UNITED STATES

N UCLEAR REGULATORY COMMISSION

REG]ON I

475 ALLENDALE ROAD

KING OF PRUSSIA. PA 19406-1415

August 72, 2OIL

Mr. Paul Freeman

Site Vice President

Seabrook Nuclear Power Plant

NextEra Energy Seabrook, LLC

c/o Mr. Michael O'Keefe

P.O. Box 300

Seabrook, NH 03874

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

REPORT 05000443/201 1 003

Dear Mr. Freeman:

On June 30, 201 1, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at

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

on July 13,2011, with Mr. E. Metcalf and other members of your statf.

These inspections 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.

The report documents three NRC-identified findings of very low significance (Green) that were

determined to involve a violation of NRC requirements. However, because of the very low

safety significance and because the issues were entered into your corrective action program,

the NRCis treating the findings as non-cited violations (NCV) consistent with Section 2.3.2.a of

the NRC Enforcement Policy.

lf you contest 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 l; the Director, Office of Enforcement, United States Nuclear Regulatory

Commission, Washington, DC 20555-0001; and the NRC Resident lnspector at the Seabrook

Station. In addition, if you disagree with the characterization of any finding in this report, you

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

your disagreement, to the Regional Administrator, Region l, and the NRC Resident lnspector at

the Seabrook Station. The information you provide will be considered in accordance with

Inspection Manual Chapter 0305.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its

enclosure, and your response (if any), will be available electronically for public inspection in the
P. Freeman 2

NRC Public Document Room or from the Publicly Available Records (PARS) component of

NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at

http://www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely, I-t- / it rfi4

t lrV

Arthur L. Burritt, Chief

Projects Branch 3

Division of Reactor Projects

Docket No. 50-443

License No: NPF-86

Enclosure: lnspection Report No. 050004431201 1003

wi Attachment: Supplemental Information

cc w/encl: Distribution via ListServ
P. Freeman 2

NRC Public Document Room or from the Publicly Available Records (PARS) component of

NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at

http://www.nrc.oov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Arthur L. Burritt. Chief

Projects Branch 3

Division of Reactor Projects

Distribution dencl: (via e-mail)

W. Dean, RA (RIORAMA|L Resource)

D. Lew, DRA (RIORAMA|L Resource)

D. Roberts, DRP (RIDRPMA|L Resource)

J. Clifford, DRP (RIDRPMA|L Resource)

C. Miller, DRS (RlDRSMail Resource)

P. Wilson, DRS (Rl DRSMail Resource)

A. Burritt. DRP

L. Cline, DRP

A. Turilin, DRP

C. Douglas, DRP

W. Raymond, DRP, SRI

J. Johnson, DRP, Rl

A. Cass, DRP, Resident OA

J. McHale, Rl OEDO

RidsN rrPMSeabrook Resource

RidsNrrDorlLpl 1 -2 Resource

ROPreports Resource

SUNSI Review Gomplete: ALE (Reviewer's Initials)

DOCUMENT NAME: G:\DRP\BRANCH3\INSPECTION\REPORTS\ISSUED\201 1 (ROP 12)\SEA1 103,DOCX

After declaring this document "An Official Agency Record" it will be released to the Public.

To receive a copy of this document, indicate in the box:"C" = Copy without altachmenUenclosure "E" = Copy with attachmenUenclosure

"N" = No copy

ML112241543

CFFICE thp RI/DRP RI/DRP, RI/DRP

NAME WRavmond/alb for LCline/alb for ABurritUalb

DATE 08t12t11 08t12t11 08t12111

OFFICIAL RECORD COPY
Report No.:

Facility:

Location:

Dates:

Inspectors:

Approved by:

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

NPF-86

05000443/201 1 003

NextEra Energy Seabrook, LLC

Seabrook Station, Unit No.1

Seabrook, New Hampshire 03874

April 1 ,2011through June 30, 2011

W. Raymond, Senior Resident Inspector

J. Johnson, Resident Inspector

T. Moslak, Health Physicist

A. Turilin, Project Engineer

J. DeBoer, Reactor Engineer

T. Burns, Reactor lnspector

Arthur Burritt, Chief

Projects Branch 3

Division of Reactor Projects

Enclosure
2

TABLE OF CONTENTS

SUMMARY OF FIND1NGS............ .........3

REPORT DETATLS ...............5

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

1R01 Adverse Weather Preparation .........'5

1R04 Equipment Alignment. ......."......'....'.6

1R05 Fire Protection ........... ............'....'."'7

1R07 Heat Sink Performance............... ..............."'....9

1R08 Inservice Inspection .....'10

1R1 1 Licensed Operator Requalification Program.............. .""..11

1R12 Maintenance Effectiveness......... ......"'...'...'."12

1R13 Maintenance Risk Assessments and Emergent Work Control...... .....12

1R15 Operability Evaluations ...".............13

1R18 Plant Modifications .......19

1R19 Post-Maintenance Testing "'-."".'..21

1R20 Refueling and Outage Activities "".21

1R22 Surveillance Testing ."..24

2. RADIATION SAFETY ...................25

2RS01 Radiological Hazard Assessment and Exposure Controls.... .............25

2RS02 OccupationalALARA Planning and Controls .............. "'.'27

2RSO3 In-Plant Airborne Radioactivity Control and Mitigation ............ "'..'....29

