspectors performed field walkdowns of accessible

portions of the systems to verify as-built system configuration matched plant

documentation, and that system components and support equipment remained

operable. The inspectors confirmed that systems and components were aligned

correctly, free from interference from temporary services or isolation boundaries,

environmentally qualified, and protected from external threats. The inspectors also

examined the material condition of the components for degradation and observed

operating parameters of equipment to verify that there were no deficiencies.

Additionally, the inspectors reviewed a sample of related CRs and WOs to ensure

NextEra appropriately evaluated and resolved any deficiencies.

b. Findings

No findings were identified.

1R05 Fire Protection

Resident Inspector Quarterly Walkdowns (71111.05Q - 5 samples)

a. Inspection Scope

The inspectors conducted tours of the areas listed below to assess the material

condition and operational status of fire protection features. The inspectors verified that

NextEra controlled combustible materials and ignition sources in accordance with

administrative procedures. The inspectors verified that fire protection and suppression

equipment was available for use as specified in the area pre-fire plan, and passive fire

barriers were maintained in good material condition. The inspectors also verified that

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station personnel implemented compensatory measures for out of service, degraded, or

inoperable fire protection equipment, as applicable, in accordance with procedures.

• Primary auxiliary building (PAB) southeast corner (PAB-F-2A-Z) on December 20

• PAB boric acid tanks and sample sink rooms (PAB-F-2B-Z) on December 20

• PAB primary component cooling water (PCCW) pump area (PAB-F-2C-Z) on

December 20

• PAB PCCW heat exchangers (PAB-F-3A-Z) on December 20

• PAB SW pipe slot (PAB-F-1K-Z) on December 20

b. Findings

No findings were identified.

1R06 Flood Protection Measures (71111.06 - 1 sample)

Internal Flooding Review

a. Inspection Scope

The inspectors reviewed the UFSAR, site flooding analysis, and plant procedures to

identify internal flooding susceptibilities for the site. The inspectors review focused on

the B residual heat removal (RHR) vault to verify the adequacy of equipment seals

located below the flood line, floor and wall penetration seals, watertight door seals,

common drain lines and sumps, sump pumps, level alarms, control circuits, and

temporary or removable flood barriers. The inspectors assessed the adequacy of

operator actions that NextEra had identified as necessary to cope with flooding in this

area and also reviewed the CAP to determine if NextEra was identifying and correcting

problems associated with both flood mitigation features and site procedures for

responding to flooding.

b. Findings

No findings were identified.

1R07 Heat Sink Performance (711111.07A - 1 sample)

a. Inspection Scope

The inspectors reviewed the A and B RHR heat exchanger to ensure readiness and

availability. The inspectors conducted a walkdown of the heat exchangers and reviewed

the results of the most recent performance test. The inspectors verified that NextEra

initiated appropriate corrective actions for identified deficiencies.

b. Findings

No findings were identified.

9

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

.1 Quarterly Review of Licensed Operator Requalification Testing and Training

(71111.11Q - 1 sample)

a. Inspection Scope

The inspectors observed licensed operator simulator annual requalification exams on

November 7, 2017, which included various failures, a transient resulting in an anticipated

transient without a scram, and a faulted steam generator requiring safety injection.

Another scenario included losing a feedwater pump, requiring a reactor scram, followed

by a loss of offsite power/loss-of-coolant accident. The inspectors evaluated operator

performance during the simulated event and verified completion of risk significant

operator actions, including the use of abnormal and emergency operating procedures.

The inspectors assessed the clarity and effectiveness of communications,

implementation of actions in response to alarms and degrading plant conditions, and the

oversight and direction provided by the control room supervisor. Additionally, the

inspectors assessed the ability of the crew and training staff to identify and document

crew performance problems.

b. Findings

No findings were identified.

.2 Quarterly Review of Licensed Operator Performance in the Main Control Room

(71111.11Q - 1 sample)

a. Inspection Scope

On October 19, 2017, the inspectors observed and reviewed routine activities in the

main control room. The inspectors observed operators respond to alarms, complete a

reactor coolant system (RCS) dilution, conduct a pre-job briefing for a surveillance test,

and perform the surveillance test. Additionally, the inspectors verified that procedure

use, crew communications, and coordination of activities between work groups met

established expectations and standards.

b. Findings

No findings were identified.

.3 Licensed Operator Requalification (71111.11A - 1 sample, 71111.11B - 1 sample)

a. Inspection Scope

The following inspection activities were performed using NUREG-1021, Operator

Licensing Examination Standards for Power Reactors, Revision 11, and Inspection

Procedure 71111.11, Licensed Operator Requalification Program.

10

Examination Results

On December 26, 2017, the results of the annual operating tests and biennial written

examinations were reviewed to determine if pass/fail rates were consistent with the

guidance of NUREG-1021, and NRC IMC 0609, Appendix I, Operator Requalification

Human Performance Significance Determination Process. The review verified that the

failure rate (individual or crew) did not exceed 20 percent.

• Five out of 42 operators failed at least one portion of requalification examination

(written, job performance measures (JPMs) or individual scenario failures). The

overall individual failure rate was 11.9 percent.

• One out of eight crews failed the simulator test. The crew failure rate was

12.5 percent

Written Examination Quality

The inspectors reviewed the written examinations administered to reactor operators

(ROs) and senior reactor operators (SRO) during the weeks 2, 4, and 5 of this cycle

(November-December 2017) for qualitative and quantitative attributes as specified in

Appendix B of Attachment 71111.11,

Operating Test Quality

Ten JPMs and five scenarios were reviewed for qualitative and quantitative attributes as

specified in Appendix C of 71111.11.

Licensee Administration of Operating Tests

Observations were made of the dynamic simulator exams and JPMs administered during

the week of December 4, 2017. These observations included facility evaluations of crew

and individual performance during the dynamic simulator exams and individual

performance of JPMs.

Examination Security

The inspectors assessed whether facility staff properly safeguarded exam material. The

JPMs, scenarios, and written examinations were checked for excessive overlap of test

items.

Remedial Training and Re-Examinations

The inspectors reviewed remediation plans and examinations for one crew failure during

the first quarter of 2016.

Conformance with Operator License Conditions

Medical records for six SRO licenses and four RO licenses were reviewed to assess

conformance with license conditions. All records reviewed were satisfactory.

11

Proficiency watch standing records for licensed operators were reviewed for the first

three quarters of 2017. All active licensed operators met the watch standing

requirements to maintain an active license.

The reactivation plan for licensed operators (three ROs and 13 SROs) were reviewed to

assess the effectiveness of the reactivation process. The reactivation was successfully

processed in accordance with site procedures.

Records for the participation of licensed operators in the requalification program for the

first three quarters in 2017 were reviewed.

Simulator Performance

Simulator performance and fidelity was reviewed for conformance to the reference plant

control room. A sample of simulator deficiency reports was also reviewed to ensure

facility staff addressed identified modeling problems. Simulator test documentation was

also reviewed.

Problem Identification and Resolution

A review was conducted of recent operating history documentation found in inspection

reports, the licensees CAP, and the most recent NRC plant issues matrix. The

inspectors also reviewed specific events from the licensees CAP which indicated

possible training deficiencies, to verify that they had been appropriately addressed.

These reviews did not detect any operational events that were indicative of possible

training deficiencies.

b. Findings

Introduction. The inspectors identified a Green NCV of 10 CFR 55.49, Integrity of

Examinations and Tests, for NextEras failure to ensure the integrity of the biennial

written examinations that were to be administered to licensed operators. This would

have resulted in examining Seabrook licensed operators with questions that had been

administered to other crews during the exam cycle that were in excess of the limits

established for question overlap.

Description. On December 6, 2017, while performing a biennial inspection in

accordance with IP 71111.11, Licensed Operator Requalification Program, the

inspectors determined that the written examination that was planned to be administered

that day for Crew E (and for Crew F in the following week) contained more than

50 percent of questions that had been used cumulatively to the licensed operators in the

previous 4 weeks of the same exam cycle.

NextEra Fleet Procedure TR-AA-220-1004, Licensed Operator Continuing Training

Annual Operating and Biennial Written Exams, Revision 2, requires that, Each biennial

comprehensive written exam version shall consist of at least 50 percent new, different, or

significantly modified test items compared to all previously administered versions of the

same exam. Since the procedure was not clear regarding the intent of this requirement,

the licensee incorrectly applied this to mean that there could be no more than 50 percent

overlap of questions in any one weeks examination with any other weeks examination

questions. In other words, the licensee was applying the question overlap criteria from

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 used on each examination, the examinations for Crews

E and F would have had 30 of 33 questions that had been previously used in this cycle.

