UNITED STATES

 NUCLEAR REGULATORY COMMISSION

REGION I

2100 RENAISSANCE BLVD., SUITE 100

KING OF PRUSSIA, PA 19406-2713

 August 5, 2014

Mr. Dean Curtland 

Site Vice President

Seabrook Nuclear Power Plant 

NextEra Energy Seabrook, LLC 

c/o Mr. Michael Ossing 

P.O. Box 300 

Seabrook, NH 03874 

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

REPORT 05000443/2014003 

Dear Mr. Curtland: 

On June 30, 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 July 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.

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

These findings did not involve a violation of NRC requirements. Further, inspectors 

documented a licensee-identified violation, which was determined to be of very low safety 

significance, in this report. 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 findings 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.
D. Curtland 2

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

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

electronically for public inspection in the 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. 05000443/2014003

w/ Attachment: Supplemental Information

cc w/encl: Distribution via ListServ
D. Curtland 2

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

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

electronically for public inspection in the 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. 05000443/2014003

w/ Attachment: Supplemental Information

cc w/encl: Distribution via ListServ

Distribution w/encl: (via e-mail)

W. Dean, RA 

D. Lew, DRA 

H. Nieh, DRP

M. Scott, DRP 

P. Krohn, DRS 

J. Trapp, DRS 

G. Dentel, DRP 

R. Barkley, DRP

M. Draxton, DRP

B. Reyes, DRP

P. Cataldo, DRP, SRI

C. Newport, DRP, RI 

A. Cass, DRP, Resident AA

A. Bowers, RI, OEDO

RidsNrrPMSeabrook Resource

RidsNrrDorlLpl1-2 Resource

ROPreports Resource

 DOC NAME: G/DRP\BRANCH3\INSPECTION\REPORTS\ISSUED\2014 (ROP 15)\SB IR14003 FINAL.DOCX

 ADAMS Accession No: ML14212A458 :

 SUNSI Review

 Non-Sensitive

 Sensitive

 Publicly Available

 Non-Publicly Available

OFFICE RI/DRP RI/DRP RI/DRP 

NAME PCataldo/ MSD per telecon MDraxton/MSD GDentel/ GTD

DATE 07/31/14 07/31/14 07/31 /14

OFFICIAL RECORD COPY
1

Enclosure

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No.: 50-443

License No.: NPF-86

Report No.: 05000443/2014003

Licensee: NextEra Energy Seabrook, LLC 

Facility: Seabrook Station, Unit No.1

Location: Seabrook, New Hampshire 03874

Dates: April 1, 2014 through June 30, 2014

Inspectors: P. Cataldo, Senior Resident Inspector

C. Newport, Resident Inspector

T. O’Hara, Reactor Inspector

B. Dionne, Health Physicist

Approved by: Glenn T. Dentel, Chief

Reactor Projects Branch 3

Division of Reactor Projects
2

Enclosure

TABLE OF CONTENTS

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

REPORT DETAILS.................................................................................................................... 5

1. REACTOR SAFETY ........................................................................................................... 5

1R01 Adverse Weather Protection .................................................................................... 5

1R04 Equipment Alignment ............................................................................................... 6

1R05 Fire Protection.......................................................................................................... 7

1R06 Flood Protection Measures ...................................................................................... 8

1R07 Heat Sink Performance ............................................................................................ 9

1R08 In-service Inspection ................................................................................................ 9

1R11 Licensed Operator Requalification Program ...........................................................14

1R12 Maintenance Effectiveness .....................................................................................15

1R13 Maintenance Risk Assessments and Emergent Work Control ................................17

1R15 Operability Determinations and Functionality Assessments ....................................18

1R18 Plant Modifications .................................................................................................18

1R19 Post-Maintenance Testing.......................................................................................19

1R20 Refueling and Other Outage Activities ....................................................................19

1R22 Surveillance Testing ...............................................................................................20

2. RADIATION SAFETY.........................................................................................................21

2RS1 Radiological Hazard Assessment and Exposure Controls ......................................21

2RS2 Occupational ALARA Planning and Controls ...........................................................23

2RS3 In-Plant Airborne Radioactivity Control and Mitigation ............................................24

2RS4 Occupational Dose Assessment .............................................................................25

2RS5 Radiation Monitoring Instrumentation .....................................................................28

4. OTHER ACTIVITIES..........................................................................................................31

4OA1 Performance Indicator Verification ..........................................................................31

4OA2 Problem Identification and Resolution ....................................................................31

4OA3 Follow-Up of Events and Notices of Enforcement Discretion ..................................33

4OA6 Meetings, Including Exit...........................................................................................36

4OA7 Licensee-Identified Violation....................................................................................36

ATTACHMENT: SUPPLEMENTARY INFORMATION...............................................................37

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

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

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

LIST OF DOCUMENTS REVIEWED.......................................................................................A-2

LIST OF ACRONYMS...........................................................................................................A-21
3

Enclosure

SUMMARY

IR 05000443/2014003; 04/01/2014-06/30/2014; Seabrook Station, Unit No. 1; Maintenance 

Effectiveness and Follow-up of Events and Notices of Enforcement Discretion.

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

inspections performed by regional inspectors. Inspectors identified two findings of very low 

safety significance (Green). The significance of most findings is indicated by their color (i.e., 

greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual 

Chapter (IMC) 0609, “Significance Determination Process,” dated 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 July 9, 2013. The NRC’s program for overseeing the safe 

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

Oversight Process,” Revision 5.

Cornerstone: Mitigating Systems

 Green. The inspectors identified a finding of very low safety significance (Green) because 

NextEra did not perform adequate evaluations of safety-related residual heat removal 

(RHR) vaults. Specifically, additional technical evaluation and analysis was not adequately 

conducted on the safety-related ‘A’ and ‘B’ RHR concrete vaults when it was determined 

that they exceeded the quantitative limits specified in NextEra procedures. NextEra entered 

the failure to perform adequate technical evaluations on concrete structures exceeding 

American Concrete Institute (ACI) Tier II quantitative requirements into the CAP (action 

request (AR) 01929460), and planned to perform a formal technical evaluation of the ‘A’ 

and ‘B’ RHR vault conditions in accordance with their structural monitoring program 

procedure and the ACI 349.3R-96 standard.

The performance deficiency was considered to be 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, the inspectors 

concluded that the reliability of the structures was affected in that they exceeded the

specified Tier II limits without the performance of further technical 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 represent an actual loss of function of at least a single train for 

greater than its Technical Specification Allowed Outage Time or two separate safety 

systems out-of-service for greater than it Technical Specification Allowed Outage Time. 

This finding is related to the cross-cutting area of Human Performance – Procedure 

Adherence, because NextEra did not follow processes, procedures, and work instructions. 

Specifically, NextEra personnel did not perform an adequate technical evaluation of two 

safety-related concrete structures that exceeded the quantitative criteria requiring such an 

evaluation [H.8]. (Section 1R12)

 Green. The inspectors identified a self-revealing finding of very low safety significance

(Green), because NextEra did not ensure that adequate procedural guidance existed in 

ON1046.12, ‘Operation of the Main Generator Breaker’ to limit the likelihood of events that 

upset plant stability. Specifically, Seabrook station experienced an automatic reactor trip 

from approximately 15 percent reactor power on April 1, 2014 when two of four reactor 
4

Enclosure

coolant pumps (RCPs) tripped on low bus voltage. The cause of the reactor trip was 

directly attributable to the main generator breaker inadvertently closing and actuating the 

main generator multi-function protective relay. NextEra entered the event into their CAP, 

and conducted a root cause evaluation to determine the root and contributing causes, 

extent of condition and extent of cause, and to identify corrective actions to prevent 

recurrence. NextEra initiated actions to revise ON1046.12 to add controls regarding the 

potential risk associated with placing the main generator breaker control in local, conducted 

briefings with Maintenance groups involved in the event, and evaluated the adequacy of 

other Operations procedures that place equipment in a configuration where protective 

features are bypassed or defeated.

The performance deficiency was more than minor because it was associated with the 

procedure quality attribute of the Initiating Events cornerstone, and it adversely affected the 

cornerstone objective to limit the likelihood of events that upset plant stability and challenge 

critical safety functions during shutdown as well as power operations. The finding was 

evaluated under IMC 0609, Attachment 4, “Phase 1 – Initial Characterization of Findings.” 

The inspectors determined that the finding is of very low safety significance (Green) 

because it did not result in both a reactor trip and the loss of mitigating equipment relied 

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

finding has a cross-cutting aspect in the area of Human Performance - Work Management, 

because NextEra did not ensure that a process of planning, controlling, and executing work 

activities such that nuclear safety is the overriding priority was implemented. Specifically, 

ON1046.12, “Operation of the Main Generator Breaker” did not contain adequate 

procedural guidance regarding the impacts of positioning the Main Generator Selector 

Switch to local, take mitigating actions, and minimize time spent at increased risk 

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

Other Findings

A violation of very low safety significance that was identified by NextEra was reviewed by 

the inspectors. Corrective actions taken or planned by NextEra have been entered into 

NextEra’s corrective action program. This violation and corrective action tracking number 

are listed in Section 4OA7 of this report.
5

Enclosure

REPORT DETAILS

Summary of Plant Status

Seabrook Unit 1 began the assessment period reducing power to begin a planned refueling 

outage (OR) 16. However, at 12:26 a.m., on April 1, 2014, an unanticipated reactor trip 

occurred while the plant was at approximately 15 percent reactor power. Following the reactor 

trip, Unit 1 remained shutdown until April 24, 2014 when the main unit generator was 

successfully synchronized with offsite power, and resumed full power operation (100 percent) 

on April 28, 2014. Seabrook Unit 1 operated at essentially 100 percent power for the remainder 

of the assessment period. Documents reviewed for each section of this inspection report are 

listed in the Attachment.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01 – 3 samples)

.1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspectors performed a review of NextEra’s readiness for the onset of seasonal high 

temperatures. The review focused on the service water cooling tower, service water 

pump house, switchyard, termination yard, and the control building. The inspectors 

reviewed the Updated Final Safety Analysis Report (UFSAR), technical specifications 

(TSs), the seasonal readiness memorandum, and the corrective action program to 

determine specific temperatures or other seasonal weather that could challenge these 

systems, and to ensure NextEra personnel had adequately prepared for these 

challenges. The inspectors reviewed station procedures, including NextEra’s seasonal 

weather preparation procedure and applicable operating procedures. The inspectors 

performed walkdowns of the selected systems to ensure station personnel identified 

issues that could challenge the operability of the systems during hot weather conditions. 

b. Findings

No findings were identified.

.2 Summer Readiness of Offsite and Alternate Alternating Current (AC) Power Systems

a. Inspection Scope

The inspectors performed a review of plant features and procedures for the operation 

and continued availability of the offsite and alternate AC power system to evaluate 

readiness of the systems prior to seasonal high grid loading. The inspectors reviewed 

NextEra’s procedures affecting these areas and the communication protocols between 

the transmission system operator and NextEra. This review focused on changes to the 

established program and material condition of the offsite and alternate AC power 

equipment. The inspectors assessed whether NextEra established and implemented 

appropriate procedures and protocols to monitor and maintain availability and reliability 
6

Enclosure

of both the offsite AC power system and the onsite alternate AC power system. The 

inspectors evaluated the material condition of the associated equipment by interviewing 

the responsible system manager, reviewing condition reports (CRs) and open work 

orders, and walking down portions of the offsite and AC power systems including the 

345 kilovolt (kV) termination yard, the 345 kV switchyard, and the relay room. 

b. Findings

No findings were identified.

.3 External Flooding

a. Inspection Scope

During the period of June 16 to June 20, 2014, the inspectors performed an inspection of 

the external flood protection measures for Seabrook Station. The inspectors reviewed 

the Updated Final Safety Analysis Report (UFSAR), Chapter 2.4.2.2, which depicts the 

design flood levels and protection areas containing safety-related equipment to identify 

areas that may be affected by external flooding. The inspectors conducted a general 

site walkdown of the outside area of the site, fuel storage building, the control building, 

and the emergency diesel generator building to ensure that NextEra erected flood 

protection measures in accordance with design specifications. The inspectors also 

reviewed operating procedures for mitigating external flooding during severe weather to 

determine if NextEra planned or established adequate measures to protect against 

external flooding events.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial System Walkdowns (71111.04Q – 4 samples)

a. Inspection Scope

The inspectors performed partial walkdowns of the following systems:

 Containment penetration return to service on April 17, 2014

 'A' residual heal removal (RHR) return to service on April 23, 2014 

 ‘B' emergency feedwater (EFW) pump return to service on June 19, 2014

 'B' service water return to service on June 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 UFSAR, TSs, work orders 

(WOs), CRs, 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 
7

Enclosure

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 corrective action program (CAP) 

for resolution with the appropriate significance characterization. 

b. Findings

No findings were identified.

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

a. Inspection Scope

On April 17, 2014, the inspectors performed a complete system walkdown of accessible 

portions of the service water system to verify the existing equipment lineup was correct. 

The inspectors reviewed operating procedures, surveillance tests, drawings, equipment 

line-up check-off lists, and the UFSAR to verify the system was aligned to perform its 

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

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

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

portions of the systems to verify system components and support equipment were 

aligned correctly and operable. The inspectors examined the material condition of the 

components and observed operating parameters of equipment to verify that there were 

no deficiencies. Additionally, the inspectors reviewed a sample of related CRs and work 

orders to ensure NextEra appropriately evaluated and resolved any deficiencies.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Resident Inspector Quarterly Walkdowns (71111.05Q – 5 samples)

a. Inspection Scope

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

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

NextEra controlled combustible materials and ignition sources in accordance with 

administrative procedures. The inspectors verified that fire protection and suppression 

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

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

station personnel implemented compensatory measures for out of service (OOS), 

degraded, or inoperable fire protection equipment, as applicable, in accordance with 

procedures. 

 Containment Building C-F-1-Z, C-F-2-Z, & C-F-3-Z on April 7, 2014

 Emergency Feedwater Pump House 27'-0" EFP-F-1-A on May 30, 2014

 Primary Auxiliary Building 53' & 81' PAB-F-3A-Z, PAB-F-3B-Z, PAB-F-4-Z on June 2, 

2014
8

Enclosure

 Site Plan/Hydrant Locations PLT-F-1-0 on June 16, 2014

 Containment Enclosure Ventilation Area, CE-F-1-A on June 23, 2014

.2 Fire Protection – Fire Brigade Response on April 16, 2014 (71111.05A – 1 sample)

a. Inspection Scope

The inspectors observed and evaluated control room operator and fire brigade response 

to fire alarms within containment on April 16, 2014, that involved flames, sparks, and 

smoke associated with the terminal box for the “A” reactor coolant pump. The inspectors 

verified that NextEra personnel identified deficiencies and took appropriate corrective 

actions as required. The inspectors evaluated specific attributes as follows:

 Proper wearing of turnout gear and self-contained breathing apparatus

 Sufficient fire-fighting equipment brought to the scene

 Effectiveness of command and control

 Communications established between fire brigade and control room

 Propagation of the fire into other plant areas

 Smoke removal operations

 Utilization of pre-planned strategies

 Procedure adherence regarding fire alarm response, fire brigade dispatch, fire alarm 

activation, containment evacuation alarm initiation, and assessment of emergency 

plan actions

b. Findings

No findings were identified.

1R06 Flood Protection Measures (71111.06 – 2 samples)

.1 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 

control 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.
9

Enclosure

.2 Annual Review of Cables Located in Underground Bunkers/Manholes

a. Inspection Scope

The inspectors conducted an inspection of two samples of underground 

bunkers/manholes subject to flooding that contain cables whose failure could affect risksignificant equipment. The inspectors performed walkdowns of risk-significant areas, 

specifically manholes No. 5 and No. 11, which contain safety-related cables routed to 

the service water pumphouse from essential switchgear. The inspectors verified water 

level in the sump/manhole to ensure that the cables were not submerged. The 

inspectors verified that the bunkers/manholes were dewatered in accordance with 

station procedures.

b. Findings

No findings were identified.

1R07 Heat Sink Performance (71111.07A – 1 sample)

a. Inspection Scope

The inspectors reviewed the ‘B’ emergency diesel generator (EDG) jacket water heat 

exchanger to determine its readiness and availability to perform its safety functions. 

The inspectors reviewed the design basis for the component and verified NextEra’s 

commitments to NRC Generic Letter 89-13. The inspectors observed actual 

performance tests for the heat exchangers and/or reviewed the results of previous 

inspections of the ‘B’ EDG jacket water and similar heat exchangers. The inspectors 

discussed the results of the most recent inspection with engineering staff and reviewed 

pictures of the as-found and as-left conditions. The inspectors verified that NextEra 

initiated appropriate corrective actions for identified deficiencies. The inspectors also 

verified that the number of tubes plugged within the heat exchanger did not exceed the 

maximum amount allowed.

b. Findings

No findings were identified.