2RS04 Occupational Dose Assessment ............. .....'30

4. OTHER ACTIV|TIES.............. ......31

4OA2 ldentification and Resolution of Problems............... ......".31

4OA5 Other Activities... '..'..'...'33

4OAO Meetings, Including Exit........... "....33

ATTACHMENT: SUPPLEMENTAL INFORMATION ......".'..'33

SUPPLEMENTAL INFORMATION .......... ........... A-1

KEY pOtNTS OF CONTACT ............. A-1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED..... ................ A-2

LIST OF DOCUMENTS REVIEWED ........... ....... A-3

Llsr oF ACRONYMS".'."'..."" """' A-11

Enclosure
3

SUMMARY OF FINDINGS

lR 0500044312011003; 0410112011-0613012011; Seabrook Station, Unit No. 1; Routine

lntegrated Report; Fire Protection; Operability Evaluations.

The report covered a three-month period of inspection by resident and regional specialist

inspectors. Three Green findings were identified. The significance of most findings is indicated

by their color (Green, White, Yellow, or Red) and determined using Inspection Manual Chapter

(lMC) 0609, "significance Determination Process" (SDP). The cross cutting aspect of a finding

is determined using the guidance in IMC 0310, "Components Within the Cross-Cutting Areas."

Findings for which the SDP does not apply may be Green or be assigned a severity level after

NRC management review. The NRC's program for overseeing the safe operation of

commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"

Revision 4, dated December 2006.

Gornerstone: Mitigating Systems

Green. The inspectors identified a non-cited violation (NCV) of Technical Specification

tt.Sl O.Z.t.h, which requires that written procedures be established and implemented for

the fire protection program. Contrary to TS 6.7.1.f , the inspectors identified combustible

materials which were not controlled per fire protection procedure FP 2.2, Revision 12'

Specifically, (i) combustible materials were stored within three feet of an energized

sample panel in the primary auxiliary building room PB404, a PRA risk significant area;

and, (ii) combustible materials in excess of the permissible amounts were stored in

waste process building area W8505. The in$pectors identified materials stored in

WB505 in excess of FP 2.2 limits on three occasions. Collectively, the NRC

observations indicate a weakness in the programmatic control of combustible materials

despite the fact that in each case the combustible materials were promptly removed

following identification by the inspector. Seabrook entered this performance deficiency

into their corrective action program.

The performance deficiency was more than minor because, if left uncorrected,

inadequate control of combustibles could affect the Mitigating Systems cornerstone

objective to assure external factors (fires) do not impact the availability and reliability of

syitems which mitigate events. The inspectors assessed the finding using Appendix F of

the Significance Determination Process (SDP) Based on a degradation rating of low,

which screens to Green in the fire protection SDP, the finding is of very low safety

significance. This finding has a cross-cutting aspect in Human Performance, Work

Piactices tH.4(b)l because Seabrook personnel did not follow procedures for the control

of transient combustibles. (Section 1R05)

Green. The inspectors identified a non-cited violation (NCV) of Technical Specification

(TS) 6.7.1.a that requires written procedures be established and implemented, including

administrative procedures that define authorities and responsibilities for safe operation

with respect to operability determinations. Contrary to TS 6.7.1.a, NextEra identified a

degraded and nonconforming condition related to reduced modulus of elasticity for

buiidings housing safety related equipment on May 27,2011 but did not complete an

operability determination until EC250348 was issued on June 28,2011 (AR1664399).

The delayed entry of the issue into the corrective action process to assess operability

was contrary to Section 4.3 of EN-AA-203-1001 that requires operability assessments be

Enclosure
4

completed in a time frame commensurate with the safety significance of the issue (within

8 hours). Seabrook subsequently completed an evaluation of the concrete issues and

determined that the buildings housing safety-related equipment remained operable.

Seabrook entered this performance deficiency into their corrective action program.