According to 10 CFR 55.49, the integrity of a test or examination is considered

compromised if any activity, regardless of intent, affected or, but for detection, could

have affected the equitable and consistent administration of the test or examination. The

inspectors concluded that exceeding the 50 percent overlap was a failure to fulfill the

requirements of NextEras procedure and constituted a compromise of examination

integrity required by 10 CFR 55.49.

The inspectors informed the licensee of this overlap issue prior to the administration of

the written examination to Crew E. The licensee then postponed this written

examination until they could develop a written examination that did not violate the

overlap requirement. The first four written examinations of this 2017 cycle did use

common questions, but did not exceed the 50 percent overlap limit. Thus, there was no

actual effect on the equitable and consistent administration of the written examinations.

(Furthermore, the licensee has operators sign a security agreement to not reveal any

information about the requalification examinations with other operators who have not yet

taken their examinations.) During the previous comprehensive written examination in

2015, the examination developer used unique questions for each of the examinations in

that cycle. This years comprehensive written examination was developed by a different

individual who, along with other fleet personnel, misapplied the fleet procedures overlap

criteria. The licensee entered this issue into their CAP as AR 2239906.

Analysis. The failure of NextEras training staff to ensure the integrity of examinations

administered to licensed operations personnel was a performance deficiency. The

performance deficiency was a finding that was more than minor because, if left

uncorrected, the performance deficiency had the potential to lead to a more significant

safety concern. Specifically, the potential to allow operators to return to the control room

without valid demonstration of appropriate knowledge on the biennial written

examinations could result in having less than adequately qualified operators

manipulating plant controls in response to events. Using IMC 0609, Significance

Determination Process, and the corresponding Appendix I, Licensed Operator

Requalification Significance Determination Process, the finding was determined to have

very low safety significance (Green) because, although the examinations were not

administered, the integrity of an examination is considered to be compromised if any

activity affected, or but for detection, would have affected the equitable and consistent

administration of the examination. This finding had a cross-cutting aspect in the area of

Human Performance, Resources, in that leaders ensure procedures are available and

adequate to support nuclear safety. Specifically, NextEra established and implemented

a procedure that contained instructions to licensed operator biennial exam writers that

were unclear regarding regulatory guidance to limit written examination questions

overlap. [H.1]

Enforcement. Title 10 CFR 55.49, Integrity of Examinations, requires, in part, that

facility licensees shall not engage in any activity that compromises the integrity of any

test or examination required by this part. The integrity of a test or examination is

considered compromised if any activity, regardless of intent, affected or, but for

detection, could have affected the equitable and consistent administration of the test or

examination. This includes activities related to the preparation, administration, and

13

grading of the tests and examinations required by this part. Contrary to the above,

during the 2017 annual examination cycle (November through mid-December), NextEra

engaged in an activity at Seabrook that compromised the integrity of a test required by

10 CFR Part 55. Specifically, two scheduled written examinations would have contained

more than 50 percent of questions previously used in the cycle but for detection by the

NRC. Administering a written examination with greater than 50 percent cumulative

overlap from previously administered questions during a cycle is considered a

compromise of the integrity in that it is a practice that, but for detection, could affect the

equitable and consistent administration of the examination. The inspectors determined

that this overlap issue did not result in an actual effect on the equitable and consistent

administration of the written examinations. Because this finding was of very low safety

significance (Green) and has been entered into NextEras CAP as AR 2239906, this

violation is being treated as an NCV consistent with Section 2.3.2.a of the NRC

Enforcement Policy. (NCV 05000443/2017004-01 Licensed Operator Examination

Integrity Not Ensured)

1R12 Maintenance Effectiveness (71111.12Q - 1 sample)

a. Inspection Scope

The inspectors reviewed the samples listed below to assess the effectiveness of

maintenance activities on structure, system, and component (SSC) performance and

reliability. The inspectors reviewed system health reports, CAP documents,

maintenance WOs, and maintenance rule (MR) basis documents to ensure that NextEra

was identifying and properly evaluating performance problems within the scope of

the MR. For each sample selected, the inspectors verified that the SSC was properly

scoped into the MR in accordance with 10 CFR 50.65 and verified that the (a)(2)

performance criteria established by NextEra staff was reasonable. As applicable, for

SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective

actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that NextEra

staff was identifying and addressing common cause failures that occurred within and

across MR system boundaries.

• Boric acid control station

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 4 samples)

a. Inspection Scope

The inspectors reviewed station evaluation and management of plant risk for the

maintenance and emergent work activities listed below to verify that NextEra performed

the appropriate risk assessments prior to removing equipment for work. The inspectors

selected these activities based on potential risk significance relative to the reactor safety

cornerstones. As applicable for each activity, the inspectors verified that NextEra

personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the

assessments were accurate and complete. When NextEra performed emergent work,

the inspectors verified that operations personnel promptly assessed and managed plant

14

risk. The inspectors reviewed the scope of maintenance work and discussed the results

of the assessment with the stations probabilistic risk analyst to verify plant conditions

were consistent with the risk assessment. The inspectors also reviewed the TS

requirements and inspected portions of redundant safety systems, when applicable, to

verify risk analysis assumptions were valid and applicable requirements were met.

• Switchyard work, startup feed pump testing, and A emergency diesel generator

(EDG) maintenance and testing on October 17

• CS-FCV-111B fail to open during the period November 1-4

• Switchyard work and supplemental emergency power system maintenance on

November 20

• B solid state protection system Mode 1 actuation logic test on November 27

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments (71111.15 - 2 samples)

a. Inspection Scope

The inspectors reviewed operability determinations for the following degraded or

non-conforming conditions based on the risk significance of the associated components

and systems:

• A EDG fuel oil return line leaks on October 16

• D vital DC battery abnormal ammeter reading on November 15

The inspectors evaluated the technical adequacy of the operability determinations to

assess whether TS operability was properly justified and the subject component or

system remained available such that no unrecognized increase in risk occurred. The

inspectors compared the operability and design criteria in the appropriate sections of the

TSs and UFSAR to NextEras evaluations to determine whether the components or

systems were operable. The inspectors confirmed, where appropriate, compliance with

bounding limitations associated with the evaluations. Where compensatory measures

were required to maintain operability, the inspectors determined whether the measures

in place would function as intended and were properly controlled by NextEra.

b. Findings

No findings were identified.

1R19 Post-Maintenance Testing (71111.19 - 6 samples)

a. Inspection Scope

The inspectors reviewed the post-maintenance tests for the maintenance activities listed

below to verify that procedures and test activities adequately tested the safety functions

that may have been affected by the maintenance activity, that the acceptance criteria in

the procedure were consistent with the information in the applicable licensing basis

15

and/or design basis documents, and that the test results were properly reviewed and

accepted and problems were appropriately documented. The inspectors also walked

down the affected job site, observed the pre-job brief and post-job critique where

possible, confirmed work site cleanliness was maintained, and witnessed the test or

reviewed test data to verify quality control hold point were performed and checked, and

that results adequately demonstrated restoration of the affected safety functions.

• B CWT bistable card replacement on October 2

• C SW pump instantaneous overcurrent relay set point adjustment on October 16

• RC-V-2832, RCS sample valve relay replacement on November 2

• CS-FCV-111-B repairs on November 4

• Limitorque maintenance for CC-V-266 on November 28

• A fire pump annual maintenance on December 14

b. Findings

No findings were identified.

1R22 Surveillance Testing (71111.22 - 3 samples)

a. Inspection Scope

The inspectors observed performance of surveillance tests and/or reviewed test data of

selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR,

and NextEra procedure requirements. The inspectors verified that test acceptance

criteria were clear, tests demonstrated operational readiness and were consistent with

design documentation, test instrumentation had current calibrations and the range and

accuracy for the application, tests were performed as written, and applicable test

prerequisites were satisfied.

Upon test completion, the inspectors considered whether the test results supported that

equipment was capable of performing the required safety functions. The inspectors

reviewed the following surveillance tests:

• Power range channel 44 resealing calibration on October 4

• A PCCW pump on October 19 (in-service test)

• B charging pump surveillance on October 20

b. Findings

No findings were identified.

16

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation (71114.06)

.1 Emergency Preparedness Drill Observation (2 samples)

a. Inspection Scope

The inspectors evaluated the conduct of routine NextEra emergency drills on August 30,

2017, and November 29, 2017, to identify any weaknesses and deficiencies in the

classification, notification, and protective action recommendation development activities.

