K. Walsh 

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

REGION I 

2100 RENAISSANCE BLVD., SUITE 100 

KING OF PRUSSIA, PA 19406-2713 

May 6, 2014

Mr. Kevin Walsh 

Site Vice President 

Seabrook Nuclear Power Plant 

NextEra Energy Seabrook, LLC 

c/o Mr. Michael Ossing 

P.O. Box 300 

Seabrook, NH 03874 

 

SUBJECT: SEABROOK STATION, UNIT NO. 1 - NRC INTEGRATED INSPECTION 

REPORT 05000443/2014002 

Dear Mr. Walsh: 

On March 31, 2014, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection 

at Seabrook Station, Unit No. 1. The enclosed inspection report documents the inspection 

results, which were discussed on April 10, 2014, with you and other members of your staff. 

The inspection examined activities conducted under your license as they relate to safety and 

compliance with the Commission’s rules and regulations and with the conditions of your license. 

The inspectors reviewed selected procedures and records, observed activities, and interviewed 

personnel. 

This report documents one NRC-identified finding of very low safety significance (Green). This 

finding was determined to involve a violation of NRC requirements. However, because of the 

very low safety significance, and because it was entered into your corrective action program 

(CAP), the NRC is treating the finding as a non-cited violation (NCV), consistent with Section 

2.3.2.a of the NRC Enforcement Policy. If you contest the subject or severity of any NCV in this 

report, you should provide a response within 30 days of the date of this inspection report, with 

the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control 

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

Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, 

DC 20555-0001; and the NRC Resident Inspector at Seabrook Station. In addition, if you 

disagree with the cross-cutting aspect assigned to the finding in this report, you should provide 

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

disagreement, to the Regional Administrator, Region I, and the NRC Resident Inspector at 

Seabrook Station. 

Additionally, as we informed you in the most recent NRC integrated inspection report, crosscutting aspects identified in the last six months of 2013 using the previous terminology were 

being converted in accordance with the cross-reference in Inspection Manual Chapter (IMC) 

0310. Section 4OA5 of the enclosed report documents the conversion of these cross-cutting 
K. Walsh 2 

aspects which will be evaluated for cross-cutting themes and potential substantive cross-cutting 

issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment review. If you 

disagree with the cross-cutting aspect assigned, you should provide a response within 30 days 

of the date of this inspection report, with the basis for your disagreement, to the Regional 

Administrator, Region I, and the NRC Resident Inspector at Seabrook Station. 

In accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390 of the NRCs “Rules 

of Practice,” a copy of this letter, its enclosure, and your response (if any) will be available 

electronically for public inspection in the NRC’s Public Document Room or from the Publicly 

Available Records component of the NRC’s Agencywide Documents Access Management 

System (ADAMS). ADAMS is accessible from the NRC website at http://www.nrc.gov/readingrm/adams.html (the Public Electronic Reading Room). 

Sincerely, 

 /RA/ 

 Glenn T. Dentel, Chief 

Reactor Projects Branch 3

Division of Reactor Projects 

Docket No. 50-443 

License No. NPF-86 

Enclosure: Inspection Report No. 05000443/2014002 

 w/ Attachment: Supplemental Information 

cc w/encl: Distribution via ListServ 
K. Walsh 2 

aspects which will be evaluated for cross-cutting themes and potential substantive cross-cutting 

issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment review. If you 

disagree with the cross-cutting aspect assigned, you should provide a response within 30 days 

of the date of this inspection report, with the basis for your disagreement, to the Regional 

Administrator, Region I, and the NRC Resident Inspector at Seabrook Station. 

In accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390 of the NRCs “Rules 

of Practice,” a copy of this letter, its enclosure, and your response (if any) will be available 

electronically for public inspection in the NRC’s Public Document Room or from the Publicly 

Available Records component of the NRC’s Agencywide Documents Access Management 

System (ADAMS). ADAMS is accessible from the NRC website at http://www.nrc.gov/readingrm/adams.html (the Public Electronic Reading Room). 

Sincerely, 

 /RA/ 

 Glenn T. Dentel, Chief 

Reactor Projects Branch 3

Division of Reactor Projects 

Docket No. 50-443 

License No. NPF-86 

Enclosure: Inspection Report No. 05000443/2014002 

 w/ Attachment: Supplemental Information 

cc w/encl: Distribution via ListServ 

Distribution w/encl: 

W. Dean, RA 

D. Lew, DRA 

H. Nieh, DRP 

M. Scott, DRP 

R. Lorson, DRS 

J. Trapp, DRS 

G. Dentel, DRP 

R. Barkley, DRP 

M. Draxton, DRP 

B. Reyes, DRP 

P. Cataldo SRI, DRP 

C. Newport, RI, DRP 

A. Cass, DRP, Resident AA 

J. Nick, RI OEDO 

RidsNrrPMSeabrook Res 

RidsNrrDorlLpl1-2 Res 

ROPreports Resource

DOCUMENT NAME: G:\DRP\BRANCH3\Inspection\Reports\Issued\2014 (ROP 15)\SB IR14002 final.docx 

ADAMS Accession No.: ML14127A376 

 SUNSI Review 

 Non-Sensitive 

 Sensitive 

 Publicly Available 

 Non-Publicly Available 

OFFICE RI/DRP RI/DRP RI/DRP 

NAME PCataldo/GTD for per 

email 

RBarkley/GTD for per 

email GDentel/GTD 

DATE 05/05/14 05/06/14 05/06/14 

OFFICIAL RECORD COPY
1 

Enclosure 

U.S. NUCLEAR REGULATORY COMMISSION 

REGION I 

Docket No.: 50-443 

License No.: NPF-86 

Report No.: 05000443/2014002 

Licensee: NextEra Energy Seabrook, LLC 

Facility: Seabrook Station, Unit No.1 

Location: Seabrook, New Hampshire 03874 

Dates: January 1, 2014 through March 31, 2014 

Inspectors: P. Cataldo, Senior Resident Inspector 

C. Newport, Resident Inspector 

 E. Burket, Emergency Preparedness Inspector 

 T. Burns, Reactor Inspector 

B. Dionne, Health Physicist 

W. Cook, Senior Reactor Analyst 

Approved by: Glenn T. Dentel, Chief 

 Reactor Projects Branch 3 

 Division of Reactor Projects 

 
2 

Enclosure 

TABLE OF CONTENTS 

SUMMARY .................................................................................................................................... 3

REPORT DETAILS ....................................................................................................................... 4

1. REACTOR SAFETY .............................................................................................................. 4

1R01 Adverse Weather Protection ....................................................................................... 4

1R04 Equipment Alignment .................................................................................................. 4

1R05 Fire Protection ............................................................................................................. 5

1R06 Flood Protection Measures ........................................................................................ 6

1R11 Licensed Operator Requalification Program ............................................................... 6

1R12 Maintenance Effectiveness ......................................................................................... 7

1R13 Maintenance Risk Assessments and Emergent Work Control .................................. 7

1R15 Operability Determinations and Functionality Assessments ....................................... 8

1R18 Plant Modifications ................................................................................................... 10

1R19 Post-Maintenance Testing ....................................................................................... 11

1R22 Surveillance Testing ................................................................................................. 11

1EP4 Emergency Action Level and Emergency Plan Changes ......................................... 12

1EP6 Drill Evaluation ......................................................................................................... 12

2. RADIATION SAFETY .......................................................................................................... 13

2RS1 Radiological Hazard Assessment and Exposure Controls ....................................... 13

2RS2 Occupational ALARA Planning and Controls ........................................................... 16

4. OTHER ACTIVITIES ............................................................................................................ 17

4OA1 Performance Indicator Verification ........................................................................... 17

4OA2 Problem Identification and Resolution ..................................................................... 18

4OA3 Follow-Up of Events and Notices of Enforcement Discretion ................................... 22

4OA5 Other Activities .......................................................................................................... 23

4OA6 Meetings, Including Exit ............................................................................................ 24

ATTACHMENT: SUPPLEMENTARY INFORMATION................................................................ 24

SUPPLEMENTARY INFORMATION ........................................................................................ A-1

KEY POINTS OF CONTACT .................................................................................................... A-1

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED ..................................... A-1

LIST OF DOCUMENTS REVIEWED ........................................................................................ A-1

LIST OF ACRONYMS ............................................................................................................. A-10

 
3 

Enclosure 

SUMMARY 

IR 05000443/2014002; 01/01/2014-03/31/2014; Seabrook Station, Unit No. 1; Operability 

Determinations and Functionality Assessments. 

This report covered a three-month period of inspection by resident inspectors and announced 

inspections performed by regional inspectors. Inspectors identified one finding of very low 

safety significance (Green), which was an NCV. The significance of most findings is indicated 

by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using IMC 

0609, “Significance Determination Process” (SDP), dated June 2, 2011. Cross-cutting aspects 

are determined using IMC 0310, “Components Within Cross-Cutting Areas,” dated December 

19, 2013. All violations of NRC requirements are dispositioned in accordance with the NRC’s 

Enforcement Policy, dated June 7, 2012. The NRC’s program for overseeing the safe operation 

of commercial nuclear power reactors is described in NUREG-1649, “Reactor Oversight 

Process,” Revision 4. 

Cornerstone: Mitigating Systems 

 Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V, 

“Procedures,” because NextEra did not ensure adequate separation was maintained 

between temporary scaffolding and safety-related equipment. Specifically, six instances 

of scaffolding installed in the plant were identified with less than the minimum standoff 

distance to safety-related equipment specified in NextEra procedures and no corresponding 

engineering evaluation to support these deviations. NextEra entered this NCV into their 

CAP as AR 01933827 and assessed the six deviations for any impact on the associated 

safety-related systems. 