The performance deficiency was more than minor because a reasonable doubt of

operability for the affected concrete structure$ existed until further engineering

evaluations were completed to demonstrate the structures and systems that they housed

would remain functional under design and licensing basis conditions. The finding

affected the Mitigating Systems cornerstone Objective to ensure the availability, reliability

and capability of systems that respond to initiating events in order to prevent core

damage. The issue was evaluated using IMC 0609, "Significance Determination

Process" (SDP), and was determined to be of very low safety significance (Green)

because the finding was not a design or qualification deficiency, did not result in an

actual loss of safety function, was not a loss of a barrier function, and was not potentially

risk significant for external events. The finding had a cross cutting aspect in the area of

problem identification and resolution, P.1(a), because NextEra did not enter identified

degraded concrete conditions for several site buildings into the corrective actions

process in a timely manner, which would have ensured the shift manager completed

timely operability evaluations for the affected structures. (Section 1R15.3)

Green. The inspectors identified a non-cited violation (NCV) of Technical Specification

(TS) 6.7.1.a that requires that written procedures be established and implemented,

including administrative procedures that define authorities and responsibilities for safe

operation with respect to operability determinations. Contrary to TS 6.7 .1.a, NextEra

identified a degraded condition related to seryice water flow to the B emergency diesel

generator (EDG) heat exchanger (HX) on June 28,2011 but did not fully evaluate the

reduced flow under all plant conditions as required by NextEra procedure EN-AA-203-

1001. Fouling of the heat exchanger tubes was subsequently identified and mitigated.

Seabrook also completed an evaluation of the B EDG service water flow issues and

determined that the EDG remained operable. Seabrook entered this performance

deficiency into their corrective action program.

The performance deficiency was more than rninor because a reasonable doubt of

operability existed untilfurther engineering evaluations were completed to demonstrate

adequate service water flow to the B EDG HX existed and the B EDG remained

functional under design and licensing basis conditions. The finding affected the

Mitigating Systems cornerstone objective to ensure the availability, reliability and

capability of systems that respond to initiating events in order to prevent core damage.

The issue was evaluated using IMC 0609, "significance Determination Process" (SDP),

and was determined to be of very low safety significance (Green) because the finding

was not a design or qualification deficiency, did not result in an actual loss of safety

function, was not a loss of a barrier function, and was not potentially risk significant for

external events. The finding had a cross cutting aspect in the area of problem

identification and resolution, P.1(c), because NextEra personnel did not thoroughly

assess EDG operability to assure reduced HX SW flow was acceptable under all

operating conditions, or assure appropriate corrective actions were timely completed.

(Section 1R15.4)

Enclosure
5

REPORT DETAILS

Summarv of Plant Status

The plant was shutdown at the start of the report period to conduct refueling outage OR14.

NextEra completed reactor refueling and maintenance activities on the reactor and plant

secondary systems. The plant was started up and the reactor was taken critical on May 23,

2011, and operation at full power resumed on May 26, 2011. The turbine was taken offline for

maintenance on the secondary plant on June 4, 2011. Seabrook returned to full power on June

6,2011.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Preparation (71111.01 - 2 sample)

.1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspector completed one seasonal extreme weather conditions inspection sample.

The inspectors assessed NextEra's readiness for the onset of hot weather. The

inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) descriptions for

related design features and verified the adequacy of the station procedures for hot

weather protection. The inspectors reviewed NextEra's actions per procedure

ON1490.09 for seasonal readiness, and procedure OS1200.03 for severe weather. The

inspectors also performed walkdowns of susoeptible systems, specifically the

emergency feedwater, electrical distribution and service water systems. The inspectors

reviewed deficiencies related to extreme weather preparation and verified the issues

were entered into the corrective action program. The documents reviewed are listed in

the Attachment.

b. Findinqs

No findings were identified.

.2 Readiness of Offsite and Alternate AC Power Svstems

a. Inspection Scope

The inspectors completed one summer readiness of offsite and alternate AC power

systems inspection sample. The review focu$ed on NextEra procedure 051246.02,

"Degraded Vital AC Power." The inspectors verified that plant features were maintained

and procedures for operation were adequate to ensure the continued availability of AC

power systems. The inspectors verified that communication protocols with the

transmission system operator were adequate to ensure that appropriate information was

exchanged when issues arose that could impact the offsite power system. The

inspectors also observed NextEra's implementation of OS1246.02 during periods that

Enclosure
6

challenged grid conditions between April 1, 2011 and June 30, 2011. The inspection

included walkdowns of the onsite normal and emergency AC power systems and the

inspectors reviewed deficiencies related to summer readiness of offsite and alternate AC

power systems and verified these issues were entered into the corrective action

program. The documents reviewed are listed in the Attachment.

b. Findinqs

No findings were identified.