The inspectors observed emergency response operations in the simulator, technical

support center, and emergency operations facility (EOF) to determine whether the event

classification, notifications, and protective action recommendations were performed in

accordance with procedures. The inspectors also attended the station drill critique to

compare inspector observations with those identified by NextEra staff in order to

evaluate NextEras critique and to verify whether the NextEra staff was properly

identifying weaknesses and entering them into the CAP.

b. Findings

Introduction. The inspectors identified a Green NCV of 10 CFR 50.47(b)(14) and

10 CFR Part 50, Appendix E, Emergency Planning and Preparedness for Production

and Utilization Facilities,Section IV.F.2.g. Specifically, Seabrook did not identify and

critique a weakness associated with a RSPS during their critique following the

August 30, 2017, emergency preparedness drill.

Description. On August 30, 2017, Seabrook conducted an emergency preparedness

exercise, which included activating the simulator control room, the technical support

center (TSC), the operational support center, and the EOF. Consistent with the exercise

scenario script, a seismic event caused an RCS leak of approximately 300 gallons per

minute and resulted in the actuation of safety injection. The SED, located in the

simulator control room responded appropriately by declaring an emergency action level

(EAL) of Alert at 8:29 a.m. because the loss of the RCS barrier (loss of single fission

product barrier) threshold criterion was met. At 10:15 a.m., a containment release was

prematurely introduced by the simulator operator. This release was indicated on the

plant stack wide range gas monitor (WRGM) and the containment enclosure ventilation

area (CEVA) radiation monitor. The drill controllers recognized the error but did not

interject and allowed the events surrounding the premature simulated release to play

out.

At 10:19 a.m., consistent with the exercise scenario script, a second seismic event

occurred that resulted in a large break loss of coolant accident (LOCA). Plant conditions

deteriorated to the point that all ECCS necessary to inject for subsequent core cooling

had failed. This plant condition met the threshold for a potential loss of the fuel clad

barrier from a valid core cooling orange entry condition. The combination of the prior

loss of RCS barrier and the potential loss of the fuel clad barrier met the criteria for

classifying the event as a site area emergency (SAE). However, the SED, located in the

TSC, did not declare a SAE, but declared a GE at 10:28 a.m. The typical threshold for

declaring a GE is the loss of two barriers and the potential loss of the third. The SEDs

basis for concluding that the GE classification threshold criteria were met was the loss of

17

the RCS barrier, the potential loss of the fuel clad barrier, and the loss of the

containment barrier. The SED determined that a loss of containment barrier occurred

based on an unisolable pathway; however, no open pathway was scripted in the

exercise scenario and there were no valid indications that this was the case. The SED

concluded that the containment barrier was unisolable even though the radiological

release data at the time was well below the Environmental Protection Agencys (EPAs)

protective action guide (PAG) levels that are incorporated in the emergency plan.

Seabrooks emergency plan directs the comparison of radiological release data with

EPAs PAGs to inform decision making regarding whether a loss of containment barrier

exists.

Seabrook completed their formal drill critique on September 19, 2017. During the

critique, Seabrook did not identify that the declaration by the SED of a GE with protective

action recommendations (PARs) was based on insufficient information. Specifically,

Seabrooks EAL for the loss of the containment barrier is driven by a containment

isolation being required and either of the following: 1) Containment integrity has been

lost based on Short-Term Emergency Director (STED)/SED judgment, or 2) an

unisolable pathway from the containment to the environment exists. ER 1.1,

Classification of Emergencies, Revision 58, defines unisolable, as an open or breached

system line that cannot be isolated, remotely or locally.

Following the formal drill critique on September 19, 2017, the inspectors questioned the

basis for considering containment integrity lost, which resulted in characterizing the

circumstances present as a loss of the containment barrier. NextEra indicated that the

SED considered the loss of containment integrity was due to containment isolation being

required and the existence of an unisolable pathway from the containment to the

environment. NextEra noted that a containment isolation signal was received as

expected and all available remote indications showed the containment isolation valves

were closed. There were no other confirmed pathways open from containment to the

environment. The licensee also confirmed that containment pressures and pressure

trends were indeterminate with respect to the status of containment integrity. The

licensee validated after the exercise that the higher than normal WRGM readings

indicated noble gases that could only come from damaged nuclear fuel inside

containment; however, the containment post-LOCA radiation monitors were reading

relatively normal with no indication of damaged fuel.

As planned by the exercise scenario script, a containment recirculation sump isolation

valve CBS-V-8, was not opening when required, to place the containment on

recirculation cooling, which led the SED to suspect the penetration and its encapsulated

valve were the possible locations of an unisolable pathway. This determination by the

SED is noteworthy, because the control room operators had confirmed that the valve

was closed based on remote indication. The status of this valve, encapsulation tank,

and penetration line were not validated locally. Taking into account the seismic events

that caused the large break LOCA, the suspect encapsulated valve, higher than normal

WRGM readings and CEVA radiation levels, the SED concluded that a loss of

containment integrity, as defined in their EAL scheme and basis, existed. A GE

declaration was made due to the loss of containment conclusion and the previously

determined potential loss of fuel clad and the loss of the RCS barrier (versus the

originally scripted SAE). Due to the lack of any other valid indications that containment

integrity was jeopardized, the SED relied upon the radiological releases seen on the

WRGM and CEVA radiation monitor as positive indication of the loss of the containment

18

barrier. The fission product barrier EAL (FG1) allows the SED to use judgment to make

a determination of containment barrier integrity based on less discrete information.

Specifically, 4.A.1 states that containment integrity has been lost when the actual

containment atmospheric leak rate likely exceeds that associated allowable leakage.

However, Seabrooks procedure, ER 1.1 states, it is expected that the SED will assess

the threshold using judgment, and with due consideration given current plant conditions,

and available operational and radiological data.

The inspectors determined that, as a result of the deviation from the preplanned

scenario script and due to the actual condition experienced during the exercise, the GE

declaration would have been an appropriate event classification if it had been based on

SED judgment instead of an unisolable pathway. The conditions presented at the time

could have warranted the use of judgement to escalate from an SAE to a GE based on

imminent fuel melt and the uncertainty recognized by the SED, regarding the fuel

condition based on radiation monitors indicating a release outside the containment.

Therefore, the GE threshold criteria (loss of two and the potential loss of the third fission

product barriers) would have been met by the loss of the RCS barrier, the potential loss

of the containment barrier and judgement that the loss of the fuel clad barrier was

imminent.

As a result, in accordance with IMC-0609, Appendix B, Emergency Preparedness

Significance Determination Process, the performance demonstrated by NextEra

participants in the drill, provided specific opportunities that could preclude effective

implementation of the emergency plan that the inspectors concluded was a weakness.

In addition, the inspectors also identified deficiencies associated with the Emergency

Classification system RSPS under 10 CFR 50.47(b)(4). These deficiencies involved the

less than adequate translation of specific guidance incorporated into the Seabrook EAL

basis document during implementation of a recent upgrade to the Seabrook emergency

plan to incorporate a revision (5 to 6) to Nuclear Energy Institute (NEI) Document 99-02,

Development of Emergency Action Levels for Non-Passive Reactors. Moreover, the

inspectors determined that the requisite training for decision-makers for the most

relevant portion of the revised guidance, was also developed and provided in a less than

adequate manner. More importantly, the germane sections of the revised guidance

associated with the Containment Barrier portion of the Fission Product Matrix EALs were

directly exercised during the August 2017 drill.

Analysis. The inspectors determined that not identifying an exercise weakness related

to a GE classification based on insufficient information during the exercise critique was a

performance deficiency that was reasonably within the ability of Seabrook to foresee and

prevent. The finding is more than minor because it is associated with the ERO attribute

of the Emergency Preparedness Cornerstone and affected the cornerstone objective to

ensure that the licensee is capable of implementing adequate measures to protect the

health and safety of the public in the event of a radiological emergency. Specifically,

Seabrook personnel did not identify an exercise weakness associated with a RSPS

when the incorrect basis for a GE declaration was used by the SED.

The inspectors assessed the finding using IMC 0609, Attachment 4, Initial

Characterization of Findings, issued October 7, 2016. This attachment directs

inspectors to use IMC 0609, Appendix B, Emergency Preparedness Significance

Determination Process, issued September 22, 2015, because the finding and the

19

associated weakness are in the emergency preparedness cornerstone. Inspectors

determined the finding was a critique finding, the drill scope was full scale, the planning

standard was risk-significant, and the performance opportunity was a success. As a

result, and using figure 5.14-1, Significance Determination for Critique Findings, the

inspectors determined this finding was of very low safety significance (Green).