This performance deficiency was considered more than minor because it affected the 

protection against external factors attribute of the Mitigating Systems cornerstone and its 

objective to ensure the availability, reliability, and capability of systems that respond to 

initiating events to prevent undesirable consequences. Specifically, NextEra did not 

evaluate scaffolding installations when insufficient separation to safety-related equipment 

existed after procedural requirements were revised to a more restrictive value. Additionally, 

it was similar to example 4.a in IMC 0612, Appendix E, “Examples of Minor Issues,” which 

states that the issue of failing to appropriately evaluate scaffold installation as required by 

procedures is more than minor if the licensee routinely failed to perform engineering 

evaluations. The issue was evaluated in accordance with IMC 0609, Appendix A, “The 

Significance Determination Process for Findings At-Power” and determined to be of very low 

safety significance (Green), because it did not involve the loss or degradation of equipment 

or function specifically designed to mitigate a seismic event. This finding has a cross-cutting 

aspect in the area of Problem Identification and Resolution, Evaluation, because NextEra 

personnel did not perform an adequate extent of condition review after revision of their 

erection of scaffold procedure. This performance deficiency directly contributed to multiple 

instances of scaffold members erected within two inches of safety-related equipment without 

an engineering evaluation [P.2]. (Section 1R15)

 
4 

Enclosure 

REPORT DETAILS 

Summary of Plant Status 

Seabrook operated essentially at full power for the entire assessment period, with the exception 

of minor downpowers for turbine control valve testing. However, on March 31, 2014, plant load 

was reduced to approximately 15% for turbine generator testing prior to a shutdown and entry 

into refueling outage No. 16 at midnight, March 31, 2014. 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 sample) 

 Readiness for Impending Adverse Weather Conditions 

a. Inspection Scope 

The inspectors reviewed NextEra’s preparations for the onset of cold weather and snow 

on February 5, 2014. The inspectors reviewed the implementation of adverse weather 

preparation procedures before the onset of and during this adverse weather 

condition. The inspectors verified that operator actions defined in NextEra’s adverse 

weather procedure maintained the readiness of essential systems. The inspectors 

discussed readiness and staff availability for adverse weather response with operations 

and work control personnel. 

b. Findings 

 No findings were identified. 

1R04 Equipment Alignment 

 Partial System Walkdowns (71111.04Q – 3 samples) 

a. Inspection Scope 

The inspectors performed partial walkdowns of the following systems: 

 'A' emergency diesel generator (EDG) while 'B' EDG was out of service (OOS) for 

annual maintenance on January 27, 2014 

 Supplemental emergency power system (SEPS) while 'A' EDG was OOS for annual 

maintenance on February 12, 2014 

 'A' emergency feedwater (EFW) pump return to service on March 26, 2014 

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

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

applicable operating procedures, system diagrams, the Updated Final Safety Analysis 
5 

Enclosure 

Report (UFSAR), technical specifications (TSs), work orders (WOs), condition reports, 

and the impact of ongoing work activities on redundant trains of equipment in order to 

identify conditions that could have impacted system performance of their intended safety 

functions. The inspectors also performed field walkdowns of accessible portions of the 

systems to verify system components and support equipment were aligned correctly and 

were operable. The inspectors examined the material condition of the components and 

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

The 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. 

1R05 Fire Protection 

 Resident Inspector Quarterly Walkdowns (71111.05Q – 5 samples) 

a. Inspection Scope 

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

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

that NextEra controlled combustible materials and ignition sources in accordance with 

administrative procedures. The inspectors verified that fire protection and suppression 

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

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

station personnel implemented compensatory measures for OOS, degraded, or 

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

 Residual heat removal (RHR) containment spray safety injection (CSSI) equipment 

vault train 'A' RHR-F-1B-Z, RHR-F-2B-Z, RHR-3B-Z, RHR-F-4B-Z, RHR-F-4B-Z1, 

RHR-F-4B-Z2 on January 15, 2014 

 RHR CSSI equipment vault train 'B' RHR-F-1A-Z, RHR-F-2A-Z, RHR-3A-Z, 

RHR-F-4A-Z, RHR-F-4A-Z1, RHR-F-4A-Z2 on January 16, 2014 

 SEPS-F-1-0 on January 21, 2014 

 Circulating water pump room SW-F-1A-Z on March 11, 2014 

 Primary auxiliary building (PAB) piping penetration area PAB-F-1A-Z, PAB-F-1J-Z 

on March 17, 2014 

b. Findings 

No findings were identified. 
6 

Enclosure 

1R06 Flood Protection Measures (71111.06 – 1 sample) 

 Internal Flooding Review 

a. Inspection Scope 

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

assess susceptibilities involving internal flooding. The inspectors also reviewed the CAP 

to determine if NextEra identified and corrected flooding problems and whether operator 

actions for coping with flooding were adequate. The inspectors also focused on the ‘B’ 

EDG building to verify the adequacy of equipment seals located below the flood line, 

floor and water penetration seals, watertight door seals, common drain lines and sumps, 

sump pumps, level alarms, control circuits, and temporary or removable flood barriers. 

b. Findings 

No findings were identified. 

1R11 Licensed Operator Requalification Program (71111.11 – 2 samples) 

.1 Quarterly Review of Licensed Operator Requalification Testing and Training 

a. Inspection Scope 

The inspectors observed licensed operator simulator training on January 23, 2014, 

which included simulated degraded equipment and subsequent equipment failures 

and initiators, which resulted in escalating degraded plant conditions that ensured 

implementation of emergency operating procedures by the operating crew, as well as 

implementation of the emergency plan. This emergency plan implementation included 

classification of specific events that warranted an Alert Event Declaration. The 

inspectors evaluated operator performance during the simulated event and verified 

completion of risk significant operator actions, including the use of abnormal and 

emergency operating procedures. The inspectors assessed the clarity and effectiveness 

of communications, implementation of actions in response to alarms and degrading plant 

conditions, and the oversight and direction provided by the control room supervisor. The 

inspectors verified the accuracy and timeliness of the emergency classification made by 

the shift manager and the TS action statements entered by the 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 

a. Inspection Scope 

The inspectors observed general control room activities, including alarm response and 

control room shift turnovers, conducted on January 13, 2014, March 14, 2014 and 
7 

Enclosure 

March 27, 2014. Additionally the inspectors observed turbine control valve testing on 

January 17, 2014, engineered safety features actuation system (ESFAS) relay testing 

on January 27, 2014, operator response to a failed open ’B’ steam generator feed 

regulating bypass valve on March 14, 2014, and restoration from enclosure building 

exhaust fan EAH-FN-4A testing on March 19, 2014. The inspectors observed test 

performance to verify that procedure use, crew communications, and coordination of 

activities between work groups similarly met established expectations and standards. 

b. Findings 

No findings were identified. 

1R12 Maintenance Effectiveness (71111.12 – 2 samples) 

a. Inspection Scope 

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

maintenance activities on structure, system, or 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. 

 SW pump P-41C increased vibration trending in January 2014 

 ED/EDE 120 VAC electrical distribution systems on February 18, 2014 

b. Findings 

No findings were identified. 

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 – 5 samples) 

a. Inspection Scope 

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

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

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

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

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

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

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

the inspectors verified that operations personnel promptly assessed and managed plant 

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

of the assessment with the station’s probabilistic risk analyst to verify plant conditions 
8 

Enclosure 

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. 

 RHR system valve maintenance and testing on January 14, 2014 

 ESFAS relay testing on January 15, 2014 

 Inverter 1B corrective maintenance following internal transformer failure on 

February 19, 2014 

 Planned SW cooling tower switchover on March 20, 2014 

 'B' feedwater regulating bypass valve M/A station replacement on March 27, 2014 

b. Findings 

No findings were identified. 

1R15 Operability Determinations and Functionality Assessments (71111.15 – 7 samples) 

a. Inspection Scope 

The inspectors reviewed operability determinations for the following degraded or nonconforming conditions: 

 East and west pipe chase low temperature impact on feedwater isolation valve 

operability on January 3, 2014 

 Operability of safety-related equipment in close proximity to temporary scaffolding on 

January 16, 2014 

 Service water pumphouse seismic monitor non-functional following monthly testing 

on January 24, 2014 

 Containment enclosure ventilation area seal gaps identified on January 28, 2014 

 'A' EDG did not trip on overspeed during return to service testing on February 14, 2014 

 Reactor coolant system (RCS) leakage into the RHR system on March 13, 2014 

 Turbine-driven EFW pump P-37B oil leak identified during testing on March 19, 2014 

The inspectors selected these issues based on the risk significance of the associated 

components and systems. The inspectors evaluated the technical adequacy of the 

operability determinations to assess whether TS operability was properly justified and 

the subject component or system remained available such that no unrecognized 

increase in risk occurred. The inspectors compared the operability and design criteria in 

the appropriate sections of the TSs and UFSAR to NextEra’s evaluations to determine 

whether the components or systems were operable. 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. The 

inspectors determined, where appropriate, compliance with bounding limitations 

associated with the evaluations. 

b. Findings 

Introduction. The inspectors identified a Green NCV of 10 CFR 50, Appendix B, 

Criterion V, “Procedures,” because NextEra did not ensure adequate separation was 

maintained between temporary scaffolding and safety-related equipment. Specifically, 
9 

Enclosure 

six instances of scaffolding installed in the plant were identified with less than the 

minimum standoff distance to safety-related equipment specified in NextEra procedures 

and no corresponding engineering evaluation to support these deviations. 

Description. 10 CFR 50, Appendix B, Criterion V, requires that activities affecting quality 

be prescribed by documented procedures and be accomplished in accordance with 

those procedures. When used in the plant, the design and installation of temporary 

scaffolding must be controlled to ensure that it is not installed too close to safety-related 

equipment. During a seismic event, scaffolding installed too close to safety-related 

equipment can come into contact with that equipment, cause damage to it, and affect its 

safety function. NextEra procedures control the installation of temporary scaffolding at 

Seabrook by specifying a minimum separation between scaffolding and safety-related 

equipment, and by requiring an engineering evaluation when the minimum separation 

cannot be met. 

NextEra mechanical maintenance procedure, MS0599.47, “Erection of Scaffolding,” 

Revision 2, states that members of scaffolding erected adjacent to operable safetyrelated equipment shall not be less than two inches unless justified by an Engineering 

Evaluation. MS0599.47 was revised in February 2013, and the requirement for scaffold 

separation from operable safety-related equipment was changed from “...should not be 

less than 2 inches and in no case less than 1⁄2 inch without an engineering evaluation” to 

“...shall not be less than 2 inches unless justified by engineering evaluation.” 