1R08 In-service Inspection (71111.08 – 1 sample)

a. Inspection Scope

From April 4 to April 11, 2014 and from April 13 to April 16, 2014, the inspectors 

conducted a review of NextEra’s implementation of in-service inspection (ISI) program 

activities for Seabrook Unit 1. These activities monitor the reactor coolant system 

pressure boundary, risk significant piping and components and the containment to 

identify degradation, complete evaluations and make repairs or replacements as 

required. Sample selection was based on the inspection procedure objectives and risk 

priority of those pressure retaining components in systems where degradation could 

result in a significant increase in risk. The inspectors observed in-process nondestructive examinations (NDE), reviewed documentation records, and interviewed 

inspection personnel to verify that the non-destructive examination activities performed 
10

Enclosure

as part of Period 2 of the third 10-year interval of the Seabrook ISI Program, during 

refueling outage OR16, were conducted in accordance with the requirements of the 

ASME Boiler and Pressure Vessel Code Section XI, 2004 Edition, No Addenda.

Nondestructive Examination (NDE) and Welding Activities

ASME Code-Required Examinations

The inspectors reviewed the equipment calibration sheet and the inspection data sheets 

from the ultrasonic testing (UT) examination of the pressurizer nozzle inner radius weld 

(RC E 10 A IR). There were no recordable indications from this examination and 100 

percent coverage was achieved. 

The inspectors reviewed the equipment calibration sheet and inspection data sheets 

from the UT examination of the pressurizer nozzle weld (RC E-10 A-NZ). There were no 

recordable indications from this examination; however coverage was limited to 74 

percent due to the nozzle configuration.

The inspectors reviewed the equipment calibration sheet and the inspection data sheets 

from the UT examination of the Steam Generator (SG) ‘A’ nozzle to pipe weld (RC E 

11A 2B NZ). There were no recordable indications from this examination and 100 

percent coverage was achieved.

The inspectors reviewed the equipment calibration sheet and inspection data sheets 

from the UT examination of the pressurizer nozzle weld (RC E-10 2A-NZ). There were 

no recordable indications from this examination; however coverage was limited to 98 

percent due to the nozzle configuration.

The inspectors reviewed the equipment calibration sheet and the inspection data sheets 

from the UT examination of the SG ‘A’ pipe to nozzle weld inner radius (RC E 11 2A-IR). 

There were no recordable indications from this examination and 100 percent coverage 

was achieved.

The inspectors reviewed the equipment calibration sheet and the inspection data sheets 

from the UT examination of the SG ‘A’ nozzle to pipe weld inner radius (RC E-11A 2B 

IR). There were no recordable indications from this examination and 98 percent 

coverage was achieved. 

The inspectors reviewed the equipment calibration sheets and the inspection data 

sheets from the UT examination of the pressurizer nozzle weld (RC E-11 2B-NZ). There 

were no recordable indications from this examination and 98 percent coverage was 

achieved. 

The inspectors reviewed the inspection data sheet from the bare metal visual 

examination of the reactor vessel closure head (RVCH) and control rod drive mechanism 

(CRDM) nozzle penetrations performed in accordance with ASME Code Case 792-1. 

The inspectors reviewed certifications of the NDE technicians performing these 

examinations and verified that the inspections were performed in accordance with 

approved procedures and that the results were reviewed and evaluated by certified 

Level III NDE personnel.
11

Enclosure

Other Containment Liner Examinations

The inspectors reviewed additional UT results completed by NextEra to examine the 

containment liner. NextEra performed ultrasonic thickness measurements of 51, onesquare foot area sample locations of the containment liner. These measurements were 

taken from the inside of the containment at elevation -26 feet and were spaced 

approximately every 6 degrees in azimuth around the containment circumference in 

accessible areas. The containment liner thicknesses were measured to be nominally 

0.375 inch in thickness at all locations, and no exterior or interior surface degradation of 

the liner was evident.

The inspectors reviewed the visual testing (VT) procedures and the recorded inspection 

results for the inspection of an anomaly on the upper dome portion of the containment 

liner. The inspectors determined these inspections were conducted in accordance with 

the ASME Code Section XI, Subsection IWE. An engineering evaluation of the 

inspection results concluded that no observable change had occurred in the anomaly 

and recommended re-inspection during the next refueling outage, OR17.

The inspectors reviewed the records of a water intrusion condition in the fuel transfer 

tube vault which had been identified in 2009. One wall of the fuel transfer tube vault is 

also part of the containment liner. In April 2011, NextEra staff made an entry into the 

vault area and conducted an inspection of the liner condition inside the vault. Some 

areas of the containment liner portion of the fuel tube vault had minor, light rust which 

was removed. UT measurements of the liner thickness in 2011 verified that all 

previously rusted areas on the liner were greater than 0.375 inch. Also, NextEra staff 

made repairs to the non-liner portion of the vault wall which had shown signs of leakage. 

Because this repair was not on the containment liner portion of the vault, no repair has 

been made under the ASME Code, Section IWE in the fuel transfer vault. The 

inspectors determined there were no examinations completed during OR16. An 

examination was not required because the leak was stopped in 2011 and a repair was 

not made to the containment liner.

Review of Originally Rejectable Indications Accepted by Evaluation

The inspectors reviewed a volumetric examination data sheet for an indication identified 

during a previous (OR15) inspection of a piping weld in line SI- 0251-07-09. A flaw 

evaluation completed by a qualified NDE Level III individual later determined that the 

flaw was acceptable per ASME Code, Subsection XI, Table IWB-3514-2. 

The inspectors reviewed the NDE report of a weld flaw detected in the J-groove weld 

during the ultrasonic (UT) examination of RVCH penetration No. 57. These inspections 

are required by 10 CFR 50.55(a) and Code Case N-729-1. NextEra staff performed a 

manual eddy current (EC) inspection to determine if the indication was fully contained 

within the weld. The EC exam showed that the indication was completely contained by 

the weld material and was acceptable per the requirements of the ASME Code without 

repair.

Repair/Replacement Activities

The inspectors reviewed WO package 40108170-01 for replacement and testing 

activities of pressure relief valve SI-V-60, the ‘D’ accumulator relief valve. The 

inspectors verified that the replacement valve was successfully tested with a leak test 
12

Enclosure

pressure of 630.5 psig. The inspectors also reviewed the VT-2 visual examination data 

sheet for SI-V-60, and the Form A of the Flanged Joint Torque Traveler for the valve. 

The inspectors also reviewed the Field Work Closeout Form and the ASME Form NIS2A Repair/Replacement Certification Record. The inspectors determined that the 

replacement was completed in accordance with ASME Section XI, Repair/Replacement 

requirements. 

The inspectors reviewed WO 40220458-03 associated with the shop fabrication of piping 

to replace the service water pump P-41 B/D discharge piping with AL6XN piping material 

per engineering change EC278717. The inspectors reviewed the ASME Section XI 

Repair/Replacement Plan Traveler, the Form A Weld Traveler, and the liquid penetrant 

test records to verify that the replacement piping had been fabricated in accordance with 

the change document. The inspectors verified that the Repair/Replacement Plan had 

been completed correctly in accordance with the ASME Section XI Code.

PWR Vessel Upper Head Penetration (VUHP) Inspection Activities

The inspectors reviewed the NextEra calculations of Effective Degradation Years (EDY) 

and Re-inspection Year (RIY) for the Seabrook Unit 1 Reactor Vessel Upper Head 

completed prior to OR16. Based on these calculation results, NextEra completed an 

inspection of the Unit 1 VUHP J-groove welds during the OR16 refueling outage. 

These calculations were performed in accordance with the requirements of 10 CFR 

50.55a(g)(6)(ii)(D) and the ASME Boiler and Pressure Vessel Code Case N 729 1, 

“Alternative Examination Requirements for PWR Reactor Vessel Upper Heads,” to 

ensure the structural integrity of the reactor vessel head pressure boundary. 

The inspectors reviewed visual inspection reports (Data Sheets) of the remote bare 

metal visual examination of the exterior surface of Unit 1 VUHP to confirm appropriate 

inspection coverage was achieved and to verify that no boric acid leakage or wastage 

had occurred.

The inspectors observed the ultrasonic examination of several RVCH penetrations and 

reviewed the data from the inspection of penetration No. 57 which identified a recordable 

indication in the J-groove weld. The inspectors reviewed the Eddy Current (ET) 

inspection data used to verify the location of the indication. This inspection verified that 

the indication was enclosed within the J-groove weld and could be accepted for 

continued use without repair. There were no other recordable indications detected in the 

other reactor vessel upper head nozzle penetrations.

Boric Acid Corrosion Control (BACC) Inspection Activities 

The inspectors reviewed the boric acid corrosion control program, which is conducted 

in accordance with Seabrook Unit 1 Station procedures and NextEra procedures, 

discussed the program with the boric acid program owner, and sampled photographic 

inspection records of boric acid leaks found on safety significant piping and components 

inside the Seabrook Unit 1 containment. The walkdowns were conducted by NextEra 

personnel and were directly observed by the NRC Resident Inspectors on April 1, 2014, 

during the initial containment entry for the OR16 outage. The inspectors reviewed a 

sample of leaks observed and reported, the identification and documentation of nonconforming conditions identified in the corrective action program and reviewed a sample 

of boric acid evaluations completed by engineering to repair or monitor the conditions.
13

Enclosure

The inspectors verified that potential deficiencies identified during the walkdowns were 

entered into NextEra’s corrective action program and reviewed a sample of engineering 

evaluations of the conditions reported to verify that the corrective actions were 

consistent with the requirements of the NextEra procedures and 10 CFR 50, Appendix 

B, Criterion XVI.

Steam Generator (SG) Tube Inspection Activities

The inspectors observed a portion of SG eddy current testing (ECT) examinations, data 

evaluation, and documentation review of the final list of pluggable tubes from this 

inspection.

The inspectors remotely observed a sample of the Unit 1 SG eddy current tube 

examinations which were based upon Seabrook's operating experience and the 

assessment of past degradation mechanisms. 

The scope consisted of the following examinations:

 100 percent bobbin coil probe examination of SG 'B'

 20 percent bobbin coil probe examination of SG 'A', SG 'C', and SG 'D', including 

Outside Diameter Stress Corrosion Cracking (ODSCC) susceptible tubes, and tubes 

susceptible to tube-to-tube wear

 100 percent +Point examination of dings/dents > 5 volts on HL (including U-Bend) 

in SG 'C'

 20 percent +Point examination of dings/dents > 5 volts on HL (including U-bend) 

in SG 'A', SG 'B', and SG 'D'

 Resolution of all bobbin coil probe "I" codes

 Visual inspection of all plugs (mechanical and welded)

 Visual inspection of channel heads in all SG's

The inspectors reviewed a sample of the indications identified in the SGs during the 

OR15 eddy current inspections to verify that they were consistent with the potential 

degradation mechanisms that may be observed during the current OR16 inspections 

as documented in SG-SGMP-13-21, Revision 3, “Steam Generator Degradation 

Assessment for the Seabrook OR16 Refueling Outage.” The inspectors also reviewed 

the Condition Monitoring and Operational Assessments from the prior outage, OR15.

The inspectors verified that the SG eddy current tube examinations were performed 

in accordance with Unit 1 Technical Specifications and the Unit 1 Steam Generator 

Program by reviewing the SG tube eddy current test results to verify that no in-situ 

pressure testing was required, no tubes required stabilization, and no increased primaryto-secondary leakage had occurred over the operating cycle.

The inspector also verified that the tubes that exhibited degradation and did not meet 

acceptance criteria were plugged (3 tubes, due to anti-vibration bar wear) or sleeved (0 

tubes) using the alternate repair criteria per Generic Letter 95-05, Voltage-Based Repair 

Criteria for Westinghouse Steam Generator Tubes Affected by ODSCC during the OR16 

inspection. 
14

Enclosure

The inspectors verified that SG tubes that exhibited degradation examination screening 

criteria was in accordance with the Electric Power Research Institute (EPRI) Steam 

Generator Guidelines and flaw sizing was in accordance with EPRI examination 

technique specification sheet. During OR16, NextEra plugged three tubes in ‘C’ SG due 

to anti-vibration bar wear. The total tubes plugged in all four SG's, at the completion of 

OR16, was 185. 

NextEra staff did not conduct foreign object search and retrieval (FOSAR) on the 

secondary side of the SGs during OR16, because the secondary side of the SG's were 

not opened and inspected during this outage.

The inspectors reviewed the Westinghouse eddy current testing procedures and verified 

that NextEra completed the steam generator inspections in accordance with the 

requirements of NEI 97-06, Pressurized Water Reactor Steam Generator Examination 

Guidelines, Revision 7. The inspectors also reviewed the Westinghouse procedure 

qualification certifications and a sample of data analysts' personnel certifications. The 

inspectors reviewed a sample of eddy current qualification records for Primary and 

Secondary resolution analysts, Independent Quality Data Analysts, and Utility Level III 

Quality Data Analysts.

Identification and Resolution of Problems

The inspectors reviewed a sample of Seabrook Station Unit 1 condition reports, which 

identified NDE indications, deficiencies and other non-conforming conditions since the 

previous OR15 outage and during the current OR16 outage. The inspectors verified that 

nonconforming conditions were properly identified, characterized, evaluated and entered 

into the corrective action program.

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 June 25, 2014, which 

included a loss of coolant accident coincident with a loss of off-site power. 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 shift technical 

advisor. Additionally, the inspectors assessed the ability of the crew and training staff to 

identify and document crew performance problems.
15

Enclosure

b. Findings

No findings were identified.

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

a. Inspection Scope

The inspectors observed licensed operator performance in the main control room during 

plant shutdown in preparation for a planned refueling outage (OR16), and observed 

operations staff event response to a reactor trip and transition to Mode 3 on April 1, 

2014. Additionally, inspectors observed containment fire response on April 16, 2014; 

initial reactor coolant system (RCS) draindown and heatup activities on April 4 and 21, 

2014; approach to criticality and transition to Mode 2 on April 23, 2014; control room 

turnover on June 10, 2014; and RCS boration and turbine control system reboot on 

June 12, 2014. The inspectors observed applicable 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, and component (SSC) performance and 

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

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

NextEra was identifying and properly evaluating performance problems within the scope 

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

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

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

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

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

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

across MR system boundaries.

 4000-RM-10 Snubber failure on April 11, 2014

 Structures monitoring program - RHR vaults on May 23, 2014

b. Findings

Introduction: The inspectors identified a finding of very low safety significance (Green) 

because NextEra did not perform adequate evaluations of safety-related RHR vaults. 

Specifically, additional technical evaluation and analysis was not adequately conducted 

on the safety-related ‘A’ and ‘B’ RHR concrete vaults when it was determined that they 

exceeded the quantitative limits specified in NextEra procedures. 
16

Enclosure

Description: NextEra’s Engineering Department Standard 36180, “Structural Monitoring 

Program” provides guidance for the conduct of the structural condition monitoring 

program to meet the requirements of 10CFR 50.65, the Maintenance Rule. The 

procedure provides a systematic approach for evaluation of plant structures to provide 

reasonable assurance that those structures are capable of fulfilling their intended safety 

function. This is accomplished, in part, by periodic reviews of the condition of plant 

structures via systematic walkdowns. Additionally, NextEra’s structural monitoring 

program commits to the requirements specified in the American Concrete Institute 

standard ACI 349.3R-96, ‘Evaluation of Existing Nuclear Safety Related Concrete 

Structures’ for the evaluation of conditions identified in safety-related concrete 

structures. This commitment was formally implemented by procedure revision on 

July 29, 2013. 

NextEra’s structural monitoring program procedure states that measurable 

discontinuities exceeding specified ACI ‘Tier II’ quantitative limits shall be considered 

unacceptable and in need of further technical evaluation and that further evaluation 

should consider the use of other inspection, testing or analytical tools to obtain condition 

and functional information of the structures in question. 

While performing a plant walkdown on May 23, 2014, the inspectors identified several 

instances of concrete conditions in the ‘A’ and ‘B’ RHR vaults that exceeded the 

quantitative Tier II criteria specified in NextEra’s procedure. These conditions included: 

spalling greater than 20 millimeter (mm) in depth, passive cracks greater than 1 mm, 

and staining of an undefined source on concrete surfaces. The load bearing walls of 

the RHR vaults are approximately 30 inches thick and are reinforced with number 

8 vertical reinforcing bars spaced at 12 inches each face and number 9 horizontal 

reinforcing bars spaced at 9 inches each face.

The inspectors reviewed NextEra’s routine structures monitoring program walkdown 

evaluations conducted on the ‘A’ RHR vault on October 10, 2013 and the ‘B’ RHR vault 

on March 25, 2014. The inspectors noted that NextEra personnel had identified the 

conditions, but had not conducted adequate technical evaluation of those conditions. 

Specifically, NextEra’s analysis of the conditions in the RHR vaults relied largely on 

engineering judgment and did not use any other inspection, testing or analytical tools to 

obtain additional condition and functional information of the two vaults.