1R04 Equipment Aliqnment (71111.04Q - 5 samples;7111 1.04S - 1 sample)

.1 PartialWalkdown

a. Inspection Scope

The inspectors completed five partial system walkdown inspection samples for the plant

systems listed below. The inspectors verified that valves, switches, and breakers were

correctly aligned in accordance with Seabrook's procedures and that conditions that

could affect system operability were appropriately addressed. The inspectors reviewed

applicable piping and instrumentation drawings and system operational lineup

procedures. The documents reviewed are listed in the Attachment.

o Primary component cooling water (PCCW) "A" Train with "8" service water (SW) and

"8" PCCW out of service for work performed on April 11, 2011.

. "A" train primary component cooling water (PCCW) during planned unavailability of

the "B" train PCCW and SW systems on April 27,2011 through May 2, 2011'

o Reactor and support system alignments on April 22, 2011, in preparation for plant

startup from Mode 6.

. Residual heat removal (RHR) system alignment for low temperature over pressure

protection during shutdown cooling operations on April 1,2011 through April 4, 2011.

. "B" train RHR on May 5,2011 through May 10,2011 during the removal and

replacement of 1-RHR-P-8A motor and seal package.

Findinqs

No findings were identified.

Complete Svstem Walkdown

Insoection Scope

The inspectors completed one complete system walkdown inspection sample on the

service water system, specifically, Train "A" during a Train "B" pipe replacement and SW

ocean outage. The inspectors walked down the accessible portions of the system to

verify the system's overall material condition; that valves were correctly positioned; that

electrical power was available; that major system components were properly labeled;

that hangers and supports were correctly installed and functional; and that ancillary

equipment or debris did not interfere with system performance. The inspectors reviewed

b.

,2

Enclosure
7

plant procedures, system drawings, the UFSAR, and the technical specifications (TS).

The documents reviewed are listed in the Attachment.

b. Findinos

No findings were identified.

1R05 Fire Protection (71111.05Q - 5 samples)

.1 Quarterlv Review of Fire Areas

Inspection Scope

The inspectors completed five quarterly fire protection inspection samples. The

inspectors examined the areas of the plant listed below to assess: the control of

transient combustibles and ignition sources; the operational status and material

condition of the fire detection, fire suppression, and manual fire fighting equipment; the

material condition of the passive fire protection features; and the compensatory

measures for out-of-service or degraded fire protection equipment. The inspectors

verified that the fire areas were maintained in accordance with applicable portions of Fire

Protection Pre-Fire Strategies and Fire Hazard Analysis. The documents reviewed are

listed in the Attachment.

. Primary auxiliary building 53 FT (PAB-F-3A-Z).

. Containment 26 FT (C-F-3-2).

. Fuel storage building 7 FT (FSB-F-1-A).

o Site yard area with focus on containment outage access (PLT-F-1-0).

. Containment 0 FT and +25 FT (C-F-Z-Z and C-F-3-Z).

Findinqs

Inadequate Control of Combustible Materials

lntroduction: The inspectors identified a non-cited violation (NCV) of Technical

Specification (TS) 6.7.1.h, which requires that written procedures be established and

implemented for the fire protection program. The inspectors identified combustible

materials that were not controlled per NextEra procedure FP 2.2 in the primary auxiliary

building and in the waste process building room W8505. The inspectors identified

materials stored in W8505 in excess of FP 2.2 limits on 3 occasions from April 15,2011

to July 1,2011. Collectively, the NRC observations indicate inadequate programmatic

control of transient combustible materials. Seabrook removed the improperly stored

material identified by the inspector and entered this performance deficiency into their

corrective action program.

Description: Procedure FP 2.2, "Control of Combustible Materials", provides

requirements for controlling combustible materials at Seabrook. The inspectors

identified the following conditions that did not meet the requirements of FP 2.2:

b.

1.

Enclosure
I

(a) During a walkdown of the Waste Processing Building area W8505 on April 15,2011,

the inspectors identified ten rolls of new plastic bags. Section 4.7 of FP 2.2 allows

permanent storage of combustible materials in room W8505 in quantities specified

for normal operations on bag rack only (i.e., three rolls of bags). In addition to the

bags on the rack, seven additional rolls of bags were beside the rack. Procedure FP

2.2, Section 4.4.3, provides a permit threshold for quantities that exceed 100 pounds

of National Fire Protection Association (NFPA) flammability category 1 solid

materials (Class A materials). No transient combustible material permit was issued.

The additional seven rolls of bags exceeded 100 pounds. The inspectors discussed

this issue with Operations Management (OM). The materials were removed.