While performing a plant walkdown on January 15, 2014, the inspectors identified 

temporary scaffold members installed less than two inches from the ‘A’ Containment 

Building Spray (CBS) pump discharge and suction lines. The ‘A’ CBS pump and its 

associated piping are classified as safety-related equipment and were operable at the 

time. The identified scaffold did not include an engineering evaluation that provided 

acceptance of separation of less than two inches. Subsequent plant walkdowns by 

NextEra personnel identified five additional instances of scaffolding installed less than 

two inches from operable safety-related equipment without an associated engineering 

evaluation. NextEra personnel determined that an inadequate extent of condition review 

following the February 2013 revision of MS0599.47, resulted in scaffolding being staged 

in the plant at less than the new, more restrictive scaffold separation requirement of two 

inches. Having identified multiple instances where NextEra personnel had not complied 

with the separation requirement of the scaffolding procedure, the inspectors concluded 

that NextEra had not been adequately controlling the design and installation of 

temporary scaffolding. 

NextEra entered the additional instances of inadequate separation identified during their 

independent walkdowns into the CAP. All discrepancies were corrected and assessed 

for any potential impact to the operability or functionality of affected systems. The 

inspectors reviewed the CRs and determined that the safety function of each system 

potentially impacted by temporary scaffolding, including those identified by the 

inspectors and NextEra, would not have been affected during a seismic event. 

Analysis. The inspectors determined that not providing adequate separation between 

temporary scaffolding and safety-related equipment without an engineering basis was 

a performance deficiency within NextEra’s ability to foresee and correct. Specifically, 

several scaffold members were observed within two inches of safety-related equipment 

without an engineering evaluation as specified by current procedural requirements. 
10 

Enclosure 

This performance deficiency was considered more than minor because it affected the 

protection against external factors attribute of the Mitigating System cornerstone and its 

objective to ensure the availability, reliability, and capability of systems that respond to 

initiating events to prevent undesirable consequences. Specifically, NextEra did not 

evaluate scaffolding installation when insufficient separation to safety-related equipment 

existed after procedural requirements were revised to a more restrictive value. 

Additionally, it was similar to example 4.a in IMC 0612, Appendix E, “Examples of Minor 

Issues,” which states that the issue of failing to appropriately evaluate scaffold 

installation as required by procedures is more than minor if the licensee routinely failed 

to perform engineering evaluations. The issue was evaluated in accordance with IMC 

0609, Appendix A, “The Significance Determination Process for Findings At-Power” and 

determined to be of very low safety significance (Green) since it did not involve the loss 

or degradation of equipment or function specifically designed to mitigate a seismic event. 

This finding is related to the cross-cutting area of Problem Identification and ResolutionEvaluation, because NextEra did not thoroughly evaluate issues to ensure that 

resolutions address causes and extent of conditions commensurate with their safety 

significance (P.2). Specifically, NextEra personnel did not perform an adequate extent 

of condition review after revision of their erection of scaffolding procedure. This 

performance deficiency directly contributed to multiple instances of scaffolding members 

erected within two inches of safety-related equipment without an engineering evaluation. 

Enforcement. 10 CFR 50, Appendix B, Criterion V, requires, in part, that activities 

affecting quality shall be prescribed by documented procedures and shall be 

accomplished in accordance with those procedures. NextEra mechanical maintenance 

procedure, MS0599.47, “Erection of Scaffolding,” Revision 2, states that members of 

scaffolding erected adjacent to operable safety-related equipment shall not be less than 

two inches from the equipment unless justified by an Engineering Evaluation. Contrary 

to the above, on January 15, 2014, the inspectors identified that certain activities 

affecting quality at Seabrook were not accomplished in accordance with documented 

procedures. Specifically, following a revision of the minimum scaffolding separation 

requirement in February 2013, multiple instances of scaffolding outside of the new 

requirements were left uncorrected and engineering evaluations were not completed. 

Installation of temporary scaffolding in the vicinity of safety-related equipment has the 

potential to adversely affect that equipment’s performance during a seismic event 

because it was installed with insufficient standoff distance. After the issue was identified 

by the inspectors, NextEra performed independent walkdowns of all scaffolding, 

identifying five additional instances of inadequate separation distance. All identified 

discrepancies were corrected or evaluated as adequate. Because this violation is of 

very low safety significance (Green) and NextEra entered this into their CAP (AR 

01933827), this violation is being treated as an NCV consistent with the Enforcement 

Policy. (NCV 05000443/2014002-01, Scaffolding Installed with Insufficient 

Separation to Safety Related Equipment)

1R18 Plant Modifications (71111.18 – 1 sample) 

 Permanent Modifications 

a. Inspection Scope 

The inspectors evaluated a modification that replaced the SW flow element to the ‘A’ 

EDG jacket water cooler implemented under engineering change EC280824, and 
11 

Enclosure 

completed on February 13, 2014. The inspectors verified that the design and licensing 

bases, as well as the performance capability of the affected SW train and associated 

components were not degraded by the modification. The inspectors reviewed 

associated modification documents, which included topic notes, implementation work 

order instructions, equivalent change revisions, applicable interface documents (for 

example, drawings), and applicable post-modification testing. 

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 ensured system operability and 

functional capability. The inspectors reviewed the test procedure to verify that the 

procedure adequately tested the safety functions that may have been affected by the 

maintenance activity, that the acceptance criteria in the procedure was consistent with 

the information in the applicable licensing basis and/or design basis documents, and that 

the procedure had been properly reviewed and approved. The inspectors also 

witnessed the test or reviewed test data to verify that the test results adequately 

demonstrated restoration of the affected safety functions. 

 Thermal barrier cooling pump monthly surveillance following electrical breaker 

testing on January 11, 2014 

 SEPS diesel generator DG-2A maintenance on January 21, 2014 

 Portable diesel driven pump B.5.b functional test following repairs on February 3, 2014 

 EDG ‘A’ exhaust valve Belleville washer replacement on February 11, 2014 

 1B vital inverter Ferro-Resonant transformer replacement on February 19, 2014 

 Service water pump P-41C following shaft sleeve replacement on March 6, 2014 

b. Findings 

No findings were identified. 

1R22 Surveillance Testing (71111.22 – 7 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: 
12 

Enclosure 

 Rod control testing and verification of proper operation of digital rod position 

indication on January 9, 2014 

 Portable diesel driven pump B.5.b annual functional test on January 14, 2014 

 'A' cooling tower pump comprehensive inservice test on January 16, 2014 (IST) 

 RCS leak rate surveillance test on March 5, 2014 (RCS leak rate) 

 RCS pump seal monthly controlled leakage surveillance on March 12, 2014 

 Containment enclosure exhaust fan EAH-FN-4A monthly testing on March 19, 2014 

 'B' EDG operability surveillance and ESFAS slave relay testing March 25, 2014 

b. Findings 

No findings were identified. 

Cornerstone: Emergency Preparedness 

1EP4 Emergency Action Level and Emergency Plan Changes (71114.04 – 1 sample) 

 Emergency Preparedness Drill Observation 

a. Inspection Scope 

NextEra implemented various changes to the Seabrook Emergency Action Levels 

(EALs), Emergency Plan, and Implementing Procedures. NextEra had determined that, 

in accordance with 10 CFR 50.54(q)(3), any change made to the EALs, Emergency 

Plan, and its lower-tier implementing procedures, had not resulted in any reduction in 

effectiveness of the Plan, and that the revised Plan continued to meet the standards in 

50.47(b) and the requirements of 10 CFR 50, Appendix E. 

The inspectors performed an in-office review of all EAL and Emergency Plan changes 

submitted by NextEra as required by 10 CFR 50.54(q)(5), including the changes to 

lower-tier emergency plan implementing procedures, to evaluate for any potential 

reductions in effectiveness of the Emergency Plan. This review by the inspectors was 

not documented in an NRC Safety Evaluation Report and does not constitute formal 

NRC approval of the changes. Therefore, these changes remain subject to future NRC 

inspection in their entirety. The requirements in 10 CFR 50.54(q) were used as 

reference criteria. 

b. Findings 

No findings were identified. 

1EP6 Drill Evaluation (71114.06 – 2 samples) 

.1 Emergency Preparedness Drill Observation 

a. Inspection Scope 

The inspectors evaluated a routine NextEra emergency drill on March 12, 2014, to 

identify any weaknesses and deficiencies in the event classification and notification 
13 

Enclosure 

activities. The inspectors observed emergency response operations in the simulator and 

technical support center to determine whether the event classification and notification 

activities were performed in accordance with procedures. The inspectors also attended 

the individual facility drill critique to compare inspector observations with those identified 

by NextEra staff, to evaluate NextEra’s critique and to verify whether NextEra staff was 

properly identifying weaknesses and entering them into the CAP. 

b. Findings 

No findings were identified. 

.2 Emergency Preparedness Training Observations 

a. Inspection Scope 

The inspectors observed a simulator training evolution for Unit 1 licensed operators on 

January 23, 2014, which involved simulated emergency plan implementation by an 

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

performance indicator data regarding drill and exercise performance. The inspectors 

observed event classification and notification activities performed by the crew. The 

inspectors also attended the post-evolution critique for the scenario. The focus of the 

inspectors’ activities was to note any weaknesses and deficiencies in the crew’s 

performance and ensure that NextEra evaluators noted the same issues and entered 

them into the CAP. 

b. Findings 

No findings were identified. 

2. RADIATION SAFETY 

Cornerstone: Public Radiation Safety 

2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01) 

a. Inspection Scope 

From March 4 to March 8, 2014, the inspectors reviewed NextEra’s performance in 

assessing the radiological hazards and exposure control in the workplace. The 

inspectors used the requirements in 10 CFR Part 20 and guidance in Regulatory Guide 

(RG) 8.38, “Control of Access to High and Very High Radiation Areas for Nuclear Plants,” 

TSs, and the NextEra procedures required by TSs, as criteria for determining 

compliance. 

Inspection Planning 

The inspectors reviewed 2013 NextEra performance indicators for the occupational 

exposure cornerstone for Seabrook Nuclear Station. The inspectors reviewed the 

results of radiation protection program audits. The inspectors reviewed any reports of 

operational occurrences related to occupational radiation safety since the last inspection. 
14 

Enclosure 

Radiological Hazard Assessment 

The inspectors determined if there have been changes to plant operations since the last 

inspection that may result in significant new radiological hazards. The inspectors 

evaluated whether NextEra assessed the potential impact of these changes and has 

implemented periodic monitoring, as appropriate, to detect and quantify the radiological 

hazard. 