The finding is related to the cross-cutting area of Human Performance, Change

Management in that leaders use a systematic process for evaluating and implementing

change so that nuclear safety remains the overriding priority. Specifically, although

recent changes to the sites emergency classification and action level standard scheme

were effective on July 2017, the new EAL procedure and training regarding the changes

lacked sufficient specificity to ensure the users understood the new scheme with respect

to the status of the containment integrity [H.3].

Enforcement. Title 10 CFR 50.54(q)(2) requires, in part, that a licensee shall follow and

maintain the effectiveness of an emergency plan that meets the requirements in

Appendix E to this part and, for nuclear power reactor licensees, the planning standards

of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(14) requires, in part, that periodic exercises

be conducted to evaluate major portions of emergency response capabilities and that

deficiencies identified as a result of exercises are corrected. Section lV.F.2.g of

Appendix E to 10 CFR Part 50 requires that all training, including exercises, shall

provide for formal critiques in order to identify weak or deficient areas that need

correction. Any weaknesses or deficiencies that are identified shall be corrected.

Contrary to the above, during a formal critique on September 19, 2017, Seabrook did not

identify a weakness needing correction that was demonstrated during a full participation

exercise on August 30, 2017. The weakness needing correction involved NextEras

declaration of a GE that was based on insufficient information. Because this violation

was of very low safety significance and was entered into Seabrooks CAP as

AR 2242073, this finding is being treated as an NCV consistent with Section 2.3.2 of the

NRC Enforcement Policy. (NCV,05000443/2017004-02, Failure of Exercise Critique

to Identify a Risk Significant Planning Standard Weakness)

.2 Training Observations (1 sample)

a. Inspection Scope

The inspectors observed a simulator training evolution for licensed operators on

November 7, 2017, which required emergency plan implementation by an operations

crew. NextEra planned for this evolution to be evaluated and included in the drill and

exercise performance indicator (PI) data. The inspectors observed event classification

and notification activities performed by the crew. The focus of the inspectors activities

was to note any weaknesses and deficiencies in the crews performance and ensure that

NextEra evaluators noted the same issues and entered them into the CAP.

b. Findings

No findings were identified.

20

2. RADIATION SAFETY

Cornerstone: Public Radiation Safety

2RS2 Occupational As Low As Is Reasonably Achievable Planning and Controls

(71124.02 - 1 sample)

a. Inspection Scope

The inspectors assessed NextEras performance with respect to maintaining

occupational individual and collective radiation exposures as low as is reasonably

achievable (ALARA). The inspectors used the requirements contained in 10 CFR

Part 20, applicable Regulatory Guides (RGs) 8.8 and 8.10, TSs, and procedures

required by TSs as criteria for determining compliance.

Verification of Dose Estimates and Exposure Tracking Systems

The inspectors reviewed the current annual collective dose estimate; basis methodology;

and measures to track, trend, and reduce occupational doses for ongoing work activities.

The inspectors evaluated the adjustment of exposure estimates, or re-planning of work.

The inspector reviewed post-job ALARA evaluations of excessive exposure.

b. Findings

No findings were identified.

2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03 - 1 sample)

a. Inspection Scope

The inspectors reviewed the control of in-plant airborne radioactivity and the use of

respiratory protection devices in these areas. The inspectors used the requirements in

10 CFR Part 20, RG 8.15, RG 8.25, NUREG/CR-0041, TS, and procedures required by

TS as criteria for determining compliance.

Self-Contained Breathing Apparatus for Emergency Use

The inspectors reviewed the following: the status and surveillance records for three

Self-Contained Breathing Apparatus (SCBAs) staged in-plant for use during

emergencies; Next Eras SCBA procedures and maintenance and test records; the

refilling and transporting of SCBA air bottles; SCBA mask size availability; and the

qualifications of personnel performing service and repair of this equipment.

b. Findings

No findings were identified.

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4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

.1 Mitigating Systems Performance Index (3 samples)

a. Inspection Scope

The inspectors reviewed NextEras submittal of the Mitigating Systems Performance

Index for the following systems for the period of July 1, 2017, through June 30, 2018:

• Safety System Functional Failures

• RHR System

• Cooling Water System

To determine the accuracy of the PI data reported during those periods, the inspectors

used definitions and guidance contained in NEI Document 99-02, Regulatory

Assessment Performance Indicator Guideline, Revision 7. The inspectors also

reviewed NextEras operator narrative logs, mitigating systems performance index

derivation reports, event reports, and NRC integrated inspection reports to validate the

accuracy of the submittals.

b. Findings

No findings were identified.

.2 Occupational Exposure Control Effectiveness (1 sample)

a. Inspection Scope

The inspectors reviewed licensee submittals for the occupational radiological

occurrences PI for the fourth quarter 2016 through the first, second, and third quarters

2017. The inspectors used PI definitions and guidance contained in the NEI Document

99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, to

determine the accuracy of the PI data reported. The inspectors reviewed electronic

personal dosimetry accumulated dose alarms, dose reports, and dose assignments for

any intakes that occurred during the time period reviewed to determine if there were

potentially unrecognized PI occurrences.

b. Findings

No findings were identified.

.3 Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual

Radiological Effluent Occurrences (1 sample)

a. Inspection Scope

The inspectors reviewed licensee submittals for the radiological effluent technical

specifications/offsite dose calculation manual radiological effluent occurrences PI for the

fourth quarter 2016 through the first, second, and third quarters of 2017. The inspectors

22

used PI definitions and guidance contained in the NEI Document 99-02, Regulatory

Assessment Performance Indicator Guideline, Revision 7, to determine if the PI data

was reported properly. The inspectors reviewed the public dose assessments for the PI

for public radiation safety to determine if related data was accurately calculated and

reported.

The inspectors reviewed the CAP database to identify any potential occurrences such as

unmonitored, uncontrolled, or improperly calculated effluent releases that may have

impacted offsite dose. The inspectors reviewed gaseous and liquid effluent summary

data and the results of associated offsite dose calculations to determine if indicator

results were accurately reported.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution (71152 - 3 samples)

.1 Routine Review of Problem Identification and Resolution Activities

a. Inspection Scope

As required by Inspection Procedure 71152, Problem Identification and Resolution, the

inspectors routinely reviewed issues during baseline inspection activities and plant

status reviews to verify NextEra entered issues into the CAP at an appropriate threshold,

gave adequate attention to timely corrective actions, and identified and addressed

adverse trends. In order to assist with the identification of repetitive equipment failures

and specific human performance issues for follow-up, the inspectors performed a daily

screening of items entered into the CAP and periodically attended CR screening

meetings. The inspectors also confirmed, on a sampling basis, that, as applicable, for

identified defects and non-conformances, NextEra performed an evaluation in

accordance with 10 CFR Part 21.

b. Findings

No findings were identified.

.2 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a semi-annual review of site issues to identify trends that

might indicate the existence of more significant safety concerns. As part of this review,

the inspectors included repetitive or closely-related issues documented by NextEra in

quarterly trend reports, site PIs, major equipment problem lists, system health reports,

MR assessments, and maintenance or CAP backlogs. The inspectors also reviewed

NextEras CAP database for the third and fourth quarters of 2017 to assess CRs written

in various subject areas (equipment problems, human performance issues, etc.), as well

as individual issues identified during the NRCs daily CR review (Section 4OA2.1). The

inspectors reviewed the NextEra trend reports for the previous six months of 2017,

conducted under PI-AA-207-1000, Station Self-Evaluation and Trend Analysis,

23

Revision 8, to verify that NextEra personnel were appropriately evaluating and trending

adverse conditions in accordance with applicable procedures.

b. Findings and Observations

No findings were identified.

Overall, the inspectors noted that the system health reports for the safety related

systems and systems important to safety to be up to date and reflective of current plant

status. The health reports were reflective of issues that were trending on the daily plant

status report and discussed on a regular basis by plant management for timely

resolution. The inspectors evaluated a sample of CRs generated over the course of the

past two quarters by departments that provide input to the quarterly trend reports. The

inspectors determined that, in most cases, the issues were appropriately evaluated by

Seabrook staff for potential trends and resolved within the scope of the CAP. Moreover,

the inspectors identified instances where potential adverse trends were identified by

department staff during the course of the assessment period, which were consistent with

similar station-level trends, and confirmed that station personnel were utilizing statistical

and trending tools to identify potential emerging trends. Additionally, the inspectors

verified that discussions between department and performance improvement staff were

occurring to ensure emerging trends were appropriately captured either in the CAP or

the quarterly trend report, as applicable. One such example was an issue with the

overall health of the preventive maintenance program, which included implementation

and knowledge issues following a program assessment documented under CR 2219903.