On June 15,2011, inspectors identified a similar condition in room W8505 in that

three additional rolls of bags were near the bag rack. The inspectors discussed the

issue with Shift Manager (SM) and Fire Brigade Leader (FBL). Condition Report

1661217 was initiated and the non-permitted materials were removed. On July 1,

2011, the inspectors identified the same conditions in room W8505. Specifically,

three additional rolls of bags were near the bag rack. Further, there was a half-full

55 gallon drum of used oil in the area. No transient combustible permit existed for

the materials. The inspectors discussed these observations with the on-duty FBL

and SM. The non-permitted materials were removed. Condition Reports 1666354

and 1666363 were initiated to enter this issue in the corrective action program.

(b) During a walkdown of the primary auxiliary building (PAB) room PB404 on June 15,

2011, the inspectors observed a cardboard box of paper towels and partially filled

plastic bag stored next to (within three feet) 1-SS-CP-166-8, an energized sample

analysis control panel. The materials were used by chemistry personnel to obtain

steam generator blowdown samples. The sample panel area is an elevated platform

that is approximately three feet wide. Section 4.2.5.c of FP 2.2 states that materials

are not to be stored within three feet of energized electrical equipment (panels, etc.).

Room PB4O4 houses the primary component cooling water pumps and heat

exchangers, and is a designated PRA risk significant area, defined as an area that

contributes the greatest majority of risk of core damage to fire initiated events. The

inspectors discussed the observations with the on-duty Fire Brigade Leader and Shift

Manager. The combustible materials were removed. Condition Report AR 1661010

was initiated to enter this issue in the corrective action program.

Collectively, these NRC observations indicate programmatic weakness in the control of

combustible materials. The failure by worker$ to follow procedures and the ineffective

NextEra actions to keep combustibles in WB505 below limits set by FP 2.2 raise a

concern with the control of combustibles which, if left uncorrected, could lead to a more

significance safety concern. Further, the inspectors identified that restrictions contained

in the Final Safety Analysis Report, Appendix A, Responses to BTP APCSB 9.5-1 -

materials near safety related tanks, were not reflected in FP 2.2. NextEra entered this

issue into the corrective action program as AR 1667113.

Analvsis: The inspectors determined that the failure to properly implement procedure FP

ZZtor the controlof transient combuStible materialwas a performance deficiency. This

finding was considered more than minor because, if left uncorrected, inadequate control

of combustibles could affect the Mitigating Systems cornerstone objective to assure

external factors (fires) do not impact the availability and reliability of systems which

mitigate events.

Enclosure
9

The inspectors performed a significance determination of this issue using IMC 0609,

"significance Determination Process" (SDP), Appendix F, "Fire Determination

Significance Determination Process."

The issue met the Phase I qualitative screening criteria as discussed in Appendix F.

Based on an evaluation using Step 1 of Appendix F, the inspectors determined the

finding affected the category of Fire Prevention and Administrative Controls in that

combustible material was not being properly Controlled; the finding had a "low"

degradation rating; and, the finding was of very low safety significance (Green). The

inspectors determined this event affected the cross-cutting area of H.4.(b), Human

Performance, Work Practices, because of the failure of workers to follow station

procedures.

Enforcement: TS 6.7.1.h requires that written procedures shall be implemented for the

Fire Protection Program (FPP). Fire Program procedure FP 2.2,"Controlof Combustible

Materials," Revisionl2, limits the quantity of transient combustible material stored in

W8505 (Section 4.7) and near electrical panels in P8404 (Section 4.2.5.c). Contrary to

the above, NextEra did not limit the quantity of transient combustible material stored in

W8505 and near electrical panels in P8404 in accordance with the requirements of

procedure FP 2.2. Specifically, NextEra stored ten rolls of plastic bags in WB505 and

stored a cardboard box of paper towels and a partially filled plastic bag within three feet

of an energized sample analysis control panel in PB404. Because the failure to control

combustible materials was of very low safety significance and has been entered into

NexEra's corrective action program (ARs 1661010, 1666354, 1666363), this violation is

being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy

(NCV 05000443/201 1 003-01, Inadequate Gontrol of Combustible Materials).

1R07 Heat Sink Performance (71111.07 - 1 sample)

Insoection Scope

The inspectors completed one heat sink performance inspection sample. Specifically

the inspectors reviewed the 2011 testing of the "B" component cooling water heat

exchanger to verify that the heat exchanger could fulfill its design function. The

inspectors reviewed thermal performance monitoring (WO 01202862), trending data for

heat exchanger temperatures and fouling factors, and ES1850.017, "SW Heat

Exchanger Program". The inspectors interviewed the system engineer to evaluate the

process used to monitor the heat exchanger and commitments in Generic Letter 89-13,

"Service Water System Problems Affecting Safety-Related Equipment." The inspectors

performed system walk downs and reviewed condition reports to verify that issues

associated with the heat exchanger were identified and corrected. The documents

reviewed are listed in the Attachment.