The inspectors reviewed the last two radiological surveys from the Fuel Transfer Canal 

and Letdown Line in the Demineralizer Alley. The inspectors evaluated whether the 

thoroughness and frequency of the surveys were appropriate for the given new 

radiological hazard. 

The inspectors conducted walk-downs and independent radiation measurements in the 

facility, including radioactive waste processing, storage, and handling areas to evaluate 

material and radiological conditions. 

The inspectors reviewed one risk-significant work activity that involved exposure to 

radiation. This activity was the initial entry, survey and decontamination of the fuel 

transfer canal following the second dry spent fuel storage cask loading campaign. For 

this work activity, the inspectors assessed whether the pre-work surveys performed were 

appropriate to identify and quantify the radiological hazard and to establish adequate 

protective measures. The inspectors evaluated the radiological survey program to 

determine if radiological hazards were properly identified (e.g., discrete radioactive hot 

particles, transuranics and hard to detect nuclides in air samples, transient dose rates 

and large gradients in radiation dose rates). 

The inspectors observed work in potential airborne radioactivity areas, and evaluated 

whether the air samples from the fuel transfer canal air sample locations were 

representative of the breathing air zone and were properly evaluated. 

Instructions to Workers 

The inspectors selected three containers of radioactive materials and assessed whether 

the containers were labeled and controlled in accordance with 10 CFR Part 20 

requirements. 

The inspectors reviewed the following radiation work permits (RWP) used to access high 

radiation areas (HRA) and evaluated if the specified work control instructions and control 

barriers were consistent with TS requirements for HRA. 

 RWP 14-0015 High Integrity Container/Liner Shipping Preparation to include 

Capping, Weighing and Transfer to Waste Processing Building, January 1, 2014 

 RWP 14-0022 Inspect CS Valves inside Letdown Valve Room at Power, 

January 15, 2014 

 RWP 14-0027 Primary Auxiliary Building Demineralizer Alley Work/Entry, 

December 31, 2013 

 RWP 14-0058 Fuel Storage Building Transfer Canal Radiation Protection Survey, 

Decontamination and Maintenance Support Activities, March 2, 2014 
15 

Enclosure 

For these RWPs, the inspectors assessed whether allowable stay-times or permissible 

dose for radiologically significant work under each RWP were clearly identified. The 

inspectors evaluated whether electronic personnel dosimeter (EPD) alarm set-points 

were in conformance with survey indications and plant procedural requirements. 

The inspectors reviewed two occurrences where a worker’s EPD malfunctioned or 

alarmed. The inspectors evaluated whether workers responded appropriately. The 

inspectors assessed whether the issue was included in the corrective action program 

and whether compensatory dose evaluations were conducted as appropriate. 

For work activities that could suddenly increase radiological conditions, the inspectors 

assessed the NextEra means to inform workers of these changes. 

Contamination and Radioactive Material Control 

The inspectors reviewed NextEra‘s criteria for the survey and release of potentially 

contaminated material. The inspectors evaluated whether there was sufficient 

procedural guidance on alarm response. 

The inspectors reviewed NextEra‘s procedures and records to verify that the radiation 

detection instrumentation was used at an appropriate sensitivity level. The inspectors 

selected six sealed sources from the NextEra inventory records and reviewed whether 

the sources were accounted for and were tested for loose surface contamination. 

The inspectors evaluated whether any recent transactions involving nationally tracked 

sources were reported in accordance with 10 CFR Part 20 requirements. 

Radiological Hazards Control and Work Coverage 

The inspectors evaluated ambient radiological conditions and performed independent 

radiation measurements during walk-downs of the facility. The inspectors assessed 

whether the conditions were consistent with applicable posted surveys, RWPs, and 

associated worker briefings. 

The inspectors examined NextEra physical and programmatic controls for highly 

activated or contaminated materials stored within spent fuel and other storage pools. 

The inspectors assessed whether appropriate controls were in place to preclude 

inadvertent removal of these materials from the pool. 

The inspectors examined the posting and physical controls for selected HRAs, locked 

high radiation area (LHRA) and very high radiation areas (VHRA) to verify conformance 

with the occupational performance indicator. 

Risk-Significant HRA and VHRA Controls 

The inspectors discussed with the Radiation Protection Manager (RPM) the controls and 

procedures for high-risk HRAs and VHRAs. The inspectors assessed whether any 

changes to NextEra relevant procedures reduce the effectiveness of worker protection. 
16 

Enclosure 

The inspectors evaluated NextEra controls for VHRAs and areas with the potential to 

become a VHRA to ensure that an individual was not able to gain unauthorized access 

to these areas. 

Problem Identification and Resolution 

The inspectors evaluated whether problems associated with radiation monitoring and 

exposure control were being identified by NextEra at an appropriate threshold and were 

properly addressed for resolution in the licensee’s corrective action program. The 

inspectors assessed the appropriateness of the corrective actions for problems 

documented by NextEra that involve radiation monitoring and exposure controls. The 

inspectors assessed NextEra processes for applying operating experience to their plant. 

b. Findings 

No findings were identified. 

2RS2 Occupational ALARA Planning and Controls (71124.02) 

a. Inspection Scope 

The inspectors assessed performance with respect to maintaining occupational 

individual and collective radiation exposures as low as is reasonably achievable 

(ALARA). The inspectors used the requirements in 10 CFR Part 20, RG 8.8, 

“Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear 

Power Plants will be As Low As Is Reasonably Achievable,” RG 8.10, “Operating 

Philosophy for Maintaining Occupational Radiation Exposure As Low as Is Reasonably 

Achievable,” TSs, and NextEra procedures required by TSs, as criteria for determining 

compliance. 

Inspection Planning 

The inspectors reviewed information regarding Seabrook’s collective dose history, 

current exposure trends, and ongoing or planned activities in order to assess current 

performance and exposure challenges. The inspectors reviewed the plant’s three year 

rolling average collective radiation exposure. 

The inspectors compared the site-specific trends in collective exposures against the 

industry average values and from similar nuclear power plants. In addition, the 

inspectors reviewed any changes in the radioactive source term by reviewing the trend 

in average contact dose rate with reactor coolant piping and steam generator primary 

channel head space and manways. The inspectors reviewed site-specific procedures 

associated with maintaining occupational exposures ALARA, which included a review of 

processes used to estimate and track exposures from specific work activities. 

Radiological Work Planning 

The inspectors assessed whether NextEra planning identified appropriate dose 

reduction techniques; considered alternate dose reduction features; and estimated 

reasonable dose goals. The inspectors evaluated whether NextEra‘s ALARA 

assessment had taken into account decreased worker efficiency from use of respiratory 

protective devices and/or heat stress mitigation equipment. The inspectors determined 
17 

Enclosure 

whether NextEra work planning considered the use of remote technologies as a means 

to reduce dose and the use of dose reduction insights from industry operating 

experience and plant-specific lessons learned. The inspectors assessed the integration 

of ALARA requirements into work procedure and RWP documents. 

Source Term Reduction and Control 

The inspectors used licensee records to determine the historical trends and current 

status of plant source term known to contribute to elevated facility collective dose. The 

inspectors assessed whether the licensee had developed contingency plans for 

expected changes in the source term as the result of changes in plant fuel performance 

issues or changes in plant primary chemistry. 

Problem Identification and Resolution 

The inspectors evaluated whether problems associated with ALARA planning and 

controls are being identified by the licensee at an appropriate threshold and were 

properly addressed for resolution in the licensee’s corrective action program. The 

inspectors assessed NextEra‘s process for applying operating experience to their plant. 

b. Findings 

No findings were identified. 

4. OTHER ACTIVITIES 

4OA1 Performance Indicator Verification (71151) 

 Unplanned Scrams, Unplanned Power Changes, and Unplanned Scrams with 

Complications (3 samples) 

a. Inspection Scope 

The inspectors reviewed NextEra’s submittals for the following Initiating Events 

Cornerstone performance indicators for the period of January 1, 2013 through 

December 31, 2013. 

 Unplanned scrams per 7000 critical hours 

 Unplanned scrams with complications 

 Unplanned power changes per 7000 critical hours 

To determine the accuracy of the performance indicator data reported during those 

periods, inspectors used definitions and guidance contained in Nuclear Energy Institute 

(NEI) Document 99-02, “Regulatory Assessment Performance Indicator Guideline,” 

Revision 6. The inspectors reviewed NextEra’s operator narrative logs, event reports, 

and NRC integrated inspection reports to validate the accuracy of the submittals. 

b. Findings 

No findings were identified. 
18 

Enclosure 

4OA2 Problem Identification and Resolution (71152 – 2 samples) 

.1 Routine Review of Problem Identification and Resolution Activities 

a. Inspection Scope 

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

inspectors routinely reviewed issues during baseline inspection activities and plant 

status reviews to verify that 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 

condition report screening meetings. 

b. Findings 

No findings were identified. 

.2 Annual Sample: Increasing Frequency of Leaks in Service Water Piping in the Vicinity of 

Installation/Fabrication Welds 

a. Inspection Scope 

During the period January 27 to January 31, 2014, inspectors reviewed a root cause 

evaluation (RCE AR 16379222) completed by NextEra staff for a service water pipe leak 

that occurred in August 2013. This problem was described in a licensee event report 

submitted to the NRC dated December 23, 2013. The inspectors determined the 

effectiveness of actions by NextEra staff to identify, characterize, correct and prevent 

reoccurrence of SW system leaks. 

The inspectors assessed problem identification threshold, apparent cause analysis, 

extent of condition reviews, and timeliness of corrective actions. The inspectors 

reviewed documents listed in the Attachment to this report and interviewed NextEra 

engineering personnel to assess the effectiveness of the planned, scheduled, and 

completed corrective actions to resolve the identified deficiency. 