The inspectors consulted with regional specialists, reviewed a subsequent licensee 

evaluation of the RHR vaults, and concluded that the RHR vaults remain operable and 

have reasonable assurance of structural integrity. This conclusion is based on the 

significant reduction in crack width below the concrete cover, and the absence of 

degradation or distress in the concrete surrounding the cracks. The observed cracks 

do not show any indications that they are alkali-silica reaction (ASR) related/induced 

cracks Rather, the cracking is more likely due to shrinkage and/or settlement-induced 

stress relief. Additional engineering review and analysis is warranted to more clearly 

identify and understand the cause(s) of the observed cracking in the RHR vaults. 

NextEra entered the failure to perform adequate technical evaluations on concrete 

structures exceeding Tier II quantitative requirements into the Corrective Action Program 

(AR 01929460), and planned to perform a formal technical evaluation of the ‘A’ and ‘B’ 

RHR vault conditions in accordance with their structural monitoring program procedure 

and the ACI 349.3R-96 standard.
17

Enclosure

Analysis: The inspectors determined that NextEra’s failure to perform an adequate 

technical evaluation of conditions identified in the ‘A’ and ‘B’ RHR vaults that exceeded 

the Tier II quantitative criteria in the structures monitoring program procedure was a 

performance deficiency within NextEra’s ability to foresee and correct. The performance 

deficiency was considered to be 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, the inspectors concluded that 

the reliability of the structures was affected in that they exceeded the specified Tier II 

limits without the performance of further technical 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 represent an actual loss of function of at least a single train for greater 

than its TS Allowed Outage Time or two separate safety systems out-of-service for 

greater than its TS Allowed Outage Time. This finding is related to the cross-cutting 

area of Human Performance – Procedure Adherence, because NextEra did not follow 

processes, procedures, and work instructions. Specifically, NextEra personnel did not 

perform an adequate technical evaluation of two safety-related concrete structures that 

exceeded the quantitative criteria requiring such an evaluation [H.8]. 

Enforcement: Enforcement action does not apply because the performance deficiency 

did not involve a violation of a regulatory requirement. (FIN 05000443/2014003-01, 

Inadequate Technical Evaluation of Safety-Related Structures)

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 – 6 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 

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.

 Lowering of reactor water level for reactor vessel head removal on April 3, 2014

 Unit Auxiliary Transformers (UATs) and RHR OOS for maintenance on April 8, 2014

 Mid-loop operations for steam generator nozzle dam removal on April 13, 2014

 Risk assessment for 1-ED-X-2-A/B OOS for main generator breaker repairs while 

entering Mode 4 from Mode 5, Mode 3 from Mode 4, and Mode 2 from Mode 3 on 

April 22, 2014

 'B' containment instrument air compressor corrective maintenance on June 2, 2014

 Battery charger return to service on June 19, 2014
18

Enclosure

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments (71111.15 – 4 samples)

a. Inspection Scope

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

 ‘A’ EFW past operability due to oil leak identified on March 19, 2014

 Vital inverter EDE-1B increased DC input amps on May 25, 2014

 RCP undervoltage (UV) time delay relays out of tolerance on June 9, 2014

 Seat leakage on 1-MS-V-393 on June 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

No findings were identified.

1R18 Plant Modifications (71111.18 – 1 sample)

Temporary Modifications

a. Inspection Scope

The inspectors reviewed the temporary modifications listed below to determine whether 

the modifications affected the safety functions of systems that are important to safety. 

The inspectors reviewed 10 CFR 50.59 documentation and post-modification testing 

results, and conducted field walkdowns of the modifications to verify that the temporary 

modifications did not degrade the design bases, licensing bases, and performance 

capability of the affected systems. In addition, the inspectors reviewed modification 

documents associated with the upgrade, including associated engineering changes, 

correspondence with the vendor, industry operating experience, environmental and 

seismic qualifications, as well as the 10 CFR 50.59 documentation and post-modification 

testing results, as applicable.

 Engineering Change 250048, main feed pump turbine digital upgrade
19

Enclosure

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.

 CBS-V2 following electrical maintenance on valve motor, on April 1, 2014

 RHR pump 8B comprehensive test following maintenance on April 14, 2014

 Retest for strainer 11 bypass leak on April 15, 2014

 BC-1A energize and parallel following maintenance on May 22, 2014

 CS-FCV-111A actuator diaphragm replacement on June 11, 2014

 ‘B’ service water train comprehensive test on June 25, 2014

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities (71111.20 – 1 sample)

a. Inspection Scope 

The inspectors reviewed the station’s work schedule and outage risk plan for the 

maintenance and refueling outage (OR16), which was conducted April 1 to 22, 2014. 

The inspectors reviewed NextEra’s development and implementation of outage plans 

and schedules to verify that risk, industry experience, previous site-specific problems, 

and defense-in-depth were considered. During the outage, the inspectors observed 

and evaluated the following outage activities:

 Shutdown and cooldown operations, and transition to Mode 5 and entry into RHR 

operations

 Configuration management, including maintenance of defense-in-depth, 

commensurate with the outage plan for the key safety functions and compliance with 

the applicable technical specifications when taking equipment out of service

 Implementation of clearance activities and confirmation that tags were properly hung 

and that equipment was appropriately configured to safely support the associated 

work or testing

 Installation and configuration of reactor coolant pressure, level, and temperature 

instruments to provide accurate indication and instrument error accounting 
20

Enclosure

 Status and configuration of electrical systems and switchyard activities to ensure that 

technical specifications were met

 Monitoring of decay heat removal (RHR) operations

 Impact of outage work on the ability of the operators to operate the spent fuel pool 

cooling system

 Reactor water inventory controls, including flow paths, configurations, alternative 

means for inventory additions, and controls to prevent inventory loss

 Activities that could affect reactivity , including fuel handling and receipt inspections 

 Fatigue management

 Tracking of startup prerequisites, including mode transition reviews, walkdown of the 

primary containment to verify that debris had not been left which could block the 

emergency core cooling system suction strainers, and startup and ascension to full 

power operation

 Reactor start-up and plant heat-up activities

 Identification and resolution of problems related to refueling outage activities

 Consistent with TI 2515/188: Inspection of Near-Term Task Force Recommendation 

2.3 – Seismic Walkdowns, reviewed deferred inspections due to inaccessibility from 

seismic walkdowns conducted in 2012. These activities were performed on: 4160-

volt safety bus No. 6, located in the ‘B’ essential switchgear room, and 480-volt 

buses E63 and E64, located in containment.

b. Findings

No findings were identified.

1R22 Surveillance Testing (71111.22 – 9 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:

 Reactor vent paths cold shutdown and 18 month surveillance on April 1, 2014

 CBS-V-17 and CBS-V-18 local leakage rate tests on April 6, 2014 (containment 

isolation valve)

 Containment enclosure emergency exhaust filter system 18 month surveillance on 

April 9, 2014 (containment isolation valve)

 RC-V-88 and RC-V-89 local leakage rate tests on April 9, 2014

 Phase ‘B’ CBS and CVI actuation 18 month surveillance on April 10, 2014

 EFW comprehensive flow test on April 23, 2014

 Containment personnel hatch door seal leakage test on May 21, 2014 (inservice 

testing)

 Primary system sample on June 24, 2014

 RCS steady state leak rate calculation on June 28, 2014 (RCS leakage detection)
21

Enclosure

b. Findings

No findings were identified.

2. RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01 – 1 sample)

a. Inspection Scope

During April 14 to 17, 2014, the inspectors reviewed NextEra performance in assessing 

the radiological hazards and exposure control for OR16. 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 NextEra 

procedures required by TSs as criteria for determining compliance. 

Radiological Hazard Assessment

The inspectors reviewed the last two radiological surveys from the primary channel 

heads in the four steam generators and from the reactor cavity. The inspectors 

evaluated whether the thoroughness and frequency of the surveys were appropriate 

for the given radiological hazard.

The inspectors selected the following radiologically-significant work activities: 

 Steam Generator Eddy Current Testing and Tube Plugging

 Reactor Cavity Work during OR16, Includes Reactor Head Lift/Set

 Replace Reactor Head O-ring

The inspectors evaluated whether continuous air monitors (CAMs) were located in areas 

that were representative of actual work areas. The inspectors evaluated the NextEra 

program for monitoring levels of loose surface contamination in areas of the plant.

The inspectors reviewed several radiation work permits (RWP) used to access locked 

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

control barriers were consistent with TS requirements for LHRA.

The inspectors observed the Access Control Point location where NextEra monitors 

material leaving the radiological control area and inspected the methods used for 

control, survey, and release of these materials. The inspectors observed the 

performance of personnel surveying and releasing material for unrestricted use and 

evaluated whether the work was performed in accordance with plant procedures. 

The inspectors assessed whether the radiation monitoring instrumentation used for 

equipment release and personnel contamination surveys had appropriate sensitivity 

for the contamination present.

The inspectors evaluated the adequacy of radiological controls, required surveys, 

radiation protection job coverage, and contamination controls. The inspectors evaluated 
22

Enclosure

NextEra‘s use of electronic personal dosimeters (EPDs) in high noise areas that were 

also high radiation areas (HRAs). 

The inspectors assessed whether radiation monitoring devices were placed on the 

individual’s body consistent with NextEra procedures. The inspectors assessed whether 

the dosimeter was placed in the location of highest expected dose or that NextEra 

properly implemented an NRC-approved method of determining effective dose 

equivalent. The inspectors reviewed the application of dosimetry to effectively monitor 

exposure to personnel in high-radiation work areas with significant dose rate gradients.

Instructions to Workers

The inspectors reviewed several RWPs for work within airborne radioactivity areas with 

the potential for individual worker internal exposures.

For these RWPs, the inspectors evaluated airborne radioactive controls and monitoring, 

including potential for significant airborne levels. The inspectors assessed applicable 

containment barrier integrity and the operation of temporary high-efficiency particulate 

air ventilation systems.

Radiological Hazards Control and Work Coverage

The inspectors discussed with first-line health physics supervisors the controls in place 

for special areas that have the potential to become very high radiation areas (VHRAs) 

during certain plant operations. The inspectors assessed whether these plant 

operations require communication beforehand with the health physics group, so as to 

allow corresponding timely actions to properly control and monitor the radiation hazards.

Radiation Worker Performance

The inspectors observed the performance of radiation workers with respect to stated 

radiation protection (RP) work requirements. The inspectors assessed whether workers 

were aware of the radiological conditions in their workplace and the RWP controls/limits 

in place, and whether their behavior reflected the level of radiological hazards present.

RP Technician Proficiency

The inspectors observed the performance of the RP technicians with respect to 

controlling radiation work. The inspectors evaluated whether technicians were aware of 

the radiological conditions in their workplace and the RWP controls/limits, and whether 

their behavior was consistent with their training and qualifications with respect to the 

radiological hazards and work activities.

The inspectors reviewed two radiological problem reports since the last inspection that 

attributed the cause of the event to RP technician error. The inspectors evaluated 

whether there was an observable pattern traceable to a similar cause. The inspectors 

assessed whether this perspective matched the corrective action approach taken by 

NextEra to resolve the reported problems.
23

Enclosure

b. Findings

No findings were identified.

2RS2 Occupational ALARA Planning and Controls (71124.02 – 1 sample)

a. Inspection Scope

During April 14 to 17, 2014, the inspectors assessed performance with respect to 

maintaining occupational individual and collective radiation exposures as low as is 

reasonably achievable (ALARA) for OR16. 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. 

Radiological Work Planning

The inspectors selected the following work activities that had the highest exposure 

significance.

 Pre-Job ALARA Review 14-02 OR16 Steam Generator Eddy Current Testing and 

Tube Plugging

 Pre-Job ALARA Review 14-03 OR16 In Service Inspection

 Pre-Job ALARA Review 14-07 OR16 Fuel Handling Project

 Pre-Job ALARA Review 14-09 OR16 RCP Seal Replacement

 Pre-Job ALARA Review 14-10 OR16 Scaffolding

 Pre-Job ALARA Review 14-13 Replace Reactor Ventillation Ducting Under Vessel 

with New Design

The inspectors reviewed the ALARA work activity evaluations, exposure estimates, 

and exposure reduction requirements. The inspectors determined whether NextEra 

reasonably grouped the radiological work into work activities, based on historical 

precedence and industry standards. The inspectors compared the results achieved 

(actual dose) with the intended dose for these work activities. The inspectors compared 

the person-hour estimates provided by maintenance planning and other groups to the 

RP group actual person-hours for the work activity, and evaluated the accuracy of these 

time estimates. The inspectors assessed the reasons for any inconsistencies between 

intended and actual work activity doses.

Verification of Dose Estimates and Exposure Tracking Systems

The inspectors reviewed the assumptions and basis for the collective dose estimates for 

routine operations and the refueling outage. The inspectors reviewed applicable 

procedures to determine the methodology for estimating exposures from specific work 

activities and for department and station collective dose goals.

The inspectors evaluated whether the licensee had established measures to track, trend, 

and to reduce occupational doses for ongoing work activities. The inspectors assessed 
24

Enclosure

whether dose threshold criteria were established to prompt additional ALARA planning 

and controls. 

The inspectors evaluated the licensee’s method of adjusting exposure estimates, or 

re-planning work, when unexpected changes in scope or emergent work were 

encountered. The inspectors assessed whether adjustments to exposure estimates 

were based on sound RP and ALARA principles or if they were just adjusted to account 

for failures to plan/control the work. 

Radiation Worker Performance

The inspectors observed radiation worker and RP technician performance during work 

activities being performed in radiation areas, airborne radioactivity areas, and LHRAs. 

The inspectors evaluated whether workers demonstrated the ALARA philosophy in 

practice and whether there were any procedure or RWP compliance issues.

b. Findings

No findings were identified.

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

a. Inspection Scope

During April 14 to 17, 2014, the inspectors verified whether in-plant airborne 

concentrations were being controlled consistent with ALARA principles and the use 

of respiratory protection devices on-site did not pose an undue risk to the wearer. 

The inspectors used the requirements in 10 CFR Part 20, the guidance in RG 8.15 

Acceptable Programs for Respiratory Protection, RG 8.25 Air Sampling in the 

Workplace, NUREG-0041 Manual of Respiratory Protection Against Airborne 

Radioactive Material, TSs, and NextEra procedures required by TSs as criteria for 

determining compliance. 

Inspection Planning

The inspectors reviewed the Updated Final Safety Analysis Report to identify areas of 

the plant designed as potential airborne radiation areas and any associated ventilation 

systems or airborne monitoring instrumentation. This review included instruments used 

to identify changing airborne radiological conditions. The inspectors reviewed the 

respiratory protection program and a description of the types of respiratory protection 

devices used. The inspectors reviewed the procedures for maintenance, inspection, and 

use of respiratory protection equipment including self-contained breathing apparatus, as 

well as, procedures for air quality maintenance. The inspectors reviewed reported 

performance indicators to identify any related to unintended dose resulting from intakes 

of radioactive material.

Engineering Controls

The inspectors reviewed the licensee’s use of permanent and temporary ventilation to 

determine whether the licensee uses ventilation systems as part of its engineering 

controls to control airborne radioactivity. The inspectors reviewed procedural guidance 
25

Enclosure

for use of installed plant systems to reduce dose and assessed whether the systems are 

used during high-risk activities. 

The inspectors selected two temporary ventilation system used to support work in 

contaminated areas. The inspectors assessed whether the use of these systems was 

consistent with NextEra procedural guidance and ALARA. The inspectors assessed 

whether the licensee had established threshold criteria for evaluating levels of airborne 

beta-emitting and alpha-emitting radionuclides.

Use of Respiratory Protection Devices

The inspectors selected two work activities where respiratory protection devices were 

used to limit the intake of radioactive materials, and assessed whether the licensee 

performed an engineering evaluation concluding that the use of respirators is not 

required based on ALARA. The inspectors also evaluated whether the licensee had 

established means (such as routine bioassay) to determine if the level of protection 

provided by the respiratory protection devices during use was at least as good as that 

assumed in the licensee’s work controls and dose assessment.

The inspectors assessed whether respiratory protection devices used to limit the intake 

of radioactive materials was certified by the National Institute for Occupational Safety 

and Health/Mine Safety and Health Administration (NIOSH/MSHA) or have been 

approved by the NRC. The inspectors evaluated whether the devices were used 

consistent with their NIOSH/MSHA certification or NRC approval.

Problem Identification and Resolution

The inspectors evaluated whether problems associated with the control and mitigation 

of in-plant airborne radioactivity were being identified by the licensee at an appropriate 

threshold and were properly addressed for resolution in the licensee corrective action 

program. The inspectors assessed whether the corrective actions were appropriate for 

a selected sample of problems involving airborne radioactivity and were appropriately 

documented by the licensee.

b. Findings

No findings were identified.