Findinos

No findings were identified.

a.

b.

Enclosure
10

1R08 Inservice Inspection (71111.08 - 1 sample)

a. Inspection Scope

The purpose of this inspection was to review and assess the effectiveness of NextEra's

In-service Inspection (lSl) program for monitoring degradation of the reactor coolant

system (RCS) boundary, risk significant piping system boundaries, and the containment

boundary. The inspectors reviewed a sample of nondestructive examination (NDE)

activities to verify compliance with American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code, Section Xl and applicable NRC Regulatory

Requirements. In addition, the inspectors reviewed samples of completed nondestructive examinations, inspection procedures and inspection test reports to verify

compliance with the ASME Code, Section Xl. The inspectors reviewed the results of the

reactor vessel nozzle dissimilar weld evaluation in the post-MSIP configuration

(AR1644106). Also, the inspectors reviewed repair and replacement activities which

involved use of welding and NDE on pressure boundary risk significant systems.

The inspectors observed the performance of NDE activities in process and reviewed

documentation and examination reports for additional nondestructive examinations.

Non-destructive test processes inspected and reviewed included Visual (W), Magnetic

Particle (MT), Penetrant (PT), Eddy Current (ECT), and Ultrasonic (UT) testing. The

sample selection was based on the inspection procedure objectives, risk significance

and sample availability. The inspectors reviewed examination procedures, procedure

and personnel qualifications and examination test results.

The inspectors reviewed the procedures used to perform visual examinations for

indications of boric acid leaks from pressure retaining components including the vessel

upper head penetrations and their connections to the control rod drive mechanisms.

The inspectors reviewed samples of operability evaluations, engineering evaluations and

corrective actions provided for active and inactive boric acid leaks and determined they

were consistent with the requirements of the ASME Code and 10 CFR 50, Appendix B,

Criterion XVl, Corrective Action.

The inspectors performed a visual examination of the containment steel shell at the zero

and minus 26 foot elevations within containment to evaluate the reported condition of the

liner coating. The inspectors reviewed a sample of test reports, photographs and

condition reports initiated as a result of the liner inspection performed by NextEra.

Corrective action specified for conditions identified were evaluated by the inspectors to

assess that the engineering organization was involved in providing evaluation and

disposition. The inspectors confirmed there was no notable damage or indication of

leakage identified during the ASME Section Xl Section IWE evaluation.

Examinations I nsoected :

. Magnetic particle test (MT) of weld F0104, field weld integrally attached preengineered pipe cap to 24 inch carbon steel SW pipe. Work order 1198488, ASME

Xl, Code Class 3. MT examination procedure ES 1807'003 R 7 Ch 1.

. Liquid penetrant (PT) test of weld F0104, examination of root pass of carbon steel

attachment weld of cap to pipe using work document 1198488, ASME Xl, Code

Class 3, PT examination procedure ES 1807.002R7 Ch 1.

Enclosure
11

Ultrasonic thickness test (UT) of carbon steel SW piping to determine if wall thinning

had occurred at various selected circumferential and axial locations of the pipe

shown on drawing 1 198488. The pipe wall thickness was measured using UT

procedure ES 1807.012 R 5.

Visual test (W-2) of reactor pressure vessel bare metal upper head surfaces with

attention to the area where control rod drives intersect the head using remote visual

techniques using test procedure ES 10-01-23.

The inspectors reviewed the steam generator (SG) degradation assessment (DA) to

determine that NextEra had reviewed and incorporated the results of the previous

outage degradation assessment, operational assessment (OA) and condition monitoring

(CM) assessment. The inspectors reviewed the eddy current test (ECT) procedure,

sample plan and data acquired. lt was noted that no steam generator tubes were

identified which specified in-situ pressure testing or specified "plugging".

The inspectors reviewed documentation of reworl</repair activities which specified the

development of ASME Section Xl repair plans with the use of welding processes to

complete the repair. The work orders (WO) which detail these repair/replacement

activities are:

. WO 1198488 02 repair of thru wall leak of SW pipe line SW 1814-1-156 and modify

support 1814-SG-02 by installation of a pre-engineered pipe cap, Drawing SKEC145189-2000. The lnspectors reviewed applicable welding and NDT procedures

to determine compliance with the ASME Xl Code requirements.

. WO 40055977 01 fabrication of new SW pipe spool and reducer for replacement of

existing spoolwhich was degraded (wall thinning). Pipe spoolwelds F0105, 0106

and 0107 were made using weld procedure ES0815.004. The inspectors reviewed

the fabrication and inspection procedures to verify compliance with the ASME Xl

Code requirements.

The inspectors reviewed the replacement material, weld procedure specifications and

qualifications, welder qualifications, weld filler metals, non-destructive tests acceptance

criteria and post work testing for each activity. The documents reviewed are listed in the

Attachment.

b. Findinos

No findings were identified.