The inspectors reviewed non-destructive test procedures, procedure qualifications 

including test personnel qualifications to determine compliance with the applicable 

American Society of Mechanical Engineers codes and standards. Also, the inspectors 

reviewed system health reports, work orders, procurement documents, drawings and 

photographs to determine if the nonconforming condition was appropriately identified, 

documented, characterized and entered into NextEra’s corrective action process. 

The inspectors reviewed root cause evaluation AR 16379222 and interviewed members 

of the evaluation team. The inspectors interviewed the qualified non-destructive test 

examiner to evaluate the ultrasonic test method used. Test results were reviewed with 

the test examiner to assess the remaining wall thickness for continued operation without 

encroaching on minimum wall requirements. 

 
19 

Enclosure 

b. Findings and Observations 

No findings were identified. The root cause evaluation and corrective actions were 

reasonable, appropriate and timely. 

NextEra’s root cause evaluation addressed a history of SW degradation (corrosion/ 

erosion) resulting in wall thinning and pressure boundary penetration and leakage. The 

areas where wall thinning and leakage occurred was determined to be associated with 

the loss of protective coating and/or liner failure at fabrication/installation welds which 

typically results in turbulent fluid flow. This turbulent flow was particularly aggressive in 

the attack of base metals and protective coatings at these weld locations and 

configuration changes. The inspectors assessed the root cause determination, results of 

the extent of condition investigation of other locations within the SW system and other 

fluid (circulating water) systems with similar piping materials, operating parameters and 

configurations. 

The inspectors noted that examination using ultrasonic testing was performed at 

selected locations with known change in flow patterns and velocity changes. The results 

of this testing identified areas exhibiting variable wear rates. An evaluation of these test 

results was made to determine pipe structural and pressure retaining integrity. 

The inspectors visually examined several portions of previous SW pipe and fittings that 

had been removed from the SW system in prior outages due to identified leaks. The 

removed samples provided confirmatory evidence of corrosive/erosive attack from 

turbulent flow at root locations of field welds, configuration changes and pipe to fitting 

intersections. These locations revealed characteristic “pin hole” leaks at weld locations 

and a general “wastage” of pipe and fitting interior diameters. These locations were 

evaluated for compliance with minimum wall thickness requirements. Those locations 

which were identified as active “leaks” at weld locations or, where areas exhibiting loss 

of wall thickness and were encroaching on minimum wall requirements, were 

dispositioned for repair/replacement in the CAP. 

The inspectors determined that this issue received appropriate management attention as 

indicated by the corrective action that was taken to perform a temporary leak repair by 

the installation of a weldolet encapsulating the leak location. At the next outage, (OR16) 

the weldolet will be removed and replaced with a more suitable “flush patch”. The patch 

will be coated internally with a corrosion/erosion resistant material. The inspectors 

discussed the licensee plans to systematically remove and replace the SW piping with a 

base metal that is significantly more resistant to erosion/corrosion attack. The 

inspectors examined numerous lengths of pipe and fittings which were staged for 

replacement in the plant during the April 2014 refueling outage and subsequent outages. 

.3 Annual Sample: Review of Activities Associated with Alkali Silica Reaction Affected 

Structures 

a. Inspection Scope 

March 12 to 13, NRC inspectors from Region I and a structural engineer from the 

Division of License Renewal, NRR, witnessed testing conducted at the Ferguson 

Structural Engineering Laboratory (FSEL) at the University of Texas – Austin. The 

testing was conducted in support of the Seabrook Alkali-Silica Reaction (ASR) 
20 

Enclosure 

Integrated Corrective Action Plan. Specifically, the inspectors witnessed load testing of 

the control beam for reinforcement anchorage (lap-splice) capacity. The testing was 

performed in accordance with MPR Project 0326-0063, Procedure 5-7, “Structural 

Testing of Shear and Anchorage Specimens,” Revision 1. 

The inspectors also reviewed the results of the December 2013, Combined Crack 

Indexing (CCI) measurements and the supporting engineering analysis. Proprietary data 

sheets and associated evaluations were made available for inspector review. 

Additionally, the inspectors’ review included discussions with the responsible Seabrook 

engineers, as well as petrography specialists consulting for the University of Texas 

Ferguson Structural Engineering Laboratory 

Lastly, the inspectors reviewed NextEra’s revised Prompt Operability Determinations 

(PODs) that address additional Seabrook structures identified as being affected by ASR 

via the Phase 3 ASR walkdown program. 

b. Findings and Observations 

The inspectors identified no findings. 

Review of CCI and Crack Width Measurements 

The inspectors examined the December 2013 CCI measurement results documented in 

Foreign Print (FP) 100847 and FP 100848, dated January 30, 2014. As documented in 

these NextEra reports, there are 32 areas currently being monitored for ASR progression 

using the CCI and crack width methodology. As of December 2013, 26 of the 32 areas 

have been monitored on a six-month basis for approximately two years. Based upon the 

data collected to date, NextEra has concluded the following: 1) the data suggests a slow 

increasing trend in CCI and crack width over the past two years; 2) at 14 interior ASR 

locations, the horizontal and vertical CCI data indicates an overall upward trend (an 

average increase of 0.04 mm/m, with a measurement tolerance of 0.05 mm); 3) at nine 

exterior locations, the horizontal and vertical CCI data indicates no significant change 

over the two-year period; 4) the six floor/ceiling/roof locations indicate flat to upward 

trends early in the period, but no change later in the two-year period; and, 5) some 

fluctuation in the measured CCI and crack width values have been observed. The 

fluctuations may be attributed to thermal effects, cyclic or constant moisture exposure, 

measurement device accuracy, or the condition of the measured surfaces as impacted 

by cleaning/preparation and weathering effects. Independent inspector review of CCI 

and crack width measurement data and photographs of selected areas confirmed 

NextEra’s conclusions. As stated in FP 100847 and FP 100848, NextEra will continue 

the six-month data collection to comply with the Structures Monitoring Program and 

validate these observed ASR progression trends. 

Review of Operability Determinations 

Based on the result of recent Phase 3 walkdowns, NextEra identified six additional areas 

with CCI > 1.0 millimeter per meter (mm/m). In accordance with the Seabrook 

Structures Monitoring Program (SMP), NextEra staff completed evaluations of the 

affected structures to assess the potential impact of ASR on continued operability. The 

six additional areas identified with a CCI value greater than 1.0 mm/m were: cooling 

tower exterior (elevation 25’, reference CTE-02); primary auxiliary building (PAB) 
21 

Enclosure 

penetration area (elevation -26’, reference MF105-01); west pipe chase (elevation 12’, 

reference MF202-02); and three areas in electrical manholes (below grade elevations, 

reference CI-W03-Wall, CI-W05-Wall, and CI-W11-Wall). The structural evaluations 

were documented as Supplement IV and V to FP 100716, “Seabrook Station: Impact of 

ASR on Concrete Structures and Attachments,” and utilized the same design capacity 

versus calculated demand margin analysis approach as the previously completed POD’s 

(reference Section 9.1, NRC Inspection Report 05000443/2012010). Inspector review of 

the six additional ASR-affected area PODs concluded that the impacted structures have 

adequate strength margin available and are fully capable of performing their safety 

functions. 

Control Beam Testing Observations 

The inspectors witnessed the performance of load testing of the first control beam (a 

specimen that has not undergone ASR aging). The beam (A-7, reinforcement 

anchorage control specimen) was tested in accordance with MPR Project 0326-0063, 

Procedure 5-7, “Structural Testing of Shear and Anchorage Specimens,” Revision 1, on 

March 13, 2014. Beam failure occurred at a load of approximately 251,000 pounds, as 

compared to the estimated failure load of 214,000 pounds. The failure load was slightly 

higher than the estimated design capacity, but within the accuracy of the design 

calculations. No test anomalies were identified. The results of the control beam test will 

be used to compare subsequent ASR-affected specimen tests to evaluate the impact of 

ASR degradation on structural performance. The inspectors observed proper procedural 

adherence, good test coordination and proper communications and safety practices 

exhibited by the testing staff, supervisory personnel and quality assurance overseers. 

The inspectors verified proper testing preparations and quality control oversight as 

specified by MPR Project 0326-0063, Procedure 5-6, “General Preparation of Test 

Facilities and Specimens,” Revision 2, and MPR Procedure 0326-0062-46, “Procedure 

for In-process Inspections of FSEL Reinforcement Anchorage Test Setup for Seabrook 

Station,” Revision 0. 

Initial Test Specimen ASR Expansion Results 

 

NextEra and the UT-Austin FSEL staff have observed in the large-scale test specimens 

that the X- and Y-direction deep pin expansion measurements (comparable to the 

Seabrook vertical and horizontal wall surface CCI measurements) do not appear to 

correlate with the through-wall (e.g., out-of-plane, or Z-direction) deep pin expansion 

measurements after the initial phase of ASR expansion. X- and Y-direction expansion 

appears to plateau while the Z-direction expansion continues to trend upward (increase). 

All large-scale reinforcement anchorage and shear specimens have demonstrated this 

expansion trend. The Z-direction expansion in the test specimens has been observed to 

be 10 times greater than the X- and Y- expansions after approximately one year. 

The preliminary implication of these test specimen expansion measurement trends is 

that the X- and Y- expansion measurement methods (CCI and crack width) currently 

used for monitoring the progression of ASR on Seabrook Station structure surfaces (per 

the Structures Monitoring Program) may not provide alone, an adequate means to 

monitor (1) ASR progression and (2) by inference (pending the completion of the testing 

program), the ASR impact on the affected building’s structural performance. The 

validation of the use of the CCI and crack width measurements for monitoring the 

structural impact of ASR has been an objective of the large specimen testing program. 
22 

Enclosure 

In considering these initial test program results, NextEra staff initiated an Action Report 

(No. 01952162) to address this issue. In addition to evaluating the future impact of using 

CCI to monitor ASR progression on affected structures, NextEra staff conducted a 

preliminary assessment of this test data for impact to their PODs completed for ASRaffected Seabrook structures. NextEra staff concluded this initial test program data does 

not adversely impact the POD margins analyses, principally because the PODs are not 

dependent upon measured CCI values for assessing the level of ASR degradation. 