2RS4 Occupational Dose Assessment (71124.04)

a. Inspection Scope

During April 14 to 17, 2014, the inspectors verified that occupational dose is 

appropriately monitored, assessed and reported by NextEra. The inspectors used the 

requirements in 10 CFR Part 20, the guidance in RG 8.13 - Instructions Concerning 

Prenatal Radiation Exposures, RG 8.36 - Radiation Dose to Embryo Fetus, RG 8.40 -

Methods for Measuring Effective Dose Equivalent from External Exposure, TSs, and the 

licensee’s procedures required by TSs as criteria for determining compliance. 
26

Enclosure

Inspection Planning

The inspectors reviewed the results of NextEra RP program audits related to internal 

and external dosimetry. The inspectors reviewed the most recent National Voluntary 

Laboratory Accreditation Program (NVLAP) report on the principal dosimetry used to 

establish dose of legal record. 

A review was conducted of NextEra procedures associated with dosimetry operations, 

including issuance/use of external dosimetry, and assessments of external and internal 

dose for radiological incidents. The inspectors evaluated whether NextEra had 

established procedural requirements for determining when external dosimetry and 

internal dose assessments are required. 

External Dosimetry

The inspectors evaluated whether the NextEra dosimetry vendor is NVLAP accredited 

and if the approved irradiation test categories for each type of personnel dosimeter used 

are consistent with the types and energies of the radiation present. 

The inspectors evaluated the onsite storage of dosimeters before issuance, during use, 

and before processing/reading. The inspectors also reviewed the guidance provided to 

radiation workers with respect to use of dosimeters.

The inspectors assessed the use of EPDs to determine if NextEra uses a “correction 

factor” to address the response of the EPD as compared to the dosimeter of legal record 

for situations when the EPD is used to assign dose.

The inspectors reviewed three dosimetry occurrence reports or corrective action 

program documents for adverse trends related to EPDs. The inspectors assessed 

whether NextEra had identified any adverse trends and implemented appropriate 

corrective actions.

Internal Dosimetry

Routine Bioassay (In Vivo)

The inspectors reviewed procedures used to assess the dose from internally deposited 

radionuclides using whole body count (WBC) equipment. The inspectors evaluated 

whether the procedures addressed methods for differentiating between internal and 

external contamination, the release of contaminated individuals, determining the route of 

intake and the assignment of dose.

The inspectors reviewed the whole body count process to determine if the frequency of 

measurements was consistent with the biological half-life of the radionuclides available 

for intake. 

The inspectors reviewed NextEra evaluation for use of its portal radiation monitors as a 

passive monitoring system. The inspectors assessed if instrument minimum detectable 

activities were adequate to determine the potential for internally deposited radionuclides 

sufficient to prompt an investigation.
27

Enclosure

Special Bioassay (In Vitro)

There were no internal dose assessments obtained using In Vitro results for the 

inspectors to review, i.e., no urinalysis or fecal sample results.

The inspectors reviewed the vendor laboratory quality assurance program and assessed 

whether the laboratory participated in an industry recognized cross-check program 

including whether out-of-tolerance results were reviewed, evaluated and resolved 

appropriately.

Internal Dose Assessment – Airborne Monitoring

NextEra‘s had not performed any internal dose assessments using airborne/derived air 

concentration monitoring during the period reviewed.

Special Dosimetric Situations

Declared Pregnant Workers

The inspectors assessed whether NextEra informs workers of the risks of radiation 

exposure to the embryo/fetus, the regulatory aspects of declaring a pregnancy, and the 

specific process to be used for voluntarily declaring a pregnancy.

The inspectors reviewed the records of one individual who had declared pregnancy 

during the current assessment period and evaluated whether the NextEra‘s radiological 

monitoring program (internal and external) for declared pregnant workers is technically 

adequate to assess the dose to the embryo/fetus. The inspectors reviewed exposure 

results and monitoring controls that were implemented for declared pregnant workers 

during the inspection period.

Dosimeter Placement and Assessment of Effective Dose Equivalent for External 

Exposures

The inspectors reviewed the NextEra methodology for monitoring external dose in nonuniform radiation fields where large dose gradients exist. The inspectors evaluated the 

NextEra criteria for determining when alternate monitoring, such as use of multi-badging, 

is to be implemented. The inspectors reviewed selected dose assessments performed 

using multi-badging to evaluate whether the assessment was performed consistent with 

procedures and dosimetric standards. 

Shallow Dose Equivalent

The inspectors reviewed one dose assessments for shallow dose equivalent for 

adequacy. The inspectors evaluated the NextEra method for calculating shallow dose 

equivalent from distributed skin contamination. 

Neutron Dose Assessment

The inspectors evaluated the NextEra‘s neutron dosimetry program, including dosimeter 

types and radiation survey instrumentation.
28

Enclosure

The inspectors reviewed neutron exposure occurrences and assessed whether 

(a) dosimetry and instrumentation was appropriate for the expected neutron spectra, 

(b) there was sufficient sensitivity for low dose and/or dose rate measurement, and 

(c) neutron dosimetry and neutron detection instruments were properly calibrated. The 

inspectors also assessed whether interference by gamma radiation had been accounted 

for in the calibration.

Assigning Dose of Record

For the special dosimetric situations reviewed in this section, the inspectors assessed 

how NextEra assigns dose of record for total effective dose equivalent, shallow dose 

equivalent, and lens dose equivalent. This included an assessment of external and 

internal monitoring results, supplementary information on individual exposures, and 

radiation surveys when dose assignment was based on these techniques.

Problem Identification and Resolution

The inspectors assessed whether problems associated with occupational dose 

assessment were being identified by NextEra at an appropriate threshold and are 

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

inspectors assessed the appropriateness of the corrective actions for a selected sample 

of problems documented by the licensee involving occupational dose assessment.

b. Findings

No findings were identified.

2RS5 Radiation Monitoring Instrumentation (71124.05)

a. Inspection Scope

During May 19 to 23, 2014, the inspectors reviewed the accuracy and operability of 

radiation monitoring instruments that are used to protect occupational workers and to 

protect the public from nuclear power plant operations. The inspectors used the 

requirements in 10 CFR Part 20, 10 CFR Part 50 Appendix A - Criterion 60 Control of 

Release of Radioactivity to the Environment and Criterion 64 Monitoring Radioactive 

Releases, 10 CFR 50 Appendix I Numerical Guides for Design Objectives and Limiting 

Conditions for Operation to meet the Criterion “As Low as is Reasonably Achievable” for 

Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, 40 CFR 

Part 190 Environmental Radiation Protection Standards for Nuclear Power Operations, 

NUREG 0737 Clarification of Three Mile Island Corrective Action Requirements, 

TSs/Offsite Dose Calculation Manual (ODCM), applicable industry standards, and 

NextEra’s procedures required by TSs as criteria for determining compliance. 

Inspection Planning

The inspector reviewed the Seabrook Station UFSAR to identify radiation instruments 

associated with monitoring area radiation, airborne radioactivity, process streams, 

effluents, materials/articles, and workers. Additionally, the inspectors reviewed the 

associated TS requirements for post-accident monitoring instrumentation. The 

inspectors reviewed a listing of in-service survey instrumentation including: air 
29

Enclosure

samplers, small article monitors, radiation monitoring instruments, personnel 

contamination monitors, portal monitors, and whole-body counters. The inspectors 

assessed whether an adequate number and type of instruments were available to 

support operations. The inspectors reviewed NextEra and third-party evaluation reports 

of the radiation monitoring program since the last inspection. The inspectors reviewed 

procedures that govern instrument source checks and calibrations, including instruments 

used for monitoring transient high radiological conditions and instruments used for 

underwater radiation surveys. 

Walkdowns and Observations

The inspectors walked down four effluent radiation monitoring systems, including 

RM6454, Storm Drain Liquid Effluent Monitor; RM6509, Liquid Waste Test Tank 

Discharge Effluent Monitor; RM6528-1, Plant Vent Wide Range Gas Monitor Effluent

Monitor; and RM6526-1, Containment Air Particulate Monitor. The inspectors assessed 

whether the effluent/process monitor configurations align with what is described in the 

UFSAR and the ODCM. 

The inspectors selected five portable survey instruments in use or available for issuance 

and assessed calibration and source check stickers for currency, as well as, instrument 

material condition and operability. 

The inspectors observed NextEra staff performance in conducting source checks for the 

following types of portable survey instruments: MGPI Telepole, Eberline AMP-100 Area 

Radiation Monitor, Ludlum 44-9 GM Frisker, Eberline RM-14 Pancake Frisker and 

Eberline RO-2A Ion Chamber. The inspectors assessed whether high-range 

instruments are source checked on all appropriate scales.

The inspectors walked down five area radiation monitors (ARMs) and four CAMs to 

determine their adequacy and operability. The inspectors compared the ARM response 

(via local readout or remote control room indications) with actual area radiological 

conditions for consistency.

The inspectors selected the Argos 4AB personnel alpha/beta contamination monitor, the 

GEM-5 gamma portal monitors, the CHRONOS-4 large article monitor, and the SAM-12 

small article monitor located at the health physics (HP) Control Point, and evaluated 

whether the periodic source checks were performed in accordance with the 

manufacturer’s recommendations and NextEra procedures.

Calibration and Testing Program

Process and Effluent Monitors

The inspectors selected three process monitor instruments: RM6532-1, Primary 

Auxiliary Building (PAB) Air particulate; RM6526-2, Containment Radiogas; and 

RM6482-1/2, Main Steam Line Monitors and evaluated whether channel calibration 

and functional tests were performed consistent with Seabrook Station’s TSs/ODCM. 

The inspectors assessed whether; (a) NextEra calibrated its monitors with National 

Institute of Standards and Technology (NIST) traceable sources; (b) the primary 

calibrations adequately represented the plant radionuclide mix; (c) when secondary 
30

Enclosure

calibration sources were used, the sources were verified by comparison with the primary 

calibration source; and (d) NextEra channel calibrations encompassed the instrument’s 

alarm set-point range. 

Laboratory Instrumentation

The inspectors assessed laboratory analytical instruments used for radiological analyses 

to determine whether daily performance checks and calibration data indicate that the 

frequency of the calibrations is adequate and there were no indications of degraded 

performance. The inspectors assessed whether appropriate corrective actions were 

implemented in response to indications of degraded performance.

Whole Body Counter (WBC)

The inspectors reviewed calibration records for the WBC and the methods and sources 

used to perform functional checks on the WBC before daily use and assessed whether 

calibration and check sources were appropriate and align with the plant’s radionuclide 

mix and that appropriate calibration phantoms were used. 

Portal Monitors, Personnel Contamination Monitors, and SAMs

The inspectors selected one of each type of these instruments and verified that the 

alarm set-point values are reasonable to ensure that licensed material is not released 

from the site. The inspectors reviewed calibration documentation for each instrument 

selected and reviewed the calibration methods to determine consistency with the 

manufacturer’s recommendations.

Portable Survey Instruments, ARMs, Electronic Dosimetry, and Air Samplers/CAMs

The inspectors reviewed calibration documentation for the following portable instruments 

in use: Eberline AMS-4 Continuous Air Monitor, Eberline RO-2A Ion Chamber, Fluke 

Biomedical 451B Ion Chamber, RADECO HD-29A Air Sampler, MGP AMP 200 Area 

Monitor Probe and DMC 2000 Electronic Personal Dosimeter (EPD). For portable 

survey instruments and ARMs, the inspectors reviewed detector measurement geometry 

and calibration methods and reviewed the use of its instrument calibrator as applicable. 

Instrument Calibrator

The inspectors reviewed the current radiation output values for the licensee’s portable 

survey and ARM instrument calibrator units. The inspectors assessed that the licensee 

had verified calibrator output over the range of the exposure rates/dose rates using an 

ion chamber/electrometer. The inspectors verified that the measuring devices had been 

calibrated by a facility using NIST traceable methods and were properly applied by the 

licensee in performing radiation survey instrument calibrations.

Calibration and Check Sources

The inspectors reviewed the NextEra’s waste stream characterization per 10 CFR Part 

61, “Licensing Requirements for Land Disposal of Radioactive Waste,” to assess 

whether calibration sources used were representative of the types and energies of 

radiation encountered in the plant. 
31

Enclosure

Problem Identification and Resolution

The inspectors evaluated whether problems associated with radiation monitoring 

instrumentation were being identified by the licensee at an appropriate threshold and 

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

inspectors assessed the appropriateness of the corrective actions for a selected sample 

of problems documented by the licensee that involve radiation monitoring 

instrumentation.

b. Inspection Findings

No findings were identified.

4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

Reactor Coolant System (RCS) Specific Activity and RCS Leak Rate (2 samples)

a. Inspection Scope

The inspectors reviewed NextEra’s submittal for the RCS specific activity and RCS 

leak rate performance indicators for the period of April 1, 2013 to March 31, 2014. To 

determine the accuracy of the performance indicator data reported during those periods, 

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

“Regulatory Assessment Performance Indicator Guideline,” Revision 7. The inspectors 

also reviewed RCS sample analysis and logs of daily measurements of RCS leakage 

and activity, and compared that information to the data reported by the performance 

indicator.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution (71152 – 1 sample)

.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. 
32

Enclosure

b. Findings 

No findings were identified.

.2 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a semi-annual review of site issues, as required by Inspection 

Procedure 71152, “Problem Identification and Resolution,” to identify trends that might 

indicate the existence of more significant safety issues. In this review, the inspectors 

included repetitive or closely-related issues that may have been documented by NextEra 

outside of the CAP, such as trend reports, performance indicators, major equipment 

problem lists, system health reports, MR assessments, and maintenance or CAP 

backlogs. The inspectors also reviewed NextEra’s CAP database for the first and 

second quarters of 2014 to assess CRs written in various subject areas (equipment 

problems, human performance issues, etc.), as well as individual issues identified during 

the NRCs daily CR review (Section 4OA2.1). The inspectors reviewed NextEra’s 

quarterly trend report for the first quarter of 2014, conducted under PI-AA-207-1000, 

‘Station Self-Evaluation and Trending, Revision 1, to verify that NextEra personnel were 

appropriately evaluating and trending adverse conditions in accordance with applicable 

procedures.

b. Findings and Observations

No findings were identified.

In general, the inspectors did not identify significant issues with the content and 

utilization of the station trending report, including any identified adverse trends, potential 

adverse trends, or management awareness areas. As a result, the inspectors evaluated 

a sample of departments that are required to provide input into the quarterly trend 

reports, which included operations, maintenance and engineering departments. This 

review included a sample of issues and events that occurred over the course of the past 

two quarters to objectively determine whether issues were appropriately considered or 

ruled as emerging or adverse trends, and in some cases, verified the appropriate 

disposition of resolved trends. The inspectors verified that these issues were addressed 

within the scope of the corrective action program, or through department review and 

documentation in the quarterly trend report for overall assessment.

For example, the inspectors noted that consistent with the issue identified in Section 

4OA3.3, regarding the surveillance test failure, NextEra had identified that a critical 

warm-up period identified in manufacturer documents was not translated to applicable 

surveillance procedures. While this was a dated issue from several years ago, this 

aspect of design interface weaknesses was identified by electrical maintenance following 

the implementation of a design change for protective relays. However, while not 

identified specifically as a trend, the issue of appropriately ensuring design interface 

documents are appropriately revised or updated is being addressed through the 

corrective action program for the issue referenced later in this report.
33

Enclosure

Another example identified by the inspectors involved the use of the CAP (versus the 

trend report) to address trends in the use of the RCS Leak Rate Program. Operations 

personnel coordinated the use of the CAP and the Training Department to ensure the 

differences seen in the calculation of identified RCS leak rate were well understood by 

the control room operators, but required some potential training due to some 

weaknesses identified in this area.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 3 samples)

.1 (Closed) Licensee Event Report (LER) 05000443/2013-001: Failure to Enter Technical 

Specification Following Discovery of SW Leak

As documented in NRC Inspection Report, 50-443/2013-005, the inspectors 

dispositioned a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, 

“Instructions, Procedures, and Drawings,” and an associated violation of technical 

specification (TS) 3.7.4, because NextEra did not follow the requirements of station 

procedure EN-AA-203-1001, “Operability Determinations/ Functionality Assessments.” 

Specifically, NextEra did not properly evaluate and document an adequate basis for 

operability, when relevant information was available that would have challenged the 

“reasonable expectation of operability” threshold for a service water (SW) through-wall 

leak that degraded incrementally from weepage on August 7, 2013, to a significantly 

larger leak on August 28, 2013. NextEra completed a temporary non-code repair of the 

flaw with the installation of a weldolet on September 1, 2013, following NRC review and 

approval of a relief request. Additionally, during refueling outage OR16, in April 2014, 

NextEra removed the temporary weldolet and installed a new piping segment to 

complete the restoration of the section of degraded piping that had developed a leak in 

August 2013. Extent of condition inspections and liner repairs were also completed for 

similar piping configurations to the degraded piping segment that were inaccessible 

during power operation. Moreover, the subject LER was submitted by NextEra when 

they had concluded, on October 30, 2013, following a review of the actions taken for the 

SW leak in question, that the plant had operated in a condition prohibited by technical 

specifications for 24 days, from August 8, 2013 to September 1, 2013. The inspector 

reviewed several extent of condition inspections, evaluations and repairs, the actual 

degraded pipe restoration activity (See Section 1R19 for the final acceptance leak test), 

and verified the adequacy of NextEra’s additional corrective actions for the performance 

issues that contributed to the identified TS violation. These actions included: (1) 

mentoring of individuals on documentation requirements for prompt operability 

determinations, (2) procedure revisions to ensure additional barriers are in place for 

future engineering evaluations regarding field information, assumptions and supervisory 

reviews, (3) corrective action program requirements were reinforced, and (4) a case 

study of this service water leak issue was identified to reinforce the relative issues to 

station personnel.