1R11 Licensed Operator Requalification Proqram (71111.11Q - 1 sample)

.1 Quarterlv Resident Inspector Review

a. Inspection Scope

The inspectors completed one quarterly licensed operator requalification program

inspection sample. Specifically, the inspectors observed simulator just-in-time training of

licensed operators on April 28,2011 for reactor and steam plant re-start activities. The

inspectors observed formal classroom and simulator activities. The inspectors examined

the operators capability to perform actions associated with high-risk activities, the

Enclosure
12

Emergency Plan, previous lessons learned items, and the correct use and

implementation of procedures. The inspectors observed and reviewed the training

evaluator's critique of operator performance and verified that deficiencies were

adequately identified, discussed and entered into the corrective action program. The

inspectors reviewed the simulator's physical fidelity in order to verify similadties between

the Seabrook control room and the simulator. The documents reviewed are listed in the

Attachment.

b. Findinqs

No findings were identified.

1R12 Maintenance Effectiveness (71111.12Q - 2 samples)

a. Inspection Scope

The inspectors completed two maintenance effectiveness inspection samples. The

inspectors reviewed performance-based problems and completed performance and

condition history reviews for the selected in-scope structures, systems or components

(SSCs) listed below to assess the effectiveness of the maintenance program. Reviews

focused on: proper Maintenance Rule (MR) scoping in accordance with 10 CFR 50.65;

characterization of reliability issues; tracking system and component unavailability;

10 CFR 50.65 (aX1) and (a)(2) classifications; identifying and addressing common

cause failures, trending key parameters, and the appropriateness of performance criteria

for SSCs classified (aX2) as well as the adequacy of goals and corrective actions for

SSCs classified (aX1). For the periodic assessment inspection sample, the inspectors

reviewed the assessment frequency, the performance criteria, the use of operating

experience and corrective actions. The inspectors reviewed system health reports,

maintenance backlogs, and MR basis documents. The documents reviewed are listed in

the Attachment.

. Residual heat removal (RH) system classified as MR

(a)(2) with a focus on component performance impacting unavailability and reliability

(AR 1647943).

. SW system classified as MR (a)(2) with a focus on pipe wall thinning identification

and repair (ARs 161 2061, 1637922, 1639537).

b. Findinqs

No findings were identified.

1 R13 Maintenance Risk Assessments and Emerqent Work Control (71111 .13 - 6 samples)

a. Inspection Scope

The inspectors completed six maintenance risk assessment and emergent work control

inspection samples. The inspectors reviewed the scheduling and control of planned and

emergent work activities in order to evaluate the effect on plant risk. The inspectors

conducted interviews with operators, risk analysts, maintenance technicians, and

engineers to assess their knowledge of the risk associated with the work, and to ensure

Enclosure
13

that other equipment was properly protected. The inspectors reviewed the availability of

opposite train guarded and protected equipment. The compensatory measures were

evaluated against Seabrook procedures, Maintenance Manual 4.14,"TroLtbleshooting,"

Revision 0 and Work Management Manual 10.1, "On-Line Maintenance," Revision 3.

Specific risk assessments were performed using Seabrook's "Safety Monitor", as

applicable. The inspectors reviewed the maintenance items listed below. The

documents reviewed are listed in the Attachment.

. Risk mitigation actions for orange risk condition associated with reactor head

removal on April 5,2011 through April 6, 2011 (WO 1205089).

. Planned work associated with SW "A" Train during "B" train pipe replacement and

maintenance WO 00626035 for work performed April 14,2011'

. Risk mitigation actions due to unplanned entry into orange risk condition due to

degraded grid during planned work to remove steam generator nozzle dams on

April 22, 201 1 (WO 1203031 ).

o Emergent work associated with the heavy lift for the removal'and replacement of

"A" train RHR pump (1-RH-P-8A) on May 6,2011 and May 9, 2011 (WO 40083875).

o Emergent work associated with the "A" train SW piping replacement and

maintenance for leak downstream of heat exchanger isolation valve SW-V-16 on

April 20, 2011 through April 22,2011 (WO 40078357).

o Emergent work associated with the temporary repair of safety injection system

check valve SI-V82 which had a body to bonnet leak that was leak sealed with the

plant at normal operating temperature and pressure on May 18,2011 through

May 22,2011 (WO 40086371).

b. Findinqs

No findings were identified.