Instead, those structures demonstrating the most significant ASR progression (as 

assumed by CCI and crack width measurements of >1.0 mm/m and >1.0 mm, 

respectively) were evaluated using conservative and bounding degradation values 

derived from published research and testing data developed from non-reinforced 

concrete specimens. At the close of the inspection period, NextEra staff had initiated 

actions to re-evaluate the use of CCI and crack width for the SMP and to re-evaluate 

their methods for monitoring and assessing ASR progression of test specimens in order 

to correlate test data to Seabrook Station. 

NextEra staff communicated with the inspectors their plans to fabricate an additional 

large-scale test specimen to instrument with strain gages and allow ASR progression in 

order to validate the use of strain gages for through-wall (Z-direction) expansion 

monitoring. The purpose of this activity is to validate the use of one or more strain gage 

designs that can subsequently be installed in Seabrook structures to accurately monitor 

through-wall expansion. In addition, NextEra staff described plans to increase the core 

sampling of control and ASR-affected large scale test specimens in order to more 

accurately measure ASR impact on concrete compressive and tensile strength and 

modulus of elasticity. NextEra staff further communicated plans to conduct petrographic 

examination of through-wall core samples from the test specimens. This additional 

concrete material property testing and data collection is intended to be used to support 

the correlation of testing program structural performance data to Seabrook structures 

(along with additional, but not yet defined, core sampling of Seabrook ASR-affected 

structures). 

In summary, the inspectors concluded that the PODs completed for ASR-affected 

Seabrook structures remain unaffected by the X-,Y- and Z-direction expansion data 

measured, to date, in the test specimens. Actions planned by NextEra to assess the 

adequacy of the SMP structural evaluation criteria and modify the ASR testing program 

were viewed appropriate by the inspectors and the Seabrook ASR Issue Technical 

Team, at this time. As stated above, the PODs use bounding assumptions not 

dependent on the degree of expansion measured. Additional inspections are planned by 

the NRC to evaluate NextEra’s ongoing corrective actions to resolve the nonconformance related to ASR-affected structures at the Seabrook Station. 

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

 Plant Events 

a. Inspection Scope 

For the degraded plant equipment transient listed below, the inspectors reviewed and/or 

observed plant parameters, reviewed personnel performance, and evaluated 

performance of mitigating systems. The inspectors communicated the plant events to 
23 

Enclosure 

appropriate regional personnel, and compared the event details with criteria contained in 

IMC 0309, “Reactive Inspection Decision Basis for Reactors,” for consideration of 

potential reactive inspection activities. The inspectors reviewed NextEra’s follow-up 

actions related to the events to assure that NextEra implemented appropriate corrective 

actions commensurate with their safety significance. 

 ‘B’ steam generator feedwater bypass valve failed to open on March 14, 2014 

b. Findings 

No findings were identified. 

4OA5 Other Activities 

.1 Cross-Cutting Aspects 

The table below provides a cross-reference from the 2013 and earlier findings and 

associated cross-cutting aspects to the new cross-cutting aspects resulting from the 

common language initiative. These aspects and any others identified since January 

2014 will be evaluated for cross-cutting themes and potential substantive cross-cutting 

issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment 

review. 

 Finding Old CrossCutting Aspect New CrossCutting Aspect 

05000443/2013004-01, Inadequate 

Operability Determination Regarding Service 

Water Leakage and Associated TS Violation 

H.1(b) H.14 

.2 Buried Piping, TI-2515/182, Phase 2 (1 sample) 

a. Inspection Scope 

The licensee’s buried piping and underground piping and tanks program was inspected 

in accordance with paragraphs 03.02.a of the Temporary Instruction (TI) 2515/182, and 

it was confirmed that activities which correspond to the completion dates specified in the 

program, that have passed since the Phase 1 inspection was conducted, have been 

completed. 

The licensee’s buried piping and underground piping and tanks program was inspected 

in accordance with paragraph 03.02.b of the TI and responses to specific questions 

found in http:www.nrc.gov/reactors/operating/ops-experience/buried-pipe-ti-phase-2-

insp-req-2011-11-16.pdf were submitted to NRC headquarters staff. 

b. Findings 

No findings were identified. 

 
24 

Enclosure 

4OA6 Meetings, Including Exit 

On April 10, 2014, the inspectors presented the inspection results to Mr. Kevin Walsh, 

Site Vice President, and other members of the Seabrook Station staff. The inspectors 

verified that no proprietary information was retained by the inspectors or documented in 

this report. 

ATTACHMENT: SUPPLEMENTARY INFORMATION 
A-1 

Attachment 

SUPPLEMENTARY INFORMATION 

KEY POINTS OF CONTACT 

Licensee Personnel 

K. Walsh, Site Vice President 

T. Vehec, Plant General Manager 

V. Brown, Senior Licensing Engineer 

M. Chevalier, RP Supervisor 

J. Connolly, Site Engineering Director 

D. Currier, Emergency Planning Manager 

K. Douglas, Maintenance Director 

D. Flahardy, Radiation Protection Manager 

M. Ossing, Licensing Manager 

V. Pascucci, Nuclear Oversight Manager 

D. Robinson, Chemistry Manager 

T. Waechter, Nuclear Plant Shift Manager 

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED 

Opened/Closed 

05000443/2014002-01 NCV Scaffolding Installed with Insufficient Separation to 

Safety Related Equipment (Section 1R15) 

 

LIST OF DOCUMENTS REVIEWED 

Section 1R01: Adverse Weather Protection 

Procedures 

ON1090.13, Response to Natural Phenomena Affecting Plant Operations, Revision 1 

OS1200.03, Severe Weather Conditions, Revision 20 

OS1090.09, Station Cold Weather Operations, Revision 2 

Section 1R04: Equipment Alignment 

Procedures 

OS1026.02, Operating the DG 1A Lube Oil System, Revision 14 

OS1026.03, Operating DG 1A Jacket Water Cooling System, Revision 11 

OS1026.04, Operating DG 1A Starting Air System, Revision 12 

OS1026.05, Operating the DG 1A Fuel Oil System, Revision 14 

OS1026.06, Operating the DG 1A Air Intake, Exhaust and Vacuum System, Revision 9 

OS1036.01, Aligning the Emergency Feedwater System for Automatic Operation, Revision 17 

OX1461.03, SEPS Operational Readiness Status Surveillance, Revision 1 

 
A-2 

Attachment 

Condition Reports 

1880681 1904097 1936382 1934562 1952234 

Section 1R05: Fire Protection 

Procedures 

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, PAB-F-1A-Z, PAB-F-1J-Z

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, RHR-F-1B-Z, RHR-F-2B-Z, 

 RHR-3B-Z, RHR-F-4B-Z, RHR-F-4B-Z1, RHR-F-4B-Z2 

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, RHR-F-1A-Z, RHR-F-2A-Z, 

RHR-3A-Z, RHR-F-4A-Z, RHR-F-4A-Z1, RHR-F-4A-Z2 

Seabrook Station Fire Protection Pre-Fire Strategies, Volume II, SEPS-F-1-0 21’-0” 

Seabrook Station Fire Protection Pre-Fire Strategies, Volume II, SW-F-1A-Z

Section 1R06: Flood Protection Measures 

Condition Reports 

00158490 01939967 

Miscellaneous 

Seabrook Station Moderate Energy Line Break Study 

Section 1R11: Licensed Operator Requalification Program 

Procedures 

ER 1.1, Classification of Emergencies, Revision 53 

Section 1R12: Maintenance Effectiveness 

Procedures 

ER-AA-201-2001, System and Program Health Reporting, Revision 4 

PEG-40, Scoping Changes and Program Interfaces, Revision 5 

PEG-45, Maintenance Rule Program Monitoring Activities, Revision 17 

Condition Reports 

1625261 1927188 1927781 1932096 1932711 1933065 

1936449 1945056 

Miscellaneous 

ED/EDE 120 VAC System Health Report 

EE-10-010, Maintenance Rule–PRA Basis Document PRA Risk Ranking and performance 

 Criteria based on SSPSS-2009, Revision 1 

NEI-99-02, Revision 7 

SW-P-41C In-Service Testing Pump Data Sheet and Data Logs 

Service Water System Health Report 

Section 1R13: Maintenance Risk Assessments and Emergent Work Control 

Procedures 

OS1046.24, Removing EDE-I-1B from Service during Power Operation, Revision 3 

Miscellaneous 

Maintenance Rule a(4) Risk Assessment Report for Workweek 1407-12 

PRA-301, MR (a)(4) Process for On-Line Maintenance Group Instruction, Revision 0 

WM-AA-100-1000, Work Activity Risk Management, Revision 0 
A-3 

Attachment 

Maintenance Orders/Work Orders 

40248951 40295189 

Section 1R15: Operability Determinations and Functionality Assessments 

Procedures 

IX1670.905, Seismic Monitoring Data Retrieval Following a Seismic Event, Revision 6 

IX1670.919, SWPH Seismic Monitor Calibration, Revision 6 

MA 4.8, Control of Scaffolding, Revision 10 

MA 5.7, Station Barriers, Penetration Seals, and Fire Barrier Wrap, Revision 17 

MS0599.16, Construction, Repair and Rework of Silicone Base Penetration Seals, Revision 7 

MS0599.47, Erection of Scaffolding, Revision 2 

MX0539.50, Emergency Diesel Generator Engine 34-Month Preventative Maintenance, 

 Revision 6 

MX0599.02, 18 Month Inspection of Technical Requirement Fire Rated Assembly Penetration 

 Seals, Revision 2 

OS1005.05, Safety Injection System Operation, Revision 25 

OX1413.01, ‘A’ Train RHR Quarterly Flow and Valve Stroke Tests, and 18 Month Valve Stroke 

 Observation, Revision 20 

OX1413.03, ‘B’ Train RHR Quarterly Flow and Valve Stroke Tests, and 18 Month Valve Stroke 

 Observation, Revision 11 

OX1426.34, Diesel Generator 1A 18 Month Operability Surveillance, Revision 10 

OX1436.02, Turbine Driven Emergency Feedwater Pump Quarterly and Monthly Valve 

 Alignment, Revision 20 

Condition Reports 

0182029 0216388 0214364 0221647 1612785 1804255 

1833819 1914234 1918208 1928775 1930569 1930855 

1933827 1934585 1935442 1936576 1937513 1937679 

1941086 1941153 1942147 1945355 1945771 1946400 

1947827 1949876 

 