As a result of this inspection, no additional findings or violations of NRC requirements 

were identified. This LER is closed. 
34

Enclosure

.2 (Closed) Licensee Event Report (LER) 05000443/2014-001: Reactor Trip Due to Delay 

in Bus Transfer Resulting in Reactor Coolant Pump (RCP) Loop Low Flow

a. Inspection Scope

On April 1, 2014, at 00:26 while operating at approximately 15 percent power following 

turbine shutdown and removal of the main generator from service, Seabrook station 

experienced an automatic reactor trip on reactor coolant two loop loss of flow. The loss 

of flow was the result of the unexpected closure of the main generator breaker (MGB) ‘B’ 

phase which caused the 345kV Bus 6 to de-energize and isolate the MGB. All buses 

transferred to the reserve auxiliary transformers as designed; however, a slight delay in 

the automatic transfer for bus 1 resulted in two RCPs tripping. The RCPs tripping 

resulted in an automatic reactor trip due to reactor coolant loop low flow. The 

emergency feedwater system actuated on low steam generator level, and all plant 

equipment functioned as expected. NextEra personnel completed a root cause 

evaluation to determine the cause of the reactor trip. Corrective actions include: revising 

procedures to add controls regarding the potential risk, ensure the use of guarded 

equipment controls, and minimizing the time spent with the main generator breaker in 

local. 

The inspectors reviewed LER 2014-001-00 and associated corrective actions and 

identified a performance deficiency that was characterized as more than minor and is 

documented below. This LER is closed. 

b. Findings

Introduction. The inspectors identified a Green self-revealing finding, because NextEra 

did not ensure that adequate procedural guidance existed in ON1046.12, “Operation of 

the Main Generator Breaker” to limit the likelihood of events that upset plant stability. 

Specifically, Seabrook station experienced an automatic reactor trip from approximately 

15 percent reactor power on April 1, 2014 when two of four reactor coolant pumps 

(RCPs) tripped on low bus voltage. The cause of the reactor trip was directly attributable 

to the main generator breaker inadvertently closing and actuating the main generator 

multi-function protective relay. 

Description. At midnight on April 1, 2014, with reactor power at approximately 

15 percent, the station turbine generator was shutdown and the main generator breaker 

was opened in preparation for the start of a scheduled refueling outage. The Main 

Generator Breaker Selector Switch was subsequently aligned to local in preparation for 

scheduled main turbine overspeed testing. At 12:26 a.m., the main generator breaker 

‘B’ phase unexpectedly closed, actuating the main generator multi-function protective 

relay. As a result of the protective relay actuation, 13.8kV buses 1, 2, 3, 4, 5, and 6 

automatically transferred from the UATs to the Reserve Auxiliary Transformers (RATs). 

However, buses 1 and 5 experienced a delayed transfer from the UATs to the RATs 

based on residual bus voltage, and the two RCPs powered from buses 1 and 5 tripped 

on low bus voltage as designed. The delayed transfer of buses 1 and 5 was due to a 

combination of the delay introduced by the ‘B’ pole remaining closed and initiating an out 

of synchronization condition, and the more restrictive dead band reset of the newer style 

synchronization check relays that were installed on buses 1 and 5. The subsequent loss 

of reactor coolant flow caused an automatic reactor protection system (RPS) actuation 

and reactor trip on loop low flow. 
35

Enclosure

NextEra determined that the cause of the ‘B’ main generator breaker pole unexpected 

closure was from the combination of an air leak from a stuck open pressure reducer 

valve in the main generator ‘B’ pole control cabinet, and inadvertent contact with the ‘B’ 

pole local closure push button inside the ‘B’ pole control cabinet during an investigation 

of the air leak by an operator. NextEra determined that the root cause for the event was 

inadequate procedural guidance contained in ON1046.12, “Operation of the Main 

Generator Breaker” to communicate the impacts of positioning the Main Generator 

Selector Switch to local, take mitigating actions, and minimize time spent at increased 

risk configurations. Specifically, ON1046.12 contained no cautions, notes or 

prerequisites to ensure that licensee personnel were aware of the potential risk 

associated with the Main Generator Selector Switch being placed in local. Specifically, 

with the switch in local, the main generator breaker pole disagreement protective 

features that would have auto-opened the pole upon inadvertent closure were disabled. 

As a result of the lack of procedural information, licensee personnel did not implement 

controls in accordance with OP-AA-102-1003, “Guarded Equipment,” or take measures 

to minimize the time spent in the risk significant configuration. NextEra noted that 

procedures used for controlling similar evolutions contain additional cautions and notes 

that identify the increased risk configuration. 

NextEra entered the event into their CAP (AR 01953543), and conducted a root cause 

evaluation to determine the root and contributing causes, extent of condition and extent 

of cause, and to identify corrective actions to prevent recurrence. NextEra initiated 

actions to revise ON1046.12 to add controls to communicate the potential risk 

associated with placing the main generator breaker control in local, conducted briefings 

with Maintenance groups involved in the event, and evaluated the adequacy of other 

Operations procedures that place equipment in a configuration where protective features 

are bypassed or defeated. The inspectors reviewed the root cause evaluation and 

associated documentation, and determined that NextEra adequately identified the root 

and contributing causes and implemented appropriate corrective actions to prevent 

recurrence.

Analysis. The inspectors determined that the inadequate procedural guidance contained 

in ON1046.12, “Operation of the Main Generator Breaker” was a performance deficiency 

that was within NextEra’s ability to foresee and correct. The performance deficiency was 

more than minor because it was associated with the procedure quality attribute of the 

Initiating Events cornerstone, and it adversely affected the cornerstone objective to limit 

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

during shutdown as well as power operations. The finding was evaluated under IMC 

0609, Attachment 4, “Phase 1 – Initial Characterization of Findings.” The inspectors 

determined that the finding is of very low safety significance (Green) because it did not 

result both in a reactor trip and the loss of mitigating equipment relied upon to transition 

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

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

NextEra did not ensure that a process of planning, controlling, and executing work 

activities such that nuclear safety is the overriding priority was implemented (H.5). 

Specifically, ON1046.12, “Operation of the Main Generator Breaker” did not contain 

adequate procedural guidance regarding the impacts of positioning the Main Generator 

Selector Switch to local, take mitigating actions, and minimize time spent at increased 

risk configurations [H.5].
36

Enclosure

Enforcement. Enforcement action does not apply because the performance deficiency 

did not involve a violation of a regulatory requirement. (FIN 05000443/2014003-02, 

Unexpected Main Generator Breaker Pole Closure Results in Reactor Trip)

.3 (Closed) Licensee Event Report (LER) 05000443/2014-002: Reactor Coolant Pump 

Undervoltage Time Delay Relay Exceeds Acceptance

On April 6, 2014, during a refueling outage, routine RPS surveillance testing identified 

that three of four reactor coolant pump (RCP) undervoltage (UV) reactor trip channels 

exceeded the TS channel response time acceptance criteria of 1.5 seconds for the RCP 

UV reactor trip function. NextEra determined that since this condition involved multiple 

similar components, there is evidence indicating that this condition may have arisen over 

time and three channels of RCP UV were concurrently inoperable. This resulted in the 

plant operating in a condition prohibited by TSs for approximately seventeen months. 

NextEra personnel initiated a root cause evaluation to determine the cause of the 

violation and determine appropriate corrective actions, replaced one relay, and adjusted 

the remaining relays to acceptable response times.

The inspectors reviewed LER 2014-002-00 and associated corrective actions and 

dispositioned the issue as a licensee identified violation of regulatory requirements. The 

enforcement aspects of this issue are discussed in Section 4OA7. This LER is closed. 

4OA6 Meetings, Including Exit

On July 10, 2014, the inspectors presented the inspection results to Mr. Dean Curtland, 

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

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

this report.

4OA7 Licensee-Identified Violation

The following violation of very low safety significance (Green) was identified by NextEra 

and is a violation of NRC requirements which meets the criteria of the NRC Enforcement 

Policy for being dispositioned as an NCV.

 Technical Specification (TS) Surveillance Requirement 4.3.1.2 requires verification of 

the response time of each reactor trip function every 18 months. During the 18 

month surveillance testing of the RCP UV channels conducted on April 6, 2014, 

three of the four RCP UV relays exceeded their allowable maximum response time, 

resulting in their associated UV reactor trip channels exceeding the limit of 1.5 

seconds. NextEra determined that the three channels were inoperable. TS 3.3.1, 

‘Reactor Trip System Instrumentation’, requires four channels of RCP UV 

instrumentation to be operable in Mode 1. With three RCP UV channels inoperable 

in Mode 1, the plant is required to initiate a shutdown within one hour in accordance 

with TS 3.0.3. NextEra determined that this condition existed from the time the 

relays were last calibrated in OR15 (September 20, 2012) until the plant entered 

OR16 (April 1, 2014). Contrary to TS 3.0.3, Seabrook station operated in Mode 1 

with three of four RCP UV channels inoperable for approximately 17 months without 

taking the required TS actions. NextEra entered this issue into the CAP as AR 

01964167 and performed a detailed analysis of the impact of the increased channel 
37

Enclosure

response time. NextEra, in consultation with Westinghouse, determined that the 

safety function of the RCP UV trip channel (prevention of departure from nucleate 

boiling) was maintained throughout the period of inoperability. NextEra planned to 

develop a maintenance procedure to allow for on-line re-calibration of the RCP 

UV relays. The inspectors determined that the finding was of very low safety 

significance (Green) in accordance with IMC 0609, Appendix A, “Determining the 

Significance of Reactor Inspection Findings for At-Power Situations” because the 

deficiency did not affect a single RPS trip signal to initiate a reactor scram and the 

function of other redundant trips or diverse methods of reactor shutdown, did not 

involve control manipulations that unintentionally added positive reactivity, and did 

not result in a mismanagement of reactivity by operators.

ATTACHMENT: SUPPLEMENTARY INFORMATION
A-1

Attachment

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Curtland, Site Vice President

T. Vehec, Site Director

R. Dodds, Plant General Manager 

J. Berg, Chemistry Supervisor

K. Boehl, Radiation Protection Supervisor

V. Brown, Senior Licensing Engineer

R. Campion, Oversight Supervisor

A. Chesno, Performance Improvement and Licensing

M. Collins, Engineering Director

J. Connolly, Engineering Director

K. Douglas, Maintenance Director

D. Egonis, Repairs and Welding Services

M. Feeney, Instrumentation and Controls Dept. Head

D. Flahardy, Radiation Protection Manager

S. Hamel, Engineering Support

Z. Kuljis, WesDyne International, EC Level III

R. Leider, Engineer

B. McAllister, Nuclear Engineer

M. Ossing, Licensing Manager

D. Perkins, Radiation Protection Analyst

D. Robinson, Chemistry Manager

D, Snyder, Boric Acid Corrosion Control Program Manager

T. Vassallo, Engineering

T. Waechter, Project Manager

B. Westovic, WesDyne International, Waltz Mill Facility

K. Whitney, ISI Program Manager

LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED

Opened/Closed

05000443/2014003-01 FIN Inadequate Technical Evaluation of SafetyRelated Structures (Section 1R12)

05000443/2014003-02 FIN Unexpected Main Generator Breaker Pole Closure 

Results in Reactor Trip (Section 4OA3.2)

Opened

None

Closed

05000443/2013-001-00 LER Failure to Enter Technical Specification Following 

Discovery of SW Leak (Section 4OA3.1)
A-2

Attachment

05000443/2014-001-00 LER Automatic Reactor Trip Due to Delay in Automatic 

Bus Transfer Resulting in RCP Loop Low Flow 

(Section 4OA3.2)

05000443/2014-002-00 LER Reactor Coolant Pump Undervoltage Time Delay 

Relay Exceeds Acceptance (Section 4OA3.3)

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Procedures

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

ON1246.03, GSU Trouble, Revision 5

OP-AA-102-1002, Seasonal Readiness, Revision 3

OS1200.03, Severe Weather Conditions, Revision 2

Condition Reports

01932588 01962633 01964824 01964840 01967006 01968990

01969460 01969599 01970614

Maintenance Orders/Work Orders

40268452 40244709

Miscellaneous

ISO New England Operating Procedure No. 4, ‘Action During a Capacity Deficiency’, 

Revision 12

Master/Local Control Center Procedure No. 1, ‘Nuclear Plant Transmission Operations’, 

Revision 13

Master/Local Control Center Procedure No. 2, ‘Abnormal Conditions Alert’, Revision 17.1

Seasonal Readiness Memo to Mano Nazar dated 5/24/14

UFSAR Chapter 2

UFSAR Section 8

Section 1R04: Equipment Alignment

Procedures

OS1001.01, Reactor Coolant System Fill and Vent, Revision 25

OS1013.03, Residual Heat Removal Train A Startup and Operation, Revision 27

OS1016.01, Service Water System Fill and Vent, Revision 18 

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

OS1056.03, Containment Penetrations, Revision 10

OS1090.05, Component Configuration Control, Revision 57

Condition Reports

00207129 00208974 00208987 01843745 01918488 01955956

01956317 01960210
A-3

Attachment

Miscellaneous

Seabrook Station Updated Final Safety Analysis Report, Chapter 9, Revision 12

Section 1R05: Fire Protection

Condition Reports

01803431 01967187 01969430 1958635 1958639

Miscellaneous

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, CE-F-1-A

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, C-F-1-Z, C-F-2-Z, & C-F-3-Z

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, EFP-F-1-A

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, PAB-F-3A-Z, PAB-F-3B-Z, & 

PAB-F-4-Z

Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, PLT-F-1-0

Procedures

OS1200.00, Response to Fire or Fire Alarm Actuation, Revision 21

FP-AA-104, Fire Protection Program, Revision 0

Section 1R06: Flood Protection Measures

Condition Reports

01969726

Miscellaneous

Seabrook Station Moderate Energy Line Break Study

Seabrook Station UFSAR, Revision 15, Section 3 & Section 9

Maintenance Orders/Work Orders

40209732-06 40209732-07 40277675-01 40241371-01

Drawings

9763-F-310248

9763-F-310249

Section 1R07: Heat Sink Procedures

Procedures

ES1850.017, SW Heat Exchanger Program, Revision 1

PEG-268, Heat Exchanger and NRC GL 89-13 Program, Revision 0

Condition Reports

01809068 01957744 01958294 01959163

Maintenance Orders/Work Orders

40262651

Miscellaneous

Calculation C-S-1-25115, DG Heat Exchanger (DG-E-42A/B) Performance After Tube Plugging, 

Revision 0
A-4

Attachment

Section 1R08: In-service Inspection

Procedures

Engineering Department Instructions, EDI N0. 30560, Boric Acid Evaluations, Revision 1

NextEra Energy Program Description ER-AP-121, Steam Generator Integrity, Revision 0

Seabrook Station, Engineering Procedure, Reactor Vessel Head Penetration Ultrasonic 

Examination Analysis, ES13-01-08, Revision 00

Seabrook Station, Engineering Procedure, Procedure for Ultrasonic Examination of Reactor 

Vessel Head Penetrations, ES13-01-07, Revision 00

Seabrook Station, Engineering Procedure, RPVH Nozzle Bottom OD Surface Eddy Current 

Inspection, ES-01-01, Revision 00

Seabrook Station, Engineering Procedure, Perform VE of RPV Top Head At Penetration 57, 

Unit 1, 4/11/14

Seabrook Station, Engineering Procedure, Eddy Current Inspection of Preservice and Inservice 

Heat Exchanger Tubing, ES01-100, Revision 07

Seabrook Station, Procedure, Steam Generator Eddy Current Data Analysis Guidelines Manual, 

Revision 7

Seabrook Station Administrative Procedure, Boric Acid Corrosion Control Program MA 10.3, 

Revision 12

Seabrook Station Engineering Procedure, Manual Ultrasonic Procedure for the Examination of 

Non-PDI Nozzle Inner Corner Regions, ES10-01-38, Revision 01

Seabrook Station Engineering Procedure, Manual Ultrasonic Procedure for the Examination of 

Non-RPV Nozzle-to-Shell Welds in Vessels >2”, ES10-01-39, Revision 01

Seabrook Station Reference Manual, Steam Generator Maintenance Reference, SGRE, 

Revision 20

Westinghouse Procedure Number MRS-2.4.2 GEN 35, Revision 15, 8/11/11; Eddy Current 