.3 Annual Sample: Ultimate Heat Sink

a. Inspection Scope

The inspectors performed an in-depth review of NextEras evaluations and corrective

actions associated with the ultimate heat sink over the last year, which includes the

ocean SW system, CWT, and PCCW system. This included degraded piping and leaks,

PCCW pump motor issues, and increasing SW pump motor winding temperatures.

The inspectors assessed NextEras problem identification threshold, cause analyses,

extent of condition reviews, compensatory actions, and the prioritization and timeliness

of NextEras corrective actions to determine whether NextEra was appropriately

identifying, characterizing, and correcting problems associated with this issue and

whether the planned or completed corrective actions were appropriate. The inspectors

compared the actions taken to the requirements of NextEras CAP and 10 CFR Part 50,

Appendix B.

b. Findings and Observations

No findings were identified.

NextEra was timely in documenting issues once they were identified and screened

appropriately for immediate operability concerns. For example, control room operators

noted an increased trend in SW pump motor winding temperatures. It did not

immediately impact the safe operation of the plant, but the issue was captured in the

CAP and the motors were systematically replaced in a timely manner.

24

An outstanding issue continues to be degraded SW piping associated with the ocean

SW and the cooling water systems. NextEra has a systematic program, reflected in

PEG-94, Service Water Inspection and Repair Trending, to ensure that long term

corrective actions are implemented to minimize unexpected leaks and challenges to the

safe operation of the plant. The inspectors verified that PEG-94 is continuously updated,

and pipe inspections and replacements are completed as scheduled. When unexpected

leaks did occur, the station demonstrated timely assessment and appropriate

compensatory measures until final corrective actions to restoration were feasible.

The inspectors noted that NextEra implemented industry initiatives to improve the

effectiveness of issue resolution, also known as CAP-002, in August 2017. The changes

are reflected in PI-AA-104-1000, Condition Reporting. The inspectors have been

closely monitoring the impact to ensure issues important to nuclear safety are addressed

appropriately. No concerns have been noted by the inspectors to date.

.4 Annual Sample: Alkali-Silica Reaction

a. Inspection Scope

The purpose of periodic site visits to Seabrook Station over the past few years has been

to review the adequacy of NextEras monitoring of alkali-silica reaction (ASR) on affected

reinforced concrete structures, per their 10 CFR 50.65 Maintenance Rule Structures

Monitoring Program (SMP), and NextEras corrective action process. In addition, the

inspectors verify on a sampling basis that significant changes or different manifestations

of ASR on the affected structures are appropriately considered for impact on the

Seabrook prompt operability determinations for the affected structure(s). Two NRC

region-based inspectors and a structural engineer from the Office of Nuclear Reactor

Regulation were on site from October 10-13, 2017, to conduct an inspection of ongoing

ASR related activities. The inspectors also conducted in-office reviews of ASR-related

documentation made available before and after the on-site inspection via an electronic

server (Certrec Inspection Management System). Although available for review, the

inspectors did not receive or take possession of these documents.

The inspectors assessed the problem identification threshold, operability and

functionality assessments, extent of condition reviews, and the prioritization and

timeliness of corrective actions to determine whether NextEra personnel were

appropriately identifying, characterizing, and correcting problems associated with the

ASR-affected structures. The inspectors evaluated NextEras actions to verify

compliance with the SMP, the CAP, and 10 CFR Part 50, Appendix B requirements.

b. Findings and Observations

No findings were identified.

The inspectors performed a review of the CEVA north wall operability determination,

including a field walkdown of the structure. The North wall is laterally deformed below

the CEVA heating, ventilation, and air conditioning (HVAC) room floor slab as measured

by the plumbness. NextEra has preliminarily concluded the movement at this location is

the result of ASR expansion of the concrete backfill confined between the wall and the

adjacent bedrock, which is a load that was not considered in the original design of the

25

wall in accordance with American Concrete Institute (ACI) 318-71. The out-of-plumb

wall section is located between the +3 and +19 foot elevation and exhibits visual

horizontal flexure cracks with evidence of delamination (identified via hammer testing) in

the vicinity of the cracks. The cracks are spaced at approximately 1 foot intervals, which

is the same spacing as the horizontal reinforcing bars. The detected delaminations were

found around the horizontal cracks where the largest displacement is occurring on the

order of approximately 1.5 inches. An initial SMP structural evaluation by NextEra staff

(simple beam finite element analysis) was performed, and with the estimated

compressive strains in the concrete in some areas and the opposing tensile strains in

the rebar in other sections, the analysis concluded that delamination is predicted.

Subsequently, a nonlinear finite element analysis based on the deformed shape of the

wall was performed by NextEra to determine the maximum allowable lateral

displacement before a modification is necessary. The inspectors reviewed this analysis

as part of the operability determination and determined that NextEras conclusions that

the structure is capable of performing its intended functions was technically supported.

The inspectors further verified that SMP Appendix C was updated with additional

qualitative monitoring requirements for the CEVA building. Discussions with the

responsible NextEra engineering staff identified that remediation methods are being

evaluated to ensure long-term continued stabilization and structural performance of the

wall. The inspectors noted that this lower portion of the north wall was identified as a

non-structural member for the CEVA structure (i.e., not part of the structural load

resisting system for the CEVA) and is not part of the boundary that establishes the

safety-related CEVA air envelope. However, the wall is required to maintain its

structural stability because it supports attached equipment.

Inspectors walkdown of the RHR/containment spray (CS) Vault confirmed the presence

of several small areas of delamination. Review of FP101055, Condition Assessment of

Cracking in RHR and CS Equipment Vault - Second Visit, dated February 4, 2016,

summarizes the results of a detailed examination of the RHR/CS Vault by NextEras staff

contractors following an earlier examination in December 2014. One of the

recommendations in FP101055 was to remove cores from areas exhibiting delamination

to better understand the extent of concrete degradation. At the request of the

inspectors, NextEra posted the results of concrete coring and associated petrographic

examination (FP101034) on their electronic server (Certrec Inspection Management

System) for review. FP101034 summarizes the petrographic examination of 19 core

samples and their associated bore holes. The examination results identified that all of

the cores taken from the external walls exhibited signs of ASR, whereas the cores taken

from the interior walls did not. The large cracks observed in the interior walls were likely

a result of upward expansion due to ASR in the exterior walls, which transferred the

resulting tension to the interior walls of the Equipment Vault. The inspectors noted that

there were no discussions on the surface delamination areas or confirmation of the

depth of delamination as was recommended in earlier reports.

The identification of delamination as either a primary (caused by internal ASR expansion

in the wall) or secondary (caused by ASR expansion of concrete backfill and associated

loading) effect of ASR is preliminarily being reviewed by the NRC inspectors as a

phenomenon associated with ASR based on plant operating experience. At the

conclusion of the on-site inspection, NextEra staff had not drawn conclusions regarding

the implications of delamination associated with ASR expansion and loading. Based

upon the inspectors initial assessment, NextEra decided to develop criteria for

identifying and monitoring delamination of ASR-affected structures and how best to use

26

hammer testing or other non-destructive examination methods (e.g., impact-echo

testing), which was captured as an action in their Change Management Plan for the

SMP. The SMP currently does not describe hammer testing or include delamination

monitoring guidance, and NextEra had not specifically identified this ASR phenomenon

in the structures Aging Management Program for their license renewal application.

On November 22, 2017, NextEra provided the inspectors with an assessment of

ASR-related delamination, to date, that concluded the delamination areas were a result

of loading on the wall and were limited to the cover concrete layer (near surface), and

therefore, not relevant to structural performance. NextEra staff planned to perform

impact-echo testing, a non-destructive test method that uses sound waves to detect

flaws within the concrete, to verify that delamination is only occurring in the cover

concrete. If delaminations deeper than the cover are identified, then NextEra staff

indicated that cores would be taken to verify the condition of the concrete. The

inspectors determined that this proposed validation plan was technically adequate to

assess the implications of delamination.