Maintenance Orders/Work Orders 

40153374 40191167 40199273 40228515 40245357 40250128 

40291172 40291812 40291925 40292272 

Miscellaneous 

Calculation C-S-1-61035, Allowable CEVA Penetration Seal Opening Size, Revision 3 

Colt-Pielstick PC2V Engine Vendor Manual 

EE-98-019, Control of Temporary Loads in Seismic Areas 

Engineering Evaluation 14-001, Scaffold & Temporary Equipment Engineering Evaluation 

FP22849, Terry Turbine Instruction Manual, Revision 1 

Preventive Maintenance Activity MS-CAT-12-SEAL-INSP 

Penetration Seal Design, Seal No. PB-021-EV101-7502, FP4490R-01 

Penetration Seal Design, Seal No. PB-021-EV101-7504, FP4492R-01 

Tech Spec and Commitment Logs dated January 2–3, 2014 

Technical Requirement TR21-4.3.3.3.1 

Drawings 

1-RH-B20662, Residual Heal Removal Sys. Train ‘A’ Detail, Revision 22 
A-4 

Attachment 

Section 1R18: Plant Modifications 

Procedures 

IS1672.141, SW-F-6181, DG-E-42A Jacket Water Cooler Service Water Outlet Flow Calibration, 

 Revision 6 

EN-AA-100, Design Control Program, Revision 1 

EN-AA-100-1003, Control of Design Interfaces, Revision 1 

EN-AA-205-1100, Design Change Packages, Revision 9 

EN-AA-205-1103, Equivalent Design Package, Revision 0 

Condition Reports 

1860416 1881903 1939926 

Maintenance Orders/Work Orders 

40233615 

Miscellaneous 

Specification 9763-006-174-1D, Data Sheets for Electronic Transmitters (Non-Class 1E), 

 Revision 14 

Drawings 

1-SW-D20795, Service Water System Nuclear Detail, Revision 43 

Section 1R19: Post-Maintenance Testing 

Procedures 

LS0556.08, Routine Preventative Maintenance 7.5 KVA Westinghouse Inverter, Revision 8 

LS0556.09, Replacement of Ferro-Resonant Transformers and Capacitors in Westinghouse 7.5 

 KVA Inverters, Revision 5 

MM-AA-100, Conduct of Maintenance, Revision 4 

MX0539.63, Emergency Diesel Generator Exhaust Valve Removal, Replacement, and Belleville 

 Washer Replacement, Revision 2 

ON0443.113, Portable Diesel Driven Pump Annual Functional Test, Revision 5 

OS1046.24, Removing EDE-I-1B from Service during Power Operation, Revision 2 

OS1047.01, Vital Inverter Operation, Revision 14 

OS1247.01, Loss of a 120VAC Instrument Panel, Revision 17 

OS1412.10, Thermal Barrier Cooling Water Pump Monthly Rotation, Revision 6 

OX1416.04, Service Water Quarterly Pump and Discharge Valve Test and Comprehensive 

Pump Test, Revision 19 

OX1446.03, Electrical Bus Weekly Operability, Revision 12 

OX1456.86, Operability Testing of IST Pumps, Revision 10 

SAG-9, PDDP and Hose Trailer Deployment, Revision 5 

Condition Reports 

0221649 1931807 1932393 1932711 1933020 1933808 

1934499 1934512 1934562 1936449 1936703 1936858 

1940192 1940751 1941180 1945142 1946434 1946440 

1946653 1947234 1947394 

Maintenance Orders/Work Orders 

01168951 01207733 01210186 0305866 0320197 40196428 

40228208 40239216 40259353 40290809 40295189 40298167 

40298355 
A-5 

Attachment 

Miscellaneous 

Calculation C-S-1-86208, Extreme Damage Mitigating Strategy Flow Capability, Revision 3 

Calculation C-S-1-50014, SW Pumps (SW-P-41A thru D) IST Uncertainties, Revision 0 

Calculation 9763-3-ED-00-34-F, AC Ground Detection System 

FP35465, SEPS Generator Set Technical Manual 

FP500076, Godwin PDDP Instruction Manual, Revision 6 

NASA TN D-8177, Apollo Experience Report–Detection and Minimization of Ignition Hazards 

From Water/Glycol Contamination of Silver-Clad Electrical Circuitry 

Standing Order SOO-14-001, B.5.b Pump Status, dated January 28, 2014 

Westinghouse Technical Bulletin NSID-TB-87-09 

Drawings 

4950C70, Sheet 4, Inverter Schematic 

1-NHY-310105, Sheet E02a, UPS 1-I-1B Vital Instrument Distribution Panel 1-PP-1B 

1-NHY-310231, Sheet I20a, Motor/Load List Motor Control Center 1-EDE-MCC-615, Revision 7 

1-NHY-310895, Sheet B4Qa, Thermal Barrier PCCW Recirc. Pump P-322B Schematic 

 Diagram, Revision 2 

Section 1R22: Surveillance Testing 

Procedures 

EN-AA-205-1102, Temporary Configuration Changes, Revision 5 

MA-AA-100-1011, Equipment Troubleshooting, Revision 0 

ON0443.113, Portable Diesel Driven Pump Annual Functional Test, Revisions 1 and 5 

ON0443.114, 18 Month B.5.b Equipment Inventory Surveillance, Revision 10 

OS1001.04, RCS Unidentified Leak Rate Action Level Exceedence, Revision 0 

OS1007.01, Automatic and Manual Rod Control, Revision 12 

OX1408.06, Controlled Leakage Monthly Surveillance, Revision 6 

OX1416.06, Service Water Cooling Tower Pumps’ Quarterly and 2 Year Comprehensive Test, 

 Revision 21 

OX1423.07, Monthly Testing of Train A Enclosure Emergency Exhaust, Revision 8 

OX1426.05, DG 1B Monthly Operability Surveillance, Revision 28 

OX1426.19, Aligning DG 1B Controls for Auto Start, Revision 3 

OX1456.46, Train B ESFAS Slave Relay K608 Quarterly Go Test, Revision 7 

Condition Reports 

1822620 1929096 1933872 1934512 1941467 1949526 

1949825 

Maintenance Orders/Work Orders 

40149909 40178159 40235580 40244087 40245450 40259353 

40264853 40246660 40246721 40246722 40287931 40290041 

40290042 

Miscellaneous 

1-SW-OT-031 IST Pump Data Log 

1-SW-OT-011 IST Pump Data Log 

Activity 1-CP-CP-113-CRDM-1, CRDM Current Command Trace Acquisition, Revision 6 

ASME OM CODE-2004 

Calc 88-002, IST Calculation of Total Developed Head for Service Water and Cooling Tower 

Pumps 

Engineering Evaluation SS-EV-98006, Revision 1 
A-6 

Attachment 

Drawings 

1-CS-B20725, Chemical & Volume Control Sys. Seal Water Detail, Revision 20 

1-NHY-310932, Cntmnt Encl Emer Exh Fan 1-FN-4A Schematic Diagram, SH-BB3a, Revision 9 

1-NHY-310932, Cntmnt Encl Emer Exh Fan 1-FN-4A Legend & SW Development, SH-BB3b, 

 Revision 10 

1-NHY-503515, EAH – Contn. Encl. Emer Exh Fltr Fan Logic Diagram, Revision 7 

Section 1EP4: Emergency Action Level and Emergency Plan Changes 

Procedures 

ER 3.3, Emergency Operations Facility Operations, Revision 51 

Section 1EP6: Drill Evaluation 

Procedures 

EP-AA-101-1000, Nuclear Division Drill and Exercise Procedure, Revision 5 

ER 1.1, Classification of Emergencies, Revision 52 

ER 1.2, Emergency Action Plan Activation, Revision 61 

ER 3.1, Technical Support Center Operations, Revision 53 

Condition Reports 

1948051 1910629 

Miscellaneous 

ER 2.0B, Seabrook Station State Notification Fact Sheet, Revision 31 

Form EPDP-03A, EP Cornerstone Reporting and Information Form, Revision 23 

Section 2RS1: Radiation Hazard Assessment and Exposure Control 

Procedures 

HD0958.04, Posting of Radiologically Controlled Areas, Revision 33 

HD0958.03, Personnel Survey and Decontamination Techniques, Revision 24 

HD0958.13, Generation and Control of Radiation Work Permits, Revision 39 

HD095817, Performance of Routine Radiological Surveys, Revision 12 

HD0958.19, Evaluation of Dosimetry Abnormalities, Revision 37 

HD0958.38, Evaluation of Isotopic Mix, Revision 29 

HN0958.25, High Radiation Area Control, Revision 37 

HN0958.30, Inventory and Control of LHRA or VHRA Keys and Locksets, Revision 26 

HX0958.23, Radioactive Source Control, Revision 20 

RP-AA-100-1001, RP Conduct of Ops, Revision 3 

RP-AA-100-1002, Radworker Instructions and Responsibilities, Revision 1 

RP-AA-101, Personnel Monitoring Program, Revision 0 

RP-AA-101-1001, Radiation Protection Conduct of Operations, Revision 3 

RP-AA-101-1002, Dosimetry Data Processes for Sentinel Software, Revision 3 

RP-AA-101-2004, Method for Monitoring and Assigning Effective Dose Equivalent for High Dose 

 Gradient Work, Revision 3 

RP-AA-102-1001, Area Rad Surveys, Revision 0 

RP-AA-102-1000, Alpha Monitoring, Revision 0 

RP-AA-102-1002, Dosimetry Data Process for Sentinel, Revision 3 

RP-AA-103-1001, Posting Requirements, Revision 1 

RP-AA-103-1002, High Rad Controls, Revision 1 

RP-AA-107-1003, Unconditional and Conditional Release of Material, Revision 1 
A-7 

Attachment 

Audits, Self-Assessments, and Surveillances 

Seabrook Station Radiation Protection Department Self Evaluation and Trend Analysis 