Inspection of Preservice and Inservice Heat Exchange Tubing

Westinghouse SG-SGMP-12-15, Revision 1, Seabrook OR15, Condition Monitoring 

Assessment and Final Operational Assessment, January 2013

Westinghouse SG-SGMP-13-21, Revision 3, Steam Generator Degradation Assessment for 

Seabrook OR16 Refueling Outage, March 2014

Program Documents

EPRI Report 1000975, November 2001; Boric Acid Corrosion Guidebook, Revision 1; Managing 

Boric Acid Corrosion Issues at PWR Power Stations

NEI 03-08, January 2010; Guidelines for the Management of Materials Issues, Revision 2

NEI 97-06, Pressurized Water Reactor Steam Generator Examination Guidelines, 

Requirements 1013706, Revision 7

Seabrook Station Reference Manual, Inservice Inspection, SIIR Revision 16, 12/23/13

WCAP-15988, Revision 1, February 2005; Generic Guidance for an Effective Boric Acid 

Inspection Program for Pressurized Water Reactors

Westinghouse Non-Proprietary Class 3, SG-SGMP-09-22, Revision 2; Seabrook OR13 

Condition Monitoring Assessment and Final Operational Assessment, March 2010

Westinghouse Owner’s Group Letter WOG 05-91, dated March 15, 2005; Subject: Transmittal 

of the Final Non-Proprietary Version of WCAP-15988-NP, Revision 1 Entitled “Generic 

Guidance for an Effective Boric Acid Inspection Program for Pressurized Water 

Reactors”, February 2005, PA-MSC-0096

Westinghouse Seabrook OR14 Condition Monitoring and Operational Assessment, SG-SGMP11-14, Revision 0; April 2011
A-5

Attachment

Westinghouse Seabrook OR16 Condition Monitoring Assessment and Final Operational 

Assessment, SG-SGMP-14-10, Revision 0, April 2014 

Westinghouse Steam Generator Degradation Assessment for Seabrook OR15 Refueling 

Outage, SG-SGMP-12-8, Revision 0, September 2012

Westinghouse Steam Generator Degradation Assessment for Seabrook OR15 Refueling 

Outage, SG-SGMP-12-8, Revision 1, September 2012

Program Health Reports

Boric Acid Corrosion Control Program, 7/1/2013 – 9/30/2013

Boric Acid Corrosion Control Program, 10/1/2013 – 12/31/2013

Eddy Current Examination Technique Sheets

Eddy Current Examination Technique Sheet, ETSS #96004.1, Revision 13, April 2010

Eddy Current Examination Technique Sheet, ETSS #96004.3, Revision 13, April 2010

Eddy Current Examination Technique Sheet, ETSS #96005.2, Revision 9, July 2006

Eddy Current Examination Technique Sheet, ETSS #24013.1, Revision 2, August 2006

Eddy Current Examination Technique Sheet, ETSS #10013.1, Revision 1, May 2012

Eddy Current Examination Technique Sheet, ETSS #128411, Revision 3, February 2011

Eddy Current Examination Technique Sheet, ETSS #128413, Revision 3, February 2011

Eddy Current Examination Technique Sheet, ETSS #27091.2, Revision 1, September 2012

Eddy Current Examination Technique Sheet, ETSS #21409.1, Revision 7, May 2010

Eddy Current Examination Technique Sheet, ETSS #21998.1, Revision 4, August 2006

Eddy Current Examination Technique Sheet, ETSS #128424, Revision 3, February 2011

Eddy Current Examination Technique Sheet, Appendix 1, ETSS #124425, Revision 3, 

February 2011

Eddy Current Examination Technique Sheet, Appendix 1, ETSS #128431, Revision 2, 

February 2011

Eddy Current Examination Technique Sheet, Appendix 1, ETSS #128432, Revision 2, 

February 2011

Eddy Current Examination Technique Sheet, ETSS #21410.1, Revision 6, October 2006

Eddy Current Examination Technique Sheet, ETSS #22401.1, Revision 4, August 2006

Eddy Current Examination Technique Sheet, ETSS #96511.1, Revision 16, August 2006

Eddy Current Examination Technique Sheet, ETSS #20510.1, Revision 7, October 2006

Eddy Current Examination Technique Sheet, ETSS #21511.1, Revision 8, July 2006

Eddy Current Examination Technique Sheet, ETSS #279101.1, Revision 1, May 2012

Eddy Current Examination Technique Sheet, ETSS #27901.3, Revision 1, May 2012

Eddy Current Examination Technique Sheet, ETSS #27902.1, Revision 1, May 2012

Eddy Current Examination Technique Sheet, ETSS #27902.2, Revision 1, May 2012

Eddy Current Examination Technique Sheet, ETSS #27902.3, Revision 1, May 2012

Eddy Current Examination Technique Sheet, ETSS #10908.4, Revision 1, May 2012

Eddy Current Examination Technique Sheet, ETSS #20407.1, Revision 7, July 2006

MRS-TRC-2163, Seabrook Appendix H & I Techniques Spring 2014 Inspection, April 2014

Steam Generator Eddy Current Inspection Parameters

ACTS# NAH-01-114, Revision 0, Jan. 15, 2014

ACTS# NAH-02-114, Revision 0, Jan. 16, 2014

ACTS# NAH-03-114, Revision 0, Jan. 16, 2014

ACTS# NAH-04-114, Revision 0, Jan. 16, 2014

ACTS# NAH-05-114, Revision 0, Jan. 16, 2014
A-6

Attachment

ACTS# NAH-06-114, Revision 0, Jan. 16, 2014

ACTS# NAH-07-114, Revision 0, Jan. 15, 2014

ACTS# NAH-08-114, Revision 0, Jan. 15, 2014

ACTS# NAH-09-114, Revision 0, Jan. 15, 2014

ACTS# NAH-10-114, Revision 0, Jan. 16, 2014

ACTS# NAH-11-114, Revision 0, Jan. 15, 2014

ANTS# NAH-A-114, Revision 0, Jan. 9, 2014

ANTS# NAH-B-114, Revision 0, Jan. 9, 2014

ANTS# NAH-C-114, Revision 0, Jan. 9, 2014

ANTS# NAH-D-114, Revision 0, Jan. 9, 2014

ANTS# NAH-D-114, Revision 0, Jan. 9, 2014

Containment References

Calculation C-S-1-10096, Containment Liner Wall Thickness Requirements Guideline

OR16 Containment Liner Dome Anomaly Examination Plan Revision 00; 4/2/14

Revision 1, 1/21/03

Seabrook Station Engineering Procedure ES1807.032, Revision 01; “Inservice Inspection 

Procedure Primary Containment Section XI IWE Program”

United Engineers and Constructors, Inc.; “Containment Design Specification For Public Service 

Company of New Hampshire Seabrook Station Unit Nos. 1 & 2”

Condition Reports (*NRC identified)

00208123 00392489 01792134 01799982 01800069 01800665

01804656 01805955 01806325 01807338 01808286 01809420

01809558 01809566 01812473 01812701 01812870 01814971

01818743 01833942 01837215 01849156 01850224 01856860

01874676 01875085 01875160 01875925 01884233 01884233

01885999 01888956 01892958 01903871 01915704 01924003

01931227 01935241 01938925 01940747 01941757 01945332

01954160 01956084 01958085 01960191* 01960193*

Maintenance Orders/Work Orders

0138224401 0138224402 4020903201 4010817001

Miscellaneous

Code Class 1, 2, and 3 Piping, 6/15/90

NEXTERA Energy, Nuclear Fleet, Process Description, PI-AA-204, Revision 24; Condition 

Identification and Screening Process, 1/30/14

NEXTERA Energy, Nuclear Fleet, Process Description, PI-AA-205, Revision 25; Condition 

Evaluation and Corrective Action, 1/30/14

NRC Letter Dated 3/28/2007, Subject: Seabrook Station Unit No. 1 – Issuance of Amendment 

Re: Technical Specification Task Force (TSTF)-449, “Steam Generator Tube Integrity,” 

(TAC NO. MD0696)

Nuclear Regulatory Commission (NRC) Generic Letter 90-05; Guidance for Performing 

Temporary Non-Code Repair of ASME

Seabrook Station Reference Manual, SIIR, Revision 16 (contains Alloy 600 examinations), 

Sections 8.2, 8.3, and 8.4

Steam Generator Management Program: Pressurized Water Reactor Steam Generator 

Examination Guidelines, Revision 7, October 2007
A-7

Attachment

Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines, 

Revision 3, November 2009

Section XI Repair/Replacement Samples

W0108170 01, Relief Valve Setpoint Pressure Test of valve 1-SV-V-60, completed on 10/26/12

W0220458 03, Replace P-41B/D Discharge Piping With AL6XN per EC278717, completed 

on 11/14/13

NDE Examination Reports & Data Sheets

40209032 01, OR16, Perform VT-2 RPV Top Head Under Insulation

40209032 01, OR16, Perform VT-2 OF RPV Top Head Under Insulation, 4/7/14

GV & VT-3 Examination Data Sheet 000446, Mechanical Penetration PEN-X62, 8/16/01

GV & VT-3 Examination Data Sheet 05-13-148, Mechanical Penetration PEN-X62, 5/12/05

GV & VT-3 Examination Data Sheet 06-13-007, Mechanical Penetration PEN-X62, 10/1/06

GV Examination, 12-13-003, 90-180 EL 119 to 192, 10/3/12, 100 percent Accept, 12-13-003

GV Examination Data Sheet 11-13-169, Mechanical Penetration PEN-X62, 5/14/11

GV Examination Data Sheet 12-13-008, Mechanical Penetration PEN-X62, 10/15/12

IWE – VT-3 Examination Data Sheet 09-06-043, Mechanical Penetration PEN-X62, 12/8/10

IWE – VT-3 Examination Data Sheet 09-06-043, Mechanical Penetration PEN-X62, 4/16/08

IWE-VT-1 Examination, 90-180 EL 119 to 192, Indication IWE-12-159-176, accept by 

engineering evaluation (11 pages) 10/3/12

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1421, 10/10/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1422, 10/10/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1429, 10/9/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1511, 10/9/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1512, 10/9/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1413, 10/24/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1514, 10/24/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1515, 10/24/13

Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1516, 9/13/13

NextEra, Seabrook Unit 1, OR16 License Renewal Containment Liner UT Exam, 5/7/14

SG A, OR16 Listing of eddy current indications detected and reported, 4/14/14

SG B, OR16 Listing of eddy current indications detected and reported, 4/14/14

SG C, OR16 Listing of eddy current indications detected and reported, 4/14/14

SG D, OR16 Listing of eddy current indications detected and reported, 4/14/14

Ultrasonic Report Data Sheet: NAH-R16-CPO2-57-4150-01Y, Head Penetration #57 (7 Pages)

Ultrasonic Thickness Examination Data Sheet 08-06-170, Mechanical Penetration PEN-X62, 

10/8/09

Ultrasonic Thickness Examination Data Sheet 11-01-033, Mechanical Penetration PEN-X62, 

5/1/11

UT Data Sheet, 09-06-031, IWE-VT-1 Examination, 90-180 EL 119 to 192, 78 percent, 

Indication IWE-12-159-176 accept by engineering evaluation, 10/29/09

UT Data Sheet 12-01-143, SI 0251-07, Pipe-to-Pipe weld, 10/8/12, Steps 4.11, 4.12, sign off of 

final NDE for SW piping fabrication

UT Data Sheet: 14-01-020, component: RC-E-10 A-IR, Nozzle Inner Radius, NRI, 4/15/14

UT Data Sheet: 14-01-021, component: RC-E-10 A-NZ, nozzle-to-vessel weld, NRI, 4/15/14

UT Data Sheet: 14-01-022, component: RC-E-11A 2A-IR, Nozzle Inner Radius, NRI, 4/15/14

UT Data Sheet: 14-01-023, component: RC-E-11A 2A-NZ, nozzle-to-vessel weld, NRI, 4/15/14

UT Data Sheet: 14-01-024, component: RC-E-11A 2B-IR, Nozzle Inner Radius, NRI, 4/15/14
A-8

Attachment

UT Data Sheet: 14-01-025, component: RC-E-11 2B-NZ, nozzle-to-vessel weld, NRI, 4/15/14

Wes Dyne, RPVH CRDM Nozzle 57 OD Manual Eddy Current (ET) Examination Report 

April 2014 OR16 Outage – Final Report, 4/16/14 (15 pages)

Eddy Current Data Acquisition Technicians and Data Analysts Certifications

10232/P4536 10271/E3848 10275/G8167 10281/P2305

10288/S1703 10298/D8884 10305/K9235 10312/S7963

10325/S1253 10339/H1487 10344/P8908 10399/Y1950

18835/P2465 19166/M2691 19257/T8398 20494/W2506

20613/B0613 23225/B4838 25553/H3260 28737/E4963

40256/M8414 45189/J5189 51753/K1753 A2945

A3502 B9540 C0042 C2886

C3274 C9055 C9162 D2172

D4576 D4816 D8021 F1726

F4008 G0071 G3127 H1748

H6377 I2393 J0009 J0145

J1515 J8109 K0073 K2676

K8138 L1107 L1342 L6066

L7003 M3442 P0017 P0116

R0609 R1509 R3919 R7770

S4256 S5339 S5760 S9385

T0042 T3673 V6207 W1758

W2545 Y0624 Z0059

Weld Traveler

40220458-03, 9/25/13, 6 pages, SW Piping Fabrication

Engineering Evaluations, Analyses, Calculations & Standards

ASME IWE VT-1 Examination Indication Containment Dome, AR1809517, 10/4/12, 

Revised 10/24/12

ASME IWE VT-1 Examination Indication Containment Dome, AR1809517, 4/3/14, 

Revised 4/12/14

Calculation C-S-1-24004, Revision 07, RC System, 9/13/2012; Seabrook Reactor Vessel Head 

Effective Degradation Years (EDY) & Re-Inspection Years 

Containment Liner IWE Examination Indications AR1646065, 5/15/13

Evaluation #14-013, REVISION 0, 4/14/14; Seabrook Station Reactor Vessel Head Effective 

Degradation Years (EDY) & Re-Inspection Year for Cycle 16

Boric Acid Corrosion Control (BACC) Leak Screening

WR94094749, AR01956281; 1-CS-V-143 (Charging to Regen HX Isolation); no 

evaluation, 4/12/14

WR94095011, AR01957408; 1-SI-V-25 (Isolation Level Column Accumulator Tank 9B); no 

evaluation, 4/12/14

WR94095012, AR01957412; 1-SI-V-223 (Drain Valve-Level Column Accumulator Tank 9D); no 

evaluation, 4/12/14

WR94095014, AR01957417; 1-SI-V-57 (Isolation for LT 956 for Accumulator Tank 9D); no 

evaluation, 4/12/14

WR94095016, AR01957419; 1-SI-V-55 (Isolation for LT 957 for accumulator Tank 9D); no 

evaluation, 4/12/14
A-9

Attachment

WR94095018, AR01957422; 1-SI-V-52 (Isolation sample connection for accumulator Tank 9D); 

no evaluation, 4/12/14

Boric Acid Corrosion Control (BACC) Leak Evaluation

AR01792134, B3 Leak from fitting on 1-CBS-P-9-A, 8/8/12

AR01884233, Active Boric Acid Leak, valve 1-CS-TCV-381-B, 6/21/13

AR01915704, Inactive Leak at 1-CS-V-250 Has Become an Active Leak, 10/28/13

AR01948324, 1-CS-FCV-121 Active Boric Acid Packing Leak, 3/14/14

AR01955327, Active Boric Acid Leak at 1-SI-V-139, 4/6/14

Drawings

United Engineers drawing 9763-F-805147, Revision 7, Seabrook Unit 1; Containment Structure 

Piping Zones 57E & F Line NOS. 1214, 1216, 1225 & 1226, 12/11/90

Section 1R12: Maintenance Effectiveness

Procedures

EX1805.01. Visual Examination and Functional Testing Program for Snubbers, Revision 11

MS0515.08, Paul-Monroe 2300, 2400, and 2500 Hydraulic Snubber Maintenance, Revision 4

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

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

Condition Reports

00140964 01282530 01686357 01688626 01725046 01804460

01811427 01929460 01932657 01957128 01959695 01959700

01978310

Maintenance Orders/Work Orders

40288714 40287787

Miscellaneous

ACI 349.3R-02, Evaluation of Existing Nuclear Safety-Related Concrete Structures

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

criteria based on SSPSS-2009, Revision 1

Engineering Department Standard 36180, Structural Monitoring Program, Revision 4

Loss of Fluid Evaluation of PMH 2000 Series Hydraulic Snubbers

Maintenance Support Evaluation 07MSE181, Remove Doors P801 and P802

NEI-99-02, Revision 7

Snubbers System Health Report

Technical Specification Section 3.7.7

Section 1R13: Maintenance Risk Assessments and Emergent Work Control

Procedures

ODI.101, Guarded Equipment Recommendations for Refueling Outages, Revision 13

OP-AA-102-1003, Guarded Equipment, Revision 4

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

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

Condition Reports

01954498
A-10

Attachment

Miscellaneous

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

Risk Assessment” 1-ED-X-2-A/B OOS for Main Generator Breaker Repairs While Entering 