Consistent with the current SMP, the B Electrical Tunnel Stage 1 structural evaluation

was recently completed. NextEra staff concluded that by including an assumed ASR

loading from the concrete backfill in the building design shear capacity calculations, the

calculated electrical tunnel wall loading (assumed demand) exceeds the design capacity

and would not conform to established standards in the ACI 318-71 structural design

code. To address this non-conforming condition, NextEra wrote a separate operability

determination and initiated further engineering evaluations to review the ASR backfill

loading assumptions and to consider potential remediation methods for the B Electrical

Building, including support struts and/or bolted plates. The inspectors noted that there

are no visual indications of loading distress or other structural integrity issues as evident

by the absence of structural cracks. The inspectors conducted a conference call with

NextEra staff and their principle ASR engineering contractor (SG&H) on October 18,

2017, to better understand the assumed backfill loading profiles used by NextEra staff in

the structural evaluations. The inspectors were informed that the concrete backfill

loading profiles differ for each Seabrook structure and that these profiles were

developed by a seven step iterative process. Based upon this conference call, the

inspectors understand that NextEra staff used as-built drawings with backfill details to

develop the initial ASR load profiles, taking into consideration whether or not the

concrete backfill was confined or unrestrained by any overburden or adjoining excavated

surfaces. If appropriate, the backfill load profile adjustments were made utilizing field

observations. Examination of NextEras methodology for assessing concrete backfill

loading is currently under review by the NRC staff, as an element of the August 1, 2016,

License Amendment Request (16-03).

Based upon discussions with the responsible engineering staff and inspector review of

the Structures Monitoring Program Manual (SMPM), the inspectors understand that as

Stage 1, 2, and 3 structural susceptibility evaluations are completed, NextEra staff intend

to update SMPM, Appendix C, Building Deformation Monitoring Tables, with critical

structural monitoring points (qualitative and/or quantitative) that are deemed appropriate

to effectively monitor ASR impacts and progression for each affected structure. The

inspectors also reviewed the current Change Management Plan for the SMP (AR No.

02148021, dated October 11, 2017), which identified numerous pending changes that

were being tracked for the next revision to the SMPM. Revision 03, dated November 17,

2017, was approved after the end of the inspection. The inspectors verified that the

27

monitoring points for the recently completed structural evaluations were added to

Appendix C.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 1 sample)

(Closed) Licensee Event Report (LER) 05000443/2017-001-00: Manual Reactor Trip in

Response to a Feedwater Isolation due to High Level in Steam Generator B

a. Inspection Scope

The inspectors reviewed the LER, root cause analysis, and event analysis, following the

April 29, 2017, plant trip, due to steam generator water level perturbations. Additionally,

the inspectors reviewed follow-up actions related to the event to assure that NextEra

staff implemented appropriate corrective actions commensurate with their safety

significance. The enforcement actions associated with this LER are discussed below.

This LER is closed.

b. Findings

Introduction. A self-revealing Green finding was identified for inadequate

implementation of procedure MA 4.5, Configuration Control, Revision 18. Specifically,

maintenance technicians failed to properly implement MA 4.5 while backfilling steam

generator instrumentation, and inadvertently left an instrumentation valve partially open

instead of fully open. This resulted in slow response of the instrument, and ultimately a

high steam generator level, a feedwater isolation signal and a manual reactor trip.

Description. On April 29, control room operators manually tripped the reactor when the

B steam generator level reached the feedwater isolation signal setpoint. The plant was

at approximately 12 percent power, and operators were raising power in preparation for

main generator synchronization. At the time, feedwater was being manually controlled

by the operators, and the wide range steam generator level indication was being used to

determine the required feedwater flow. The wide range level indication was responding

slowly to level changes which resulted in overfeeding the steam generator. This caused

the steam generator level to increase to the feedwater isolation signal setpoint.

NextEra personnel determined that the slow response of the steam generator level

indication was due to an instrumentation valve left partially open instead of fully open as

required. On April 26, instrumentation and control technicians had performed a

backfilling of the steam generator reference legs. The technicians used procedure

MA 4.5, including Form MA 4.5A, Configuration Change, to track the valve

manipulations to maintain configuration control. MA 4.5 requires that all component

manipulations and changes to component and plant configuration are performed only to

a detailed procedure or written instruction, and shall be documented on form MA 4.5A or

in an operating procedure WO, or job plan. The technicians did not properly use

place-keeping and concurrent verification during the performance of the backfilling

activity, and one instrumentation valve was left in a nearly full closed position instead of

the full open position. NextEra promptly rechecked other similar valves, then performed

a root cause evaluation that eventually led to additional technician training and improved

configuration controls during such evolutions.

28

Analysis. The inspectors determined that NextEras failure to properly implement

MA 4.5 was a performance deficiency within NextEras ability to foresee and correct, and

should have been prevented. Specifically, instrumentation and control technicians failed

to open an instrumentation valve at the end of a steam generator level indicating system

backfill maintenance activity. This resulted in operators unable to properly control steam

generator water level during startup operations, and ultimately led to a required plant trip

due to high steam generator level and a feedwater isolation signal.

This finding is more than minor because it is associated with the configuration control

attribute of the Initiating Events cornerstone and affected the cornerstone objective to

limit the likelihood of events that upset plant stability and challenge critical safety

functions during shutdown as well as power operations. Specifically, the failure to

effectively implement MA 4.5 resulted in a valve being left out of its required position, a

subsequent lack of steam generator water level control during low power operations, and

ultimately required a manual reactor trip. Additionally, the finding is similar to

Example 4.b of IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples

of Minor Issues, issued August 11, 2009, in that the performance deficiency caused a

reactor trip. In accordance with IMC 0609.04, Initial Characterization of Findings,

issued June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The Significance

Determination Process for Findings At-Power, issued June 19, 2012, the inspectors

determined that this finding is of very low safety significance (Green), because the

finding did not cause a reactor trip and the loss of mitigation equipment relied upon to

transition the plant from the onset of a trip to a stable shutdown condition.

In accordance with IMC 0310, the finding has a cross-cutting aspect in the area of

Human Performance, Work Management, because the organization did not implement a

process of planning, controlling, and executing the work activity such that nuclear safety

was the overriding priority. Specifically, NextEra did not ensure that a steam generator

backfilling activity was properly executed, which resulted in the slow response of a

steam generator level indication, the overfeeding of the steam generator, a feedwater

isolation signal, and the ultimate requirement to trip the reactor [H-5].

Enforcement. This finding does not involve enforcement action because no violation of a

regulatory requirement was identified. Because this finding does not involve a violation

and is of very low safety significance, it is identified as a finding.

(FIN 05000443 /2017004-03, Inadequate Procedure Implementation Results in a

Manual Reactor Trip)

4OA6 Meetings, Including Exit

On January 23, 2018, the inspectors presented the inspection results to Mr. Eric

McCartney, Regional Vice President, Northern Region, and other members of the

Seabrook Station staff. The inspectors verified that no proprietary information was

retained by the inspectors or documented in this report.

ATTACHMENT: SUPPLEMENTARY INFORMATION

A-1

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

E. McCartney, Regional Vice President, Northern Region

C. Domingos, Site Director

K. Boehl, Senior Radiation Protection Analyst

K. Browne, Licensing Manager

E. Carley, License Renewal Supervisor

A. Giotos, Senior Analyst

J. Hulbert, Nuclear Engineer

D. Robinson, Chemistry Manager

D. Strand, Radiation Protection Manager

T. Smith, Radiation Protection Supervisor

C. Thomas, Licensing Engineer

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened/Closed

05000443/2017004-01 NCV Licensed Operator Examination Integrity Not

Ensured (Section 1R11.3)05000443/2017004-02 NCV Failure of Exercise Critique to Identify a RSPS

Weakness (Section 1EP6)05000443/2017004-03 FIN Inadequate Procedure Implementation Results in

a Manual Reactor Trip (Section 4OA3)

Closed

05000443/2017-001-00 LER Manual Reactor Trip in Response to a

Feedwater Isolation due to High Level in Steam

Generator B (Section 4OA3)

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Procedures

OP-AA-102-1002, Seasonal Readiness, Revision 20

Condition Reports

2225659 2227085 2227175

Maintenance Orders/Work Orders

40500528

Attachment

A-2

Miscellaneous

Seabrook Station certification of seasonal readiness, Winter 2017-2018, dated 9/22/17

Section 1R04: Equipment Alignment

Procedures

OS0443.36, Fire Pump House Weekly Valve Alignment, Revision 6

OS1016.03, A Service Water Operation, Revision 17

OS1016.04, B Service Water Operation, Revision 20

OS1016.05, Service Water Cooling Tower Operation, Revision 34

OX1416.01, Service Water Monthly Valve Verification, Revision 12

OX1416.05, Service Water Quarterly Operability Test Cooling Tower Pump, Revision 27

OX1416.03, Cooling Tower Fan Monthly Operability Test, Revision 10

OX1456.02, ECCS Monthly System Verification, Revision 20

Miscellaneous

UFSAR 9.2.1, Revision 18

Drawings

1-CS-B20725, Chemical & Volume Control Charging System Detail, Revision 32

1-CS-B20729, Chemical & Volume Control System Boric Acid Detail, Revision 20

1-FP-B20266, Fire Protection Fire Pump House Detail, Revision 25

1-SI-B20446, Safety Injection System Intermediate head Injection System Detail, Revision 18