Report for 4th Quarter 2013, January 31, 2014 

Quick Hit Assessment Report 1928716, NRC 71124.01 and .02 Radiological Hazard 

Assessment and ALARA Planning and Control, February 3, 2014 

Seabrook Nuclear Oversight Report SBK-14-001, Radiation Protection and Radwaste 

Programs, February 24, 2014 

Condition Report 

01836289 01855852 01898310 01903346 01906680 01934952 

01940807 01941338 01943228 01945353 01945687 

Miscellaneous 

Seabrook Updated Final Safety Analysis Report 

Seabrook Survey M-20140304-5, Initial Entry into Fuel Transfer Canal, March 4, 2014 

Seabrook Survey M-20140122-3, HSYQ066A FSB-21-FB202 Quarterly, January 22, 2014 

Seabrook Survey M-20140203-2, HSYQ-082B WPC-(-26)-MF-106 Quarterly, February 3, 2014 

Seabrook Survey M-20130904-3, HSXQ-079A PAB-7-PB309 Quarterly, September 4, 2013 

Seabrook Survey M-20131123-2, HSYQ-083A WPC-(-20)-MF102 Quarterly, November 23, 2013 

Seabrook RWP 14-0015 High Integrity Container/Liner shipping Preparation to Include Capping, 

 Weighing and Transfer to WPB, January 1, 2014 

Seabrook RWP 14-0022 Inspect CS Valves inside Letdown Valve Room at Power, January 15, 2014

Seabrook RWP 14-0027 PAB Demin Alley Work/Entry, December 31, 2013 

Seabrook RWP 14-0058 FSB Transfer Canal RP Survey, Decon and Maintenance Support 

 Activities, March 2, 2014 

Seabrook Non Exempt Source Inventory and Leak Test, September 14, 2013 

Seabrook Exempt Source Index, July, 22, 2013 

Seabrook Dosimetry Abnormality Occurrence Report CR 01904744, November 18, 2013 

Seabrook Dosimetry Abnormality Occurrence Report CR 01906680, December 4, 2013 

Seabrook 2014 Air Sample Log, March 7, 2014 

Seabrook Lesson Plan HP1188C Alpha Monitoring Course, April 15, April 29 and May 13, 2013 

Seabrook Log of VHRA and LHRA Access Points, March 5, 2014 

Seabrook LHRA In Service Key Box Log, January 31, 2014 

Seabrook LHRA/VHRA Key Issue Log, March 4, 2014 

Seabrook HRA/LHRA Briefing Acknowledgement Form, March 5, 2014 

Work Order 

40235669 

Section 2RS2: Occupational ALARA Planning and Controls 

Procedures 

RP-AA-104 ALARA Program, Revision 2 

RP-AA-104-1000, ALARA Implementing Procedure, Revision 5 

Audits, Self-Assessments, and Surveillances 

Seabrook Station Radiation Protection Department Self Evaluation and Trend Analysis 

 Report for 4th Quarter 2013, January 31, 2014 

Quick Hit Assessment Report 1928716, NRC 71124.01 and .02 Radiological Hazard 

Assessment and ALARA Planning and Control, February 3, 2014 

Seabrook Nuclear Oversight Report SBK-14-001, Radiation Protection and Radwaste 

 Programs, February 24, 2014 
A-8 

Attachment 

Corrective Action Document 

01836312 01843713 01856278 01867573 01872019 01883752 

01890162 01893578 01896323 01904259 01930630 01944341 

Miscellaneous 

EPRI Standard Radiation Monitoring Program Results through OR 15, September 25, 2012 

Seabrook Updated Final Safety Analysis Report

Seabrook Post Outage Critique: ALARA and Station Dose Performance, June 2013 

Seabrook Station Nuclear Plant 5-Year ALARA Plan 2013-2017, July 31, 2013 

Seabrook ALARA Review Board Meeting 13-04, December 11, 2013 

Seabrook ALARA Review Board Meeting 14-01, March 3, 2014 

Seabrook Temporary Shielding Log for OR 16, March 2014 

Pre-Job ALARA Review Package: 13-01 Dry Fuel Transfer from Pool to Pad and Associated 

 Tasks for 8 ISFSI Casks, June 27, 2013 

Post-Job ALARA Review: 13-01 Dry Fuel Transfer from Pool to Pad and Associated Tasks 

 for 8 ISFSI Casks, December 4, 2013 

Pre-Job ALARA Review Package 14-01 OR 16 Reactor Dissassembly and Reassembly, 

 December 26, 2013 

Pre-Job ALARA Review Package 14-02 OR 16 Steam Generator Eddy Current Testing and 

 Tube Plugging, Febuary 25, 2014 

Pre-Job ALARA Review Package 14-03 OR 16 In Service Inspection, Febuary 25, 2014 

Pre-Job ALARA Review Package 14-07 OR 16 Fuel Handling Project, Febuary 25, 2014 

Pre-Job ALARA Review Package 14-09 OR 16 RCP Seal Replacement, Febuary 25, 2014 

Pre-Job ALARA Review Package 14-10 OR 16 Scaffolding, Febuary 25, 2014 

Pre-Job ALARA Review Package 14-13 Replace Rx Ventillation Ducting Under Vessel with 

New Design, Febuary 25, 2014 

Post Project Critique Dry Fuel Storage, Seabrook Station 2nd Loading Campaign, October 2013: 

Section 4OA1: Performance Indicator Verification 

Procedures 

NAP-206, NRC Performance Indicators, Revision 6 

Miscellaneous 

LIC-13017, Documentation Supporting the Seabrook Station NRC 1st Quarter 2013 

 Performance Indicator Submittal 

LIC-13036, Documentation Supporting the Seabrook Station NRC 2nd Quarter 2013 

 Performance Indicator Submittal 

LIC-13037, Documentation Supporting the Seabrook Station NRC 3rd Quarter 2013 

 Performance Indicator Submittal 

LIC-14004, Documentation Supporting the Seabrook Station NRC 4th Quarter 2013 

 Performance Indicator Submittal 

MSPI Derivation Reports 

Section 4OA2: Problem Identification and Resolution 

Non-Destructive Test Reports 

40265240-01, UT Extent of Condition Thickness Examination, B Train 

40265234-01, UT Extent of Condition Thickness Examination, A Train 

Condition Reports 

01897164 01637922 
A-9 

Attachment 

Maintenance Orders/Work Orders 

40265234 40268662 40268965 40268967 94080896 94080893 

Drawings 

1-SW-B20794, Service Water System Nuclear Detail (Service Water Pump House) 

1-SW-B20795, Service Water System Nuclear Detail (Turbine Bldg., Aux Bldg.) 

SK-EC270504-2000, Installation Detail Service Water Piping Repairs 

SK-EC156603-2001, Installation Detail of Weldolet Service Water Pipe Repair 

SW 1802-09-EC 2080429, SW Piping Repair (Flush Patch) Line No 1-SW-1802-004 

Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion 

Procedures 

MA-AA-100-1011, Equipment Troubleshooting, Revision 0 

OS1235.03, SG Level Instrument Failure, Revision 14 

Condition Reports 

1948268 1952067 

Maintenance Orders/Work Orders 

40300038 

Miscellaneous 

Instrument Loop Diagram ILD-1-FW-L04220, Steam Generator RC-E-11B Feedwater Bypass 

 Flow (Loop 2) 1-FW-L-4220, Revision 14 

Operational Decision Making Bulletin, dated 3/18/2014 

Section 4OA5: Other Activities 

Procedures 

ER-AA-102 Underground Piping and Tank Integrity Program, Revision 6 

ER-AA-102-1000 Underground Piping and Tanks Integrity Examination Procedure, Revision 2 

Seabrook Station Underground Piping and Tanks Inspection Program, Revision 2 

SH 6.4 Dig Safe (01/06/12) Excavation of Site Locations Penetrating Plane of Ground, 

Revision 13 

Miscellaneous 

AR 00213052-01-00, Complete Initiative Action 1 Status Complete 

AR 00213052-02-00, Complete Initiative Action 2 Risk Ranking Buried Piping 

AR 00213052-03-00, Complete Initiative Action 3 Develop Inspection Plan by 06/30/11 

AR 00213052-04-00, Complete Initiative Action 4 Implement Inspection Plan 06/30/12 

AR00213052-05-00, Develop Asset Management Plan Status Complete 01/26/10 

AR00213052-06-00, Inspect Buried Piping Containing radioactive material 09/16/10 

AR00213052-07-00, Underground Piping and Tanks Procedure Oversight 12/22/11 

AR00213052-08-00, Prioritize Underground Piping and Tanks 06/27/12 

AR 01600464 (12/09/10) Aux Steam and Aux Steam Condensate Leak w/i guard pipe 

NRC Temporary Instruction 2515/182, Issue 11/17/11 and 8/8/13; Review of the Implementation 

 of the Industry Initiative to Control Degradation of Underground Piping and Tanks 

NEI 09-14 Initial Issue, November 2009 Guideline for the Management of Underground Piping 

and Tank Integrity 

NEI 09-14 Guideline for the Management of Underground Piping and Tank Integrity, Revision 1, 

December 2010 
A-10 

Attachment 

NEI 09-14 Guideline for the Management of Underground Piping and Tank Integrity Inspection 

and Analysis Methodologies, Revision 3 

LIST OF ACRONYMS 

ADAMS Agencywide Document Access and Management System 

ALARA as low as reasonably achievable 

CAP corrective action program 

CFR Code of Federal Regulations

CSSI containment spray safety injection 

EAL emergency action level 

EDG emergency diesel generator 

EFW emergency feedwater 

EPD electronic personal dosimeter 

ESFAS engineered safety features actuation system 

HRA high radiation area 

IMC Inspection Manual Chapter

LHRA locked high radiation area 

MR Maintenance Rule 

NCV non-cited violation 

NEI Nuclear Energy Institute 

NRC Nuclear Regulatory Commission 

OOS out of service 

PAB primary auxiliary building 

RCS reactor coolant system 

RG Regulatory Guide 

RHR residual heat removal 

RPM Radiation Protection Manager 

RWP radiation work permit 

SDP significance determination process 

SEPS supplemental emergency power system 

SSC structure, system, or component 

SW service water 

TI temporary instruction 

TS technical specification 

UFSAR Updated Final Safety Analysis Report 

VAC volts alternating current 

VHRA very high radiation area 

WO work order