Mode 4 from 5, Mode 3 from Mode 4, or Mode 2 from Mode 3

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

Section 1R15: Operability Determinations and Functionality Assessments

Procedures

EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 16

LX0563.02, Reactor Coolant Pump Undervoltage Channel Calibration and Relay PM, 

Revision 11

Condition Reports

01846345 01889301 01896874 01967934 01969615 01973026

01973578 01974039 1949876 1965480

Maintenance Orders/Work Orders

40288839 40245357

Miscellaneous

Adverse Condition Monitoring and Contingency Plan for Seat Leakage on 1-MS-V-393

Breaker on Higher Bus Voltages, Revision 5 

Calculation SBC-128, Technical Specifications – Setpoints and Allowable Values, Revision 15

DBD-ESF-01, Engineered Safety Features Response Times Design Basis Document, 

Revision 2

EC-280600, Change TAP Settings of Westinghouse Inverters to prevent Tripping AC Input

Maintenance Rule Functional Failure Evaluation, under AR 1949876

UFSAR Section 6.8, Emergency Feedwater System, Revision 15

FP-22849, Terry Turbine Instruction Manual, Revision 1

Engineering Evaluation EE-11-031, FPL – Seabrook Mitigating System Performance Indicator 

Basis Document, Revision 00

Section 1R18: Plant Modifications

Procedures

ON1435.05, Feed Pump Turbine Overspeed Trip Test, Revision 5

PR-AA-1008-F01, Owner’s Acceptance Review Checklist, Revision 0

Condition Reports

01957267 01961655

Miscellaneous

DBD-ED-04, 120 VAC Vital & Non-Vital Instrument Power Design Basis Document, Revision 2

DBD-FW-01, Feedwater System Design Basis Document, Revision 3

Engineering Change 250048, Revision 1

Feedwater System Health Report

Standing Order 14-002, Main Feed Pump Master/Slave Controller Auto/Manual Transfer 

Operation

UFCR 12-006, Steam Generator Feed Pump Turbine Digital Upgrade Project, Revision 1

UFSAR Section 7.1.2.5
A-11

Attachment

Drawings

1-NHY-503100 1-NHY-503226 1-NHY-503581

1-NHY-503590 1-NHY-503593 1-NHY-504138

1-NHY-504168 1-NHY-506484 1-NHY-509053

Section 1R19: Post-Maintenance Testing

Procedures

OX1413.08, RH-P-8B Comprehensive Pump Test, Revision 8

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

Test, Revision 19

OX1456.81, Operability Testing of IST Valves, Revision 19

OS1048.13, Vital Bus 11A Operation, Revision 10

OS1008.01, Chemical and Volume Control System Makeup Operations, Revision 33

ES1850.012, Air Operated Valve Program Procedure, Revision 4

MN0520.16, Copes-Vulcan Model D-100/1000 Operator Maintenance, Revision 2

MS0517.12, Application and Repair of Protective Coating(s), Revision 12

MS0517.43, Piping Installation and Maintenance, Revision 2

ES1807.025, Inservice Inspection (ISI) Visual Examination Procedure, Revision 5

Condition Reports

01799084 01970614 1956739 1953484 1973225

Maintenance Orders/Work Orders

40202481 40265447 40303546 40238124 40238125 40238501

40238143 40263623 40260904

Miscellaneous

IST Pump Data Sheet, RH-P-8B, 1-RH-OT-0007

IST Pump Data Sheet, SW-P41B, 1-SW-OT-008

IST Pump Data Sheet, SW-P41B, 1-SW-OT-038

IST Pump Data Sheet, SW-P41D, 1-SW-OT-008

IST Pump Data Sheet, SW-P41D, 1-SW-OT-038

Section 1R20: Refueling and Other Outage Activities

Procedures

AD-AA-101-1004, Work Hour Controls, Revision 14

ODI.82A, Mode Change Notice, Revision 19

ODI.82E, Mode Change Checklist, Revision 19

ON1090.04, Containment Entry, Revision 27

OS1000.01, Heatup From Cold Shutdown to Hot Standby, Revision 39

OS1000.02, Plant Startup From Hot Standby to Minimum Load, Revision 28

OS1000.03, Plant Shutdown From Minimum Load to Hot Standby, Revision 26

OS1000.04, Plant Cooldown From Hot Standby to Cold Shutdown, Revision 44

OS1000.05, Power Increase, Revision 25

OS1000.06, Power Decrease, Revision 18

OS1000.07, Approach to Criticality, Revision 13

OS1000.08, Post Trip Review, Revision 29

OS1000.09, Refueling Operation, Revision 28

OS1000.12, Operation With RCS at Reduced Inventory/Midloop Conditions, Revision 12
A-12

Attachment

OS1000.13, Operation With the Reactor Defueled, Revision 03

OS1000.14, Reactor Coolant System Evacuation and Fill, Revision 18

OS1001.01, Reactor Coolant System Fill and Vent, Revision 25

OS1001.02, Draining the Reactor Coolant System for Vessel Head Removal, Revision 17

OS1001.11, Reactor Coolant System Shutdown Level Instrumentation, Revision 9

OS1013.03, Residual Heat Removal Train A Startup and Operation, Revision 27

OS1015.10, Refueling Canal and Cavity Drain, Revision 17

OS1015.18, Setting Containment Integrity for Mode IV Entry, Revision 10

OS1016.11, Contingency Ocean Pump Restoration for SW Work Activities With Ocean Service

Water Pumps Not in Service, Revision 5

OX1406.12, 18 Month Containment and Containment Spray Recirculation Sump Surveillance, 

Revision 11

OX1436.13, Turbine Driven Emergency Feedwater Pump Post Cold Shutdown or Post 

Maintenance Surveillance and Comprehensive Pump Test, Revision 26

OS1246.01, Loss of Offsite Power Plant Shutdown, Revision 22

PEG-10, System Walkdowns, Revision 21

RS1737, Post Refueling Low Power Physics Testing, Revision 7

Condition Reports

1960185 1960175 1960177 1803194 1799753

Miscellaneous

OS1000.04, Form A, RCS and PZR Cooldown Log, Revision 45, from 4/1 to 4/2/14

PRAE-14-001, OR16 Outage Schedule Shutdown Risk Review, Revision 0

SY-AA-100-1011-F01, Fatigue Assessment Form, Revision 4

Work Order 40203281, Shop Testing of Reactor Coolant Pump Cartridge

Work Order 40202561, Containment and Containment Spray Recirculation Sump Surveillance

Section 1R22: Surveillance Testing

Procedures

CX0901.02, Determination of Dose Equivalent I-131, Revision 12

CS0910.01, Primary Systems Sampling at SS-CP-166A, Revision 18

EX1803.003, Reactor Containment Type B and C Leakage Rate Tests, Revision 12

EX1806.001, RPS And ESFAS Response Time Summation Procedure, Revision 07

EX1808.014, Containment Enclosure Emergency Exhaust Filter System 18 Month Surveillance, 

Revision 9

OS1023.66, Containment Enclosure Ventilation System Operation, Revision 17

OX1401.02, RCS Steady State Leak Rate Calculation, Revision 8

OX1401.09, Reactor Vent Paths Cold Shutdown And 18 Month Surveillance, Revision 11

OX1456.94, Train B Phase B, CBS, & CVI Actuation 18 Month Surveillance, Revision 05

OX1456.97, Train A Phase B, CBS, & CVI Actuation 18 Month Surveillance, Revision 02

OX1436.13, Turbine-Driven Emergency Feedwater Pump Post Cold Shutdown or Post 

Maintenance Surveillance and Comprehensive Test, Revision 26

OX1460.01, Form A, Weekly Personnel Hatch Air Lock Door Seal testing, Revision 11

Condition Reports (*NRC identified)

01673025 01673033 01954574* 01954972 01955131 01955370

01956011 01958150 01958882 1960453
A-13

Attachment

Maintenance Orders/Work Orders

40201086 40202286 40203229 40204623 40204624 40204884

40204886 40307194 40204750 40258847

Drawings

1-NHY-310900 SH

Section 2RS1: Radiological Hazard Assessment and Exposure Controls

Procedures 

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

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

HD0958.19, Evaluation of Dosimetry Abnormalities, Revision 37

HN0958.25, High Radiation Area Control, Revision 37

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

HN0960.10, Radiological Requirements for Entry Beneath the Reactor Vessel, Revision 26

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

RP-AA-101-1001, Personnel Monitoring Device Issue, Revision 0

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

Gradient Work, Revision 3

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

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

Audits, Self-Assessments, and Surveillances 

Assessment and ALARA Planning and Control, February 3, 2014Seabrook 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 

Condition Reports 

01836289 01855852 01955642 01957485

Miscellaneous

EPRI Standard Radiation Monitoring Program Data for RO 16, April 9, 2014

RWP 14-0004, OR16 Reactor Cavity Decontamination, April 1, 2014

Seabrook 2014 Air Sample Log, March 29 - April 15, 2014

Seabrook HRA/LHRA Briefing Acknowledgement Form, April 15, 2014

Seabrook LHRA In-Service Key Box Log (Containment Alternate Control Point), April 15, 2014

Seabrook LHRA In-Service Key Box Log (RP Access Control Point), April 15, 2014

Seabrook LHRA/VHRA Key Issue Log (Containment Alternate Control Point), April 15, 2014

Seabrook LHRA/VHRA Key Issue Log (RP Access Control Point), April 15, 2014

Seabrook Log of VHRA and LHRA Access Points, April 15, 2014

Seabrook RWP 14-0031, Rx Cavity Work during OR16, Includes Rx Head Lift/Set, April 1, 2014

Seabrook RWP 14-0032, Replace Reactor Head O-ring include Preparations, Decontamination 

and QC Inspections, April 11, 2014

Seabrook RWP 14-0038, OR16 Steam Generator Eddy Current Testing and Tube Plugging, 

April 16, 2014

Seabrook RWP 14-0052, Replace Lower Plenum, Replace Rx Ventilation Ducting UnderVessel, March 21, 2014

Seabrook Survey M-20140409-14, CTB S/G A Initial Survey, April 9, 2014
A-14

Attachment

Seabrook Survey M-20140409-16, CTB S/G B Initial Survey, April 9, 2014

Seabrook Survey M-20140409-18, CTB S/G D Initial Survey, April 9, 2014

Seabrook Survey M-20140409-19, CTB S/G C Initial Survey, April 9, 2014

Seabrook Survey M-20140415-8, CTB-00 Cavity Reactor Cavity with Head Raised 18” Above 

Flange, April 15, 2014

Seabrook Survey M-20140416-1, CTB-00 Cavity Reactor Cavity with Head on Flange After 

Cavity Decon, April 16, 2016

Seabrook Updated Final Safety Analysis Report

Section 2RS2: Occupational ALARA Planning and Controls

Procedures 

RP-AA-104 ALARA Program, Revision 2

RP-AA-104-1000, ALARA Implementing Procedure, Revision 3

Audits, Self-Assessments, and Surveillances 

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

Programs, February 24, 2014

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

Assessment and ALARA Planning and Control, February 3, 2014

Condition Reports 

01953467

Miscellaneous

EPRI Standard Radiation Monitoring Program Results through OR15, September 25, 2012

Job In-Progress ALARA Review AR No. 14-JIP-01, Reactor Vessel Dissassembly & 

Reassembly 25 percent Review, April 12, 2014

Job In-Progress ALARA Review AR No. 14-JIP-02, Reactor Vessel Dissassembly & 

Reassembly 75 percent Review, April 17, 2014

Job In-Progress ALARA Review AR No. 14-JIP-03, OR16 Steam Generator Eddy Current 

Testing and Tube Plugging - 25 percent Review, April 15, 2014

Job In-Progress ALARA Review AR No. 14-JIP-08, OR16, Valve Maintenance- 50 percent 

Review, 

April 14, 2014

Job In-Progress ALARA Review AR No. 14-JIP-09, OR16 Fuel Handling Project – 50 percent 

Review, 

April 13, 2014

Job In-Progress ALARA Review AR No. 14-JIP-16, Replace Lower Plenum. Replace Rx 

Ventilation Ducting Under-Vessel, April 14, 2014

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, OR16 In Service Inspection, Febuary 25, 2014

Pre-Job ALARA Review Package 14-07, OR16 Fuel Handling Project, Febuary 25, 2014

Pre-Job ALARA Review Package 14-09, OR16 RCP Seal Replacement, Febuary 25, 2014

Pre-Job ALARA Review Package 14-10, OR16 Scaffolding, Febuary 25, 2014

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

New Design, Febuary 25, 2014

Seabrook Temporary Sheilding Log for OR16, March 2014
A-15

Attachment

Section 2RS3: In-Plant Airborne Radioactivity Control and Mitigation

Procedures 

HD0955.01, Analysis of Smears & Air Samples, Revision 28

HD0955.53, Use of AMS-4, Revision 2

HD0955.71, Setup of ASI Breathing Air System, Revision 0

HD0958.01, Air Sampling, Revision 12

HD0965.01, Respiratory Protection Quality Assurance and Maintenance, Revision 19

HD0965.07, Air Supplied Respiratory Protection Equipment and Distribution System,

Revision18

HD0965.12, Respiratory Equipment Issue and Use, Revision 38

OS1023.68, Containment Air Purge Operation, Revision 19

Audits, Self-Assessments, and Surveillances 

SFA 01815118, Perform Quick Hit Self-Assessment - Respiratory Protection Program, 

April 5, 2013 

SFA 01868045, 2012 Respiratory Protection Program, September 17, 2013

Condition Reports 

01880090 01953797 01956903 01957319 01957775

Miscellaneous

Air Sample Result 14-0338, Containment B Steam Generator General Area Platform, 

April 15, 2014

Air Sample Result 14-179, Containment El 26’ Decon Loop one, April 3, 2014

Air Sample Result 14-206, Containment Cavity during Cavity Flood, April 5, 2014

Air Sample Result 14-266, Containment D Steam Generator Hot Leg Inside Manway, 

April 9, 2014

HEPA Portable Ventilation Unit Inventory, April 14, 2014

HPSTID 14-002, Air Systems Breather Box Testing with V4F1 R Delta Suit, March 19, 2014

Respirator Issue Log, April 10-17, 2014

Section 2RS4: Occupational Dose Assessment

Procedures 

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

HD0992.02, Issuance and Control of Personnel Monitoring Device, Revision 38

HN0958.39, Multi-Badge Control and Exposure Tracking, Revision 35

HN0958.42, Determination and Control of Dose to an Embryo Fetus, Revision 25

HN0961.29, Internal Dosimetry Assessment, Revision 27

RP 5.3, Expected or Declared Pregnant Worker Exposure Control, Revision 4

RP-AA-101-2004, Method for Monitoring and Assigning Effective Dose Equivalent (EDE) for

High Dose Gradient Work, Revision 2

Audits, Self-Assessments, and Surveillances 

SFA 01892653 Self Assessment Quick Hit on Electronic Dosimeter Setpoints, 

September 19, 2013

Condition Reports 

01896167 01903346 01951842 01952711 01954641 01956098

01956989 01958425
A-16

Attachment

Miscellaneous

HPD0958.03 Form A - Personnel Contamination Report, Westinghouse Worker Right Hand 

April 10, 2014

HPD0958.03 Form A - Personnel Contamination Report, Master Lee Worker Left and Right 

Forearm, April 11, 2014 

HPD0958.03 Form A - Personnel Contamination Report, Operation’s Worker Right Shoe, 

April 11, 2014

HPD0958.03 Form A - Personnel Contamination Report, Bartlett RP Technician Stomach, 

April 13, 2014

HPD0958.03 Form A - Personnel Contamination Report, Master Lee Worker Chin, 

April 11, 2014

RP 5.1A - Increased Radiation Exposure Request and Authorization, Day Zimmerman Worker 

April 1, 2014

RP 5.1A - Increased Radiation Exposure Request and Authorization, Westinghouse Worker, 

April 8, 2014

RP 5.3 Expected or Declared Pregnant Worker Exposure Control, Attachment A - Declaration of 

Pregnancy, August 28, 2012

RP 5.3 Expected or Declared Pregnant Worker Exposure Control, Attachment B – Statement of 

Expected Pregnancy, August 28, 2012

RP 5.3 Expected or Declared Pregnant Worker Exposure Control, Attachment C – Radiological 