1-SW-B20792, Service Water System Nuclear Overview, Revision 6

1-SW-B20794, Service Water System Nuclear Detail, Revision 39

1-SW-B20795, Service Water System Nuclear Detail, Revision 44

Section 1R05: Fire Protection

Miscellaneous

Seabrook Station Fire Protection Pre-Fire Strategies, Volume 1

Section 1R06: Flood Protection Measures

Procedures

OS1212.01, PCCW System Malfunction, Revision 13

OS1213.01, Loss of RHR During Shutdown Cooling Revision 19

OP-AA-109, Control of Time Critical Operator Actions and Time Sensitive Actions, Revision 2

Miscellaneous

Report TP-7, Seabrook Station Moderate Line Break Study, Revision 5

UFSAR Section 3.6B, Revision 8; Section 9.2, Revision 14

Section 1R07: Heat Sink Performance

Miscellaneous

A RHR Heat Exchanger Performance Monitoring Data from OR18

B RHR Heat Exchanger Performance Monitoring Data from OR18

EPRI NP-7552, Heat Exchanger Performance Monitoring Guidelines

A-3

Drawings

1-CC-B20204, Primary Component Cooling Loop A Overview, Revision 4

1-CC-B20205, Primary Component Cooling Loop A Detail, Revision 26

1-RH-B20660, Residual Heat Removal System Overview, Revision 3

1-RH-B20663, Residual Heat Removal System Train B Cross-tie Detail, Revision 21

1-RH-B20660, Residual Heat Removal System Overview, Revision 3

9763-F-805203, PAB Vaults Piping Zone 30D Plan at EL(-) 9-0, Revision 12

Section 1R11: Licensed Operator Requalification Program

Procedures

OP 9.2, Transient Response Procedure Users Guide, Revision 18

OP-AA-100-1001, License Maintenance and Activation, Revision 4

TR-AA-220-1004, Licensed Operator Continuing Training Annual Operating and Biennial

Written Exams, Revision 2

TR-AA-230-1007, Conduct of Simulator Training and Evaluation, Revision 5

Condition Reports

2114495 2117035 2202358

Miscellaneous

Seabrook 2016-2017 Requalification Training Program Annual Examination Sample Plan

Simulator-Related Test Documents

NT-3730-1, SBT Package for L15R11, Rev. 11, dated 9/23/16

NT-3730-1, Seabrook Transient No. 1, Manual Reactor Trip, Rev. 17, dated 3/25/17

NT-3730-1, Seabrook Transient No. 11, Large Break LOCA with Loss of Offsite Power, Rev. 17

dated 3/16/17

NT-3730-1, Seabrook Transient No. 2, Simultaneous Trip of Both Main Feedwater Pumps,

Rev. 17, dated 3/25/17

NT-3730-1, Seabrook Transient No. 3, Simultaneous Closure of All Main Steam Isolation Valves,

Rev. 17, dated 3/25/17

NT-3730-1, Seabrook Transient No. 6a, Main Turbine Trip Below the P-9 Permissive, Rev. 17,

dated 3/25/17

NT-3730-1, Seabrook Transient No. 8, Slow Primary Depressurization, Rev. 17, dated 3/15/17

NT-3730-1, Steady State Value Comparison Test - 100% Power, Rev. 17, dated 5/16/17

NT-3730-1, Steady State Value Comparison Test - 46% Power, Rev. 17, dated 9/13/16

NT-3730-1, Steady State Value Comparison Test - 79% Power, Rev. 17, dated 5/15/17

NT-3730-1, Steady State Value Comparison Test - Post Event Test A Water Box Isolated,

Rev. 17 dated 1/20/16

Section 1R12: Maintenance Effectiveness

Condition Reports

021307 222005 592531 1682547 2234042 2234311

Maintenance Orders/Work Orders

4052273 40125669 40568790

Miscellaneous

EC 273524

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, Revision 27

MA-AA-100-1011, Equipment Troubleshooting, Revision 3

MA3.5, Post Maintenance Testing, Revision 23

OX0443.01, Diesel Fire Pump Weekly Test, Revision 16

OX1456.81, Operability Testing of IST Valves, Revision 29

OX1456.86, Operability Testing of IST Pumps, Revision 15

OX1490.05, Miscellaneous Systems ASME Quarterly Testing, Revision 9

A-5

Condition Reports

0289856 2227780 2230622 2234042 2238019 2238020

2238038 2238053 2240790

Maintenance Orders/Work Orders

40189098 40496829 40497318 40516877 40531737 40563635

40568790 94170738

Miscellaneous

ECs 288964, 286645

Calculation 9763-3-ED-00-23-F, Medium Voltage Protective Relay Coordination, Revision 5

Drawings

1-NHY-250000, Revision 83

1-NHY-506839, Service Water Pumps P-41B & P41D Control Loop Diagram, Revision 9

Section 1R22: Surveillance Testing

Procedures

IX1656.938, NI-N-44 Power Range NI Rescaling Calibration, Revision 12

OPMM, Operations, Management Manual, Revision 107

OS1412.13, PCCW Train A Quarterly Operability, 18 Month Position Indication, and

Comprehensive Pump Testing, Revision 0

OX1456.86, Operability Testing of IST Pumps, Revision 15

Condition Reports

2227744

Maintenance Orders/Work Orders

40508512 40515051 40561271

Drawings

PID-1-CC-B20205, Revision 27

Section 1EP6: Drill Evaluation

Procedures

ER 1.1, Classification of Emergencies, Revision 58

ER 3.1, Technical Support Center Operations, Revision 64

EP-AA-100-1000, Conduct of Emergency Preparedness, Revision 6

EP-AA-101-1000, Nuclear Division Drill and Exercise Procedure, Revision 20

Condition Reports

2223189 2229621 2232420

Miscellaneous

CFD 17-03 Drill Scenario

Combined Functional Drill Report, CFD-17-03, dated October 11, 2017

NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6

Training Lesson Plan, E01090I, Emergency Classifications, Revision 6

Training Lesson Plan, L1809C, Nuclear Energy Institute (NEI), Emergency Action Levels,

Revision 6

A-6

Section 2RS2: Occupational ALARA Planning and Controls

Procedures

RP-AA-104-1000, ALARA Implementing Procedure, Revision 13

Condition Reports

02173460 02198478 02198480 02198735 02199920 02215940

02220919 02221333 02223171

Miscellaneous

2017 Department Exposure Goals and Year to Date Department Exposures, December 5, 2017

2017 Routine Operating Dose Report, December 3, 2017

ALARA Dose Estimate Report for Work Week 1749 (December 3-8, 2017), December 4, 2017

ALARA Review Board Meeting 17-04 Subcommittee, September 19, 2017

ALARA Review Board Meeting 17-03 Subcommittee, August 14, 2017

ALARA Review Board Meeting 17-02 Subcommittee, July 26, 2017

Level 1 Assessment for NRC ALARA RETS, REMP Inspections, AR 2233688, October 30, 2017

Post-Job ALARA Review No. 17-0031, Dry Fuel Storage Project Activities, December 6, 2017

Post-Job ALARA Review No. 17-0140, OR 18 Scaffolding, December 6, 2017

Post-Job ALARA Review No.17-002, Steam Generator Primary Eddy Current Test,

December 6, 2017

Post-Job ALARA Review No.17-001, Reactor Vessel Disassembly and Reassembly,

December 6, 2017

Section 2RS3: In-Plant Airborne Radioactivity Controls and Mitigation

Procedures

HD0965.01, Respiratory Protection Quality Assurance and Maintenance Program, Revision 22

HD0965.02, Repair, Inspection, Inventory and Maintenance of Respiratory Protection

Equipment, Revision 27

HD0965.08, Breathing Air Certification, Revision 17

HD0965.10, Respirator Fit Testing Using the TSI Portacount, Revision 19

HD0965.12, Respiratory Equipment Issue and Use, Revision 42

RP-AA-106, Respiratory Protection Program, Revision 0

Condition Reports

02122162 02149186 02168471 02178320

Miscellaneous

Annual Assessment of the 2016 Respiratory Protection Program, AR 2206817, June 7, 2017

FireHawk M7 SCBA Use: Inspection and Donning Instructions, Operator Aide, Revision 9