Controls for Declared Pregnant Workers, August 28, 2012

Section 2RS5: Radiation Monitoring Instrumentation

Procedures

RP SOPs

HD0955.03, Use Tennelec APC 175, Revision 9

HD0955.05, Operation Portable Rad Mont Instruments, Revision 21

HD0955.19, Use of Sheppard Model 81 Beam Calibrator, Revision 12

HD0955.30, Use Mini Buck Calibrator, Revision 6

HD0955.39, Use Sheppard Model 89 Box Calibrator, Revision 1

HD0955.42, Operation SAM and Chronos Monitors, Revision 7

HD0955.50, Far West REM-500 Operation, Revision 5

HD0955.53, Use of AMS-4, Revision 4

HD0955.54, Operation TSA Model SPM 906 Portal Monitor, Revision 1

HD0955.62, Use Argos 4AB, Revision 3

HD0955.63, Use Sirius 2 Hand and Foot Counter, Revision 1

HD0955.64, Use MGP DRM 1 and 2 Area Rad Monitors, Revision 5

HD0955.69, Use of GEM 5 Portal Monitor, Revision 2

HD0957.01, Calibration Environmental Air Samplers, Revision 8

HD0958.01, Air Sampling, Revision 14

HD0958.38, Evaluation of 10CFR61 Isotopic Mix, Revision 29

HD0961.31, Canberra Whole Body Counting System Operation, Revision 10

HD0961.32, Canberra WBC Calibration, Revision 1

HD0961.34, Canberra FASTSCAN WBC Operation, Revision 8

HD0963.03, Calibration Eberline E140N Ratemeter, Revision 5

HD0963.08, Calibration Air Sampling Equipment, Revision 15

HD0963.20, Calibration DCA Area Rad Monitor, Revision 2

HD0963.24, Calibration Johnson Extender Teletector, Revision 4

HD0963.27, Calibration Eberline RO2 Bicron RSO5 Ion Chamber, Revision 8
A-17

Attachment

HD0963.28, Calibration MGP DMC 2000 E Dosimeters, Revision 16

HD0963.30, Calibration RO 7 Ion Chamber, Revision 6

HD0963.31, Calibration of Eberline RM-14, Revision 7

HD0963.34, Calibration PNR 4 Neutron Rem Counter, Revision 6

HD0963.44, Calibration of the Bicron MicroRem Meter, Revision 2

HD0963.45, Calibration AMS-4 CAM, Revision 1

HD0963.46, Calibration TSA Model SPM-906 Portal Mont, Revision 3

HD0963.48, Calibration of the MGP AMP-100 AMP-200, Revision 1

HD0963.51, Calibration Argos 4AB, Revision 5

HD0963.52, DRM 2 Area Rad Mont Calibration, Revision 1

HD0963.53, Calibration Sirius 2 Hand Foot Counter, Revision 5

HD0963.54, Calibration Fluke 451B Ion Chamber, Revision 0 

HD0963.55, Calibration Fluke 451P Ion Chamber, Revision 1

HD0963.57, Calibration MGP Telepole, Revision 2

HD0963.58, Calibration SAM 12, Revision 2

HD0963.60, Calibration Canberra Chronos 4, Revision 2

HD0963.61, Calibration and maintenance of RADECO HD-29A and AVS-28A Air Sampler 

Pump, Revision 1

HD0963.62, Calibration of Canberra GEM-5, Revision 2

HD0963.64, Calibration Ludlum 177, Revision 0

HD0963.65, Calibration Model 9-7 Ion Chamber, Revision 0

HD0963.56, Calibration Canberra S5-APC-GM, Revision 1

HN0955.08, Operation RDMS Continuous Air Monitor, Revision 9

HX0955.32, RDMS Setpoint Determination Rad Monitors, Revision 29

IC SOPs

IN1660.601, Dual Channel Calibration ARM 6508, 6517, 6536, 6563, Revision 6

IN1660.604, Single Channel Calibration ARM 6518, 6529, 6540, Revision 6

IN1660.611, RD 10B RD 12 Calibration ARM 6534, 6537, 6550, Revision 6

IN1660.622, Non Safety Related Area Rad Monitors Calibration, Revision 7

IN1660.714, RM 6522, 6531 PAB WPB CAM Calibration, Revision 3

IN1660.731, RM6495 Plant Vent Mid/Hi Range Rad Monitor Calibration, Revision 3

IN1660.812, RM-R-6502 Carbon Delay Beds Inlet Radiation Monitor Calibration, Revision 6

IN1660.813, RM-R-6503 WG Compressors Inlet Radiation Monitor Calibration, Revision 5

IN1660.817 RM 6510, 6511, 6512, 6513 SG Blowdown Calibration, Revision 8

IN1660.990, RM 6486, 6487, 6488, 6489 Portable Continuous Atmosphere Radiation Monitor 

Calibration, Revision 6

IN1660.992, RM-R-6454 Storm Drain Effluent Monitor Calibration, Revision 5 

IS1660.120, RM-F-6497 Plant Vent Stack Accident Sample Flow Control Calibration, Revision 7

IX1660.110, Plant Vent Stack Flow Transmitter Calibration, Revision 7

IX1660.612, RM-R-6535 A B Manipulator Crane ARM Calibration, Revision 9

IX1660.639, RM 6576A, 6576B Containment Hi Range Rad Mont Calibration Revision 10

IX1660.662, RM-R-6535-A Fuel Manipulator Crane Train A ARM Operation Test, Revision 8

IX1660.663, RM-R-6535 Fuel Manipulator Crane Train B ARM Operation Test, Revision 8

IX1660.689, RM-R-6576-A Containment Hi Range Rad Monitor Operation Test, Revision 9

IX1660.690, RM-R-6576-B Containment Hi Range Rad Mont Operation Test, Revision 8

IX1660.710, RM-R-6506 6507 Control Room Air Intake A B Rad Mont Calibration, Revision 9

IX1660.718, RM 6526 Containment Rad Mont Calibration, Revision 12

IX1660.719, RM-R-6548 Containment Rad Monitor Calibration, Revision 8
A-18

Attachment

IX1660.720, RM-R-6527 COP Trains A B Rad Mont Calibration, Revision 8

IX1660.724, RM-6562 Fuel Storage Bldg Airborne Rad Mont Calibration, Revision 6

IX1660.730, RM-R-6528 Plant Vent Wide Range Gas Monitor Calibration, Revision 9

IX1660.760, RM-R-6506-A Control Room East Air Intake Operation Test, Revision 7

IX1660.761, RM-R-6506-B Control Room West Air Intake Operation Test, Revision 7

IX1660.762, RM-R-6507-A Control Room West Air Intake Operation Test, Revision 7

IX1660.763, RM-R-6507-B Control Room West Air Intake Operation Test, Revision 7

IX1660.768, RM-R-6526 Containment Atmosphere Operation Test, Revision 9

IX1660.769, RM-R-6548 Containment Atmosphere Backup Operation Test, Revision 6

IX1660.770, RM-R-6527-A Containment On-Line Purge Operation Test, Revision 8

IX1660.771, RM-R 6527-B Containment On-Line Purge Operation Test, Revision 9

IX1660.774, RM-R-6562 Fuel Storage Building Ventilation Exhaust Operation Test, Revision 6

IX1660.780, RM-R-6528 Plant Vent Wide Range Gas Monitor Operation Test, Revision 7

IX1660.801, RM-R-6481 6482 Main Steam Line Rad Mont Calibration, Revision 7

IX1660.814, RM-R-6504 Waste Gas Compressor Rad Monitor Calibration, Revision 7

IX1660.814, RM-6503 WG Compressors Discharge Radiation Monitor Calibration, Revision 7

IX1660.815, RM-R-6505 Condenser Air Ejector Discharge Rad Mont Calibration, Revision 2

IX1660.816, RM-R-6509 WLTT Discharge Rad Mont Calibration, Revision 8

IX1660.816, RM-6509 Waste Liquid Test Tanks Discharge Radiation Monitor Calibration, 

Revision 7

IX1660.823, RM-R-6515 6516 Loop A B PCCW Rad Mont Calibration, Revision 6

IX1660.824, RM-R-6519 SGBD Flash Tank Discharge Rad Mont Calibration, Revision 9

IX1660.826, RM-R-6521 Turbine Bldg Sump Rad Mont Calibration, Revision 6

IX1660.864, RM-R-6504 WG Compressor Discharge Operation Test, Revision 8

IX1660.872, RM-R-6516 Loop A PCCW Operation Test, Revision 8

IX1660.873, RM-R-6515 Loop B PCCW Operation Test, Revision 8

IX1660.874, RM-R-6519 SB Flash Tank Discharge Operation Test, Revision 7

IX1660.876, RM-R-6521 Turbine Building Sump Pump Discharge Operation Test, Revision 6

Chem SOPs

CS0920.07, Tritium Analysis Liquid Scintillator, Revision 14

CX0917.01, Liquid Effluent Release Setpoints, Revision 20

HD0963.38, Calibration of Ludlum 220 Portable Scaler Ratemeter, Revision 5

HD0963.47, Tennelec Series 5 XLB Smear Counter Calibration, Revision 1

IX1660730, RM-R-6528 Plant Vent Wide Range Gas Monitor Calibration, Revision 9

JS0999.200, Operation of Countroom Analysis System, Revision 3

JS0999.300, Calibration of Gamma Spectroscopy Detectors Using the Count Analysis System, 

Revision 2

Audits, Self-Assessments, and Surveillances 

Nuclear Oversight Report SBK 14-001, Audit of Radiation Protection and Radwaste, Feb 24, 

2014

Quick Hit 1930254 Instruments QHSA-2, December 31, 2013

Quick Hit 1961974, NRC 71124.05 Radiation Monitoring Instrumentation, May 7, 2014

Condition Reports 

01767528 01833664 01847095 01856452 01879999 01939604

01941338 01947933 01948312 01961385 01967570 01969397
A-19

Attachment

Miscellaneous

2013 2nd Quarter RM System Health Report, June 30, 2013

2013 3rd Quarter RM System Health Report, September 30, 2013

2013 4th Quarter RM System Health Report, December 31, 2013

2014 1st Quarter RM System Health Report, March 31, 2014

Canberra Argos-4AB SN 0503-107 Calibration Record Printout, January 30, 2014

Canberra Sirius-2B SN 0607-021 Calibration Record Printout, March 20, 2014

Canberra Sirius-2B SN 0607-021 Calibration Record Printout, September 4, 2012

EC 281092, Wide Range Gas Monitor: Replacement of Thomas Model 727CM39 Air Sample 

Pump with Metal Bellows Model MB-602, May 2014

HD0963.27 Form A RO-2A Ion Chamber Calibration Record SN 3713, October 29, 2013

HD0963.27 Form A RO-2A Ion Chamber Calibration Record SN 3713, December 1, 2010

HD0963.45 Form A AMS-4 Flow Calibration Record SN 1181, January 25, 2013

HD0963.45 Form A AMS-4 Flow Calibration Record SN 1181, January 30, 2013

HD0963.45 Form A AMS-4 Flow Calibration Record SN 991, January 17, 2014

HD0963.45 Form A AMS-4 Flow Calibration Record SN 991, January 29, 2013

HD0963.45 Form B - AMS-4 Calibration Data Serial No. 1181, December 30, 2013

HD0963.45 Form B AMS-4 Calibration Data SN 1181, January 30, 2013

HD0963.48 Form B AMP-200 Calibration Sheet SN 7702-001, March 7, 2014

HD0963.48 Form B AMP-200 Calibration Sheet SN 7702-001, July 26, 2012

HD0963.54 Form A Fluke Biomedical 451B Ion Chamber Calibration Record SN 0038, 

January 17, 2014

HD0963.54 Form A Fluke Biomedical 451B Ion Chamber Calibration Record SN 0038, 

August 10, 2012

HD0963.60 Form A Chronos-4 Calibration Record SN 1009-058, March 20, 2014

HD0963.60 Form A Chronos-4 Calibration Record SN 1009-058, September 10, 2012

HD0963.61 Form A HD-29A Air Sampler Calibration Record SN 2253, March 10, 2014

HD0963.61 Form A HD-29A Air Sampler Calibration Record SN 2253, March 14, 2013

Health Physics Instruments - Product Application Literature PAL- REM-500 Calibration 

Recommendation, August 2010

Health Physics Instruments- Operations and Repair Manual: Model REM-500 Neutron Survey 

Meter Revision A, September 1998

HPSTD-96-004, Basis for Performing Operability Checks for Use of the Far West Technology 

REM-500 Neutron Meter, February 10, 1996

HPSTID 05-007, Primary Calibration of RD-64-52 ATL Effluent Radiation Monitor RM-6473, 

6509, and 6521, September 26, 2011

HPSTID 11-002 Additional Information on Containment Atmosphere Radiation Monitor Setpoint 

Basis, March 9, 2011

HPSTID 13-003, Calibration of WBC Systems – 2013, March 19, 2014

HPSTID 13-008, Verification/Calibration on the Shepherd Model 81-12 (Serial No. 7015) Cs-137 

Irradiator, October 28, 2013

HPSTID 14-005, 2013 10 CFR Part 61 Data, December 2013

HPSTID 14-007, Calibration of WBC Systems – 2014, March 19, 2014

SB Offsite Dose Calculation Manual, Revision 35

SB Updated Final Safety Analysis Report, Revision 8A

Seabrook Station Maintenance Rule (a)(1) Improvement Plan for Radiation Monitoring RM-03B 

Subsystem Airborne Monitoring Instruments, Revision 3

Seabrook Station Radiation Protection Manual, Revision 64
A-20

Attachment

Maintenance Orders/Work Orders

40086692 40092985 40104805 40107205 40111868 40114187

40122692 40122999 40162465 40194413 40205154 40215550

40218071 40235509 40235518 40235521 40235526 40238054

40263637

Section 4OA1: Performance Indicator Verification

Procedures

CS0910.01, Primary Systems Sampling at SS-CP-166A, Revision 18

CX0901.02, Determination of Dose Equivalent I-131, Revision 12

NAP-206, NRC Performance Indicators, Revision 6

OX1401.02, RCS Steady State Leak Rate Calculation, Revision 8

Miscellaneous

LIC-13026, Documentation Supporting the Seabrook Station NRC 2nd Quarter 2013 

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

LIC-14018, Documentation Supporting the Seabrook Station NRC 1st Quarter 2014 

Performance Indicator Submittal

Technical Specification, Section 3.4.8

Section 4OA2: Problem Identification and Resolution

Condition Reports

1966262 1963207 1949610 1974681 1974682 1894548

1965182 1965183 1952682

Miscellaneous

Seabrook Station Self-Evaluation and Trending Analysis Report for 1st Quarter 2014 

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

Procedures

EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 16

LX0563.02, Reactor Coolant Pump Undervoltage Channel Calibration and Relay PM, 

Revision 11

ON1046.12, Operation of the Main Generator Breaker, Revision 21

OS1000.08, Post Trip Review, Revision 19

PI-AA-103-1001-F01, Human Performance Review Worksheet, Review 2

Condition Reports

01953543 01955993 01956944 01956945 01969615 01930049

01900249 01904703

Miscellaneous

Calculation SBC-128, Technical Specifications – Setpoints and Allowable Values, Revision 15

DBD-ESF-01, Engineered Safety Features Response Times Design Basis Document, 

Revision 2

LER 2014-001-000
A-21

Attachment

Reactor Trip Due to RCP Loop Low Flow Low Maintenance Rule Functional Failure Evaluation

Work Order 40260904-03, Repair of Degraded Pipe Wall on SW-1802-4 per EC280429, 

completed on April 15, 2014. 

LIST OF ACRONYMS

AC alternating current

ACI American Concrete Institute

ADAMS Agencywide Document Access and Management System

ALARA as low as reasonably achievable 

AR action request

ARM area radiation monitor

ASME American Society of Mechanical Engineers

BACC Boric Acid Corrosion Control Program

CAM continuous air monitor

CAP corrective action program

CFR Code of Federal Regulations

CR condition report

CRDM control rod drive mechanism

EC eddy current

ECT eddy current testing

EDG emergency diesel generator

EDY effective degradation years

EFW emergency feedwater

EPD electronic personal dosimeter

EPRI Electric Power Research institute

HRA high radiation area

IMC Inspection Manual Chapter

kV kilovolt

LER licensee event report

LHRA locked high radiation area

MR maintenance rule

MSHA Mine Safety and Health Administration

NCV non-cited violation

NDE nondestructive examination

NEI Nuclear Energy Institute

NIOSH National Institute for Occupational Safety and Health

NIST National Institute of Standards and Technology

NRC Nuclear Regulatory Commission

NVLAP National Voluntary Laboratory Accreditation Program

ODCM offsite dose calculation manual

ODSCC Outside Diameter Stress Corrosion Cracking

OOS out of service

OR Refueling Outage

PCCW primary component cooling water

PAB primary auxiliary building
A-22

Attachment

RAT reserve auxiliary transformer

RCP reactor coolant pump

RCS reactor coolant system

REMP radiological environmental monitoring program

RG Regulatory Guide

RHR residual heat removal

RIY re-inspection year

RP radiation protection

RPS reactor protection system

RVCH reactor vessel upper closure head

RWP radiation work permit

SG steam generator

SSC structure, system, or component

SW service water

TS technical specification

UAT unit auxiliary transformer

UFSAR Updated Final Safety Analysis Report

UT ultrasonic testing

VHRA very high radiation area

VT visual examination

VUHP vessel upper head penetration

WBC whole body count

WO work order