Fit Test Report for MSA Ultra Elite 1000 (medium) using Portacount # 8030142409,

December 7, 2017

Fit Test Report for MSA Ultra Rubber (medium) using Portacount # 8030142409,

December 7, 2017

HD0965.02, Figure 2: SCBA Inventory, November 30, 2017

HD0965.02, HRE-M1 SCBA Inspection and Inventory, November 30, 2017

HD0965.14, Form B: SCBA Face Piece Test for Ultra Elite 1000 (medium) FH-022,

December 4, 2017

HD0965.14, Form B: SCBA Face Piece Test for Ultra Elite 1000 (medium) FH-037,

September 7, 2017

HD0965.14, Form A: SCBA/ PremAire Cadet Regulator Flow Test ANAD063768,

December 4, 2017

A-7

HD0965.14, Form A: SCBA/ PremAire Cadet Regulator Flow Test APAB279701,

August 11, 2017

Honeywell Certificate of Calibration No. 56041717L02497 Serial No. L02497, April 1, 2017

MSA CARE Authorized Repair Center and MSA MMR Certified CARE Technician Certification,

March 3, 2015

Posi3 USB Test Results Serial No. L02497 for MSA Ultra Elite (medium) FH-022,

December 4, 2017

Posi3 USB Test Results Serial No. L02497 for MSA FireHawk M7 Air Mask (medium) PR 14,

December 4, 2017

SBK HPT HP0090J, RP Technician Respirator Training, June 2, 2014

SBK GET GT1074J, Firehawk M7 SCBA Training, July 11, 2013

Service History for Instrument Model SCBA Regulator (including maintenance/repair notes),

December 6, 2017

TRI Air Testing, Inc Laboratory Report Compressed Air/Gas Quality Test for Firefighting Annex

Breathing Air, September 15, 2017

TRI Air Testing, Inc Laboratory Report Compressed Air/Gas Quality Test for Firefighting Annex

Breathing Air, June 15, 2017

TRI Air Testing, Inc Laboratory Report Compressed Air/Gas Quality Test for Firefighting Annex

Breathing Air, June 23, 2017

TSI Certificate of Testing PortaCount 8030 Bench 2 Serial No. 8030134708,

September 20, 2017

TSI Certificate of Testing PortaCount 8030 Bench 2 Serial No. 8030134713, July 12, 2017

Section 4OA1: Performance Indicator Verification

Procedures

CS0917.02, Gaseous Effluent Releases, Revision 14

CX0917.01, Liquid Effluent Releases, Revision 20

HD0958.33, Performance of Radiation Protection Supervisory Plant Walkdowns, Revision 6

JD0999.910, Reporting Key Performance Indicators per NEI 99-02, Revision 8

Condition Reports

02093824 02162340 02195218

Miscellaneous

CP 4.1C, Release Index Log 2016, November 6, 2017

CX0917.01, Form C: LEW Release Data, Permit # 17-448, Waste Test Tank B,

October 29, 2017

CX0917.01, Form C: LEW Release Data, Permit # 17-458, Storm Drain/Groundwater Extraction

Wells, October 31, 2017

CX0917.01, Form C: LEW Release Data, Permit # 17-462, Steam Generator Blowdown Drain

Flash Tank, November 8, 2017

CX0917.01, GEW Sample Collection Data, Permit # 17-451, Plant Vent, October 31, 2017

JD0999.910, Figure 1, Occupational Exposure Occurrence, January, February and March 2017,

dated April 25, 2017

JD0999.910, Figure 1, Occupational Exposure Occurrence, April, May and June 2017,

dated July 7, 2017

JD0999.910 Figure 1 Occupational Exposure Occurrence, July, August and September 2017,

dated October 27, 2017

LIC-17010, Seabrook Station NRC Third Quarter 2017 Performance Indicator Submittal

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

NextEra - Seabrook Station 2016 Annual Radioactive Release Report, April 28, 2017

A-8

MSPI Derivation Reports for MSPI Systems Residual Heat Removal System and Cooling Water

System, November 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

November 2017, December 1, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

October 2017, November 1, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

September 2017, October 2, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

August 2017, September 5, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

July 2017, August 1, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

June 2017, July 5, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

May 2017, June 2, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

April 2017, May 2, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

March 2017, April 2, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

February 2017, March 1, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

January 2017, February 1, 2017

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

December 2016, January 3, 2016

Reactor Coolant Specific Activity and RETS ODCM Radiological Effluent Occurrence KPIs,

November 2016, December 1, 2017

SBK-PRAE-15-001

Section 4OA2: Problem Identification and Resolution

Procedures

ER-AA-101, Equipment Reliability, Revision 7

ER-AA-201-2001, System Health Reporting, Revision 12

ER-AA-201-2002, System Performance Monitoring, Revision 4

OP-AA-108-1000, Operator Challenges Program Management, Revision 5

PI-AA-207-1000, Station Self-Evaluation and Trending Analysis, Revision 8

PI-AA-207, Trend Coding and Analysis, Revision 12

PI-AA-101, Assessment and Improvement Programs, Revision 23

SMPM, Structures Monitoring Program Manual, Revisions 2 and 3

Condition Reports

1637922 2053980 2144822 2151482 2153374 2157499

2162430 2162696 2162696 2164268 2164482 2168700

2175840 2178962 2178962 2181193 2205604 2207649

2214502 2215560 2215959 2216230 2216936 2217146

2217211 2219903 2222763 2222809 2223576 2224985

2227328 2232578 2235442 2236473 2237328 2237940

2238405 2111108 2148021 2240426

A-9

Maintenance Orders/Work Orders

01209317 01209321 40176613 40260904 40395367 40531735

40538714 40540846 40568543

Miscellaneous

160268-CA-05, Susceptibility Evaluation of Containment Enclosure Ventilation Area, Revision 0,

dated March 22, 2017

170400-SVR-04-RA, 2017 Tier 2 Inspections - ASR Inspections and Cracking Index

Measurements on Concrete Structures, dated October 10, 2017

170400-SVR-05-RA, 2017 Tier 2 Inspections - Measurements for ASR Expansion on Concrete

Surfaces, dated October 10, 2017

Evaluation - North Wall of Containment Enclosure Ventilation Area (CEVA) Near-Surface

Delamination (Cover Concrete Separation), dated October 30, 2017

FP 101034, Petrographic Examinations of Equipment Vaults, Revision 1

FP 101044, Identify and Measure Seismic Gaps Between the CEB and CB at 4 Missile Shields,

Revision 0

FP 101055, Condition Assessment of Cracking in RHR and CS Equipment Vault - Second Visit,

Revision 0

PEG-94, Revision 11

Prompt Operability Determination (POD) for AR 01664399, Consolidation of PODs for Reduced

Concrete Properties in Alkali Silica Reaction (ASR) Affected Seismic Category I

Structures, Revision 2, dated October 6, 2017

POD AR 02014325, Consolidation of Building Deformation Prompt Operability Determinations,

Revision 1, dated October 6, 2017

POD for AR 02193235, Alkali Silica Reaction (ASR) effects on CEVA Structure North Wall,

Revision 1, dated September 28, 2017

POD for AR 02215578, Evaluation of B Electrical Cable Tunnel as an Alkali Silica Reaction

(ASR) Affected Seismic Category I Structure, Revision 2, dated July 19, 2017

Drawings

9763-F-101620, Sheet 1, Containment Enclosure Ventilation Area Concrete Sections,

Revision 5

9763-F-113230, Sheet 1, Schedule of Required Backfill Concrete and Isolation Material for

Structures, Revision 5

Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion

Procedures

MA 4.5, Configuration Control, Revision 18

MA-AA-100, Conduct of Maintenance, Revision 16

MA-AA-203-1001, Work Order Planning, Revision 8

OP-AA-100-1000, Conduct of Operations, Revision 20

Condition Reports

2202358

Maintenance Orders/Work Orders

40532423

A-10

Miscellaneous

LER 2017-001-00, Manual Reactor Trip in Response to a Feedwater Isolation due to High Level

in Steam Generator B, June 27, 2017

Manual Reactor Trip in Response to a Feedwater Isolation due to High Level in Steam

Generator B, Event Date: 4/29/17, Root Cause Evaluation

P-14 Event Analysis

LIST OF ACRONYMS

ACI American Concrete Institute

ADAMS Agencywide Documents Access and Management System

ALARA As Low As is Reasonably Achievable

ASR alkali silica reaction

CAP corrective action program

CEVA containment enclosure ventilation area

CFR Code of Federal Regulations

CR condition report

CS containment spray

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

JPM job performance measure

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

RHR residual heat removal

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

SW service water

TS technical specification

TSC technical support center

UFSAR September 22, 2015, because the finding Updated Final Safety

Analysis Report

WO work order

WRGM wide range gas monitor