1R15 Operabilitv Evaluations (71111.15 * 5 samples)

a. Inspection Scope

The inspectors completed five operability evaluation inspection samples. The inspectors

reviewed operability evaluations and condition reports to verify that identified conditions

did not adversely affect safety system operability or overall plant safety. The evaluations

were reviewed using criteria specified in NRC Regulatory lssue Summary 2005-20,

"Revision to Guidance formerly contained in NRC Generic Letter 91-18, Information to

Licensees Regarding two NRC lnspection Manual Sections on Resolution of Degraded

and Nonconforming Conditions and on Operability'' and Inspection Manual Part 9900,

"Operability Determinations and Functionality Assessments for Resolution of Degraded or

Nonconforming Conditions Adverse to Quality or Safety." In addition, where a component

was determined to be inoperable, the inspectors verified that TS limiting condition for

operation implications were properly addressed. The documents reviewed are listed in

the Attachment. The inspectors also performed field walk downs and interviewed

personnel involved in identifying, evaluating or correcting the identified conditions. The

following items were reviewed:

. AR1 662418, operability of the pressurizer code safety relief valve (1-RC-V-1 17) due

to seat leakage, June 20, 2011.

Enclosure
b.

14

. AR1641413, evaluation of containment shellwith

April 20, 2011.

. AR1644074, operability of containment enclosure

craze cracking in concrete,

building with reduced modulus of

elasticity, April 21, 2011.

. AR1664399, operability of concrete structures with reduced modulus of elasticity,

June 27,2011.

. AR1664708, operability of B diesel generator with cooling water flow oscillations,

June 28,2011

Findinqs

NextEra wrote ARs1644074 and 1664399 to document the preliminary laboratory results

for concrete core samples taken for the containment enclosure building (CEB) and four

other seismic Category I buildings. Twenty core samples were taken as part of an

extent of condition investigation for AR 581434 in which NextEra determined that

sections of the below grade concrete walls could be affected by alkali-silica reaction

(ASR). Prior NRC review of this area was documented in Inspection Reports 2010-04,

201 0-05, 2011-02 and 2011-07 .

Inadequate 50.59 Screeninq for Desiqn Chanqe EC 272057 - AR1664074

NextEra issued EC272057, "Concrete Modulus of Elasticity Evaluation," on

April24,2Q11 to address the results of testing that showed a reduction in the concrete

modulus of elasticity in the CEB (AR 1644074). EC272057 also address the reduced

modulus in the Control Building/Electric Tunnel CB/ET (AR581434). The lowest

measured modulus was 2.16E+03 ksifor the CEB and 2.1E+03 ksi for the CB/ET, both

less than the design value of 3.62E+03 ksi. EC272057 was supported by calculations

C-S-1-10150 and C-S-1-10156 which reflected the degraded conditions in the design

calculations CD-20-CALC and CE-4-CALC for the control and containment enclosure

buildings, respectively.

NextEra concluded the structures remained operable and used EC272057 to disposition

the degraded condition as "use-as-is," by incorporating the degraded condition into the

design basis. In a safety evaluation screen per 10 CFR 50.59 for EC272Q57, NextEra

concluded the change to the facility did not require a complete evaluation per 50.59(cX2)

because adequate design margin existed and there was no adverse affect on an UFSAR

described design function.

The inspectors determined the 50.59 Screen for EC272057 did not correctly address

Screen Question 5.a: "Does the proposed activity involve a change to an SSC that

adversely affects an IJFSAR design function? Using the guidance of the Seabrook

1OCFR5059 Resource Manual and NEI 96-07, Revision 1, the inspectors determined

that a 50.59 evaluation is specified for changes that adversely affect design function. ln

this situation, the ASR impacted concrete with reduced modulus of elasticity which

reduces the flexural capacity of the walls would be an adverse effect. Therefore,

NextEra should have evaluated the change to the facility per 10 CFR 50.59(cX2).

The item is unresolved pending action by NextEra to complete a full 50.59 evaluation for

EC272057 and subsequent NRC review of that evaluation to determine whether the

.1

Enclosure
15

performance deficiency is more than minor. (URl 05000443/2011003-02, lnadequate

50.59 Screening for Design Ghange EC 272057).

Effects of Reduced Modulus on Concrete Structures - AR1644074 and 1664399

NextEra's analysis of the CEB samples found that the concrete has acceptable

compressive strength and reduced but acceptable modulus of elasticity. To evaluate the

effects of the reduced modulus, NextEra assessed the increase in strain for CEB

building elements and found that the strain at the most limiting element remained less

than the American Concrete Institute ACI-318 design stress limit and thus was

acceptable. NextEra evaluated the impact on flexural capacity by reviewing the change

in bending moment of structural elements. The reduced modulus causes the concrete to

have increased flexure which has the effect of shifting the balance point in how load is

transferred between the concrete and the imbedded steel (rebar). The reduced modulus

causes a shift toward the reinforced steel in tension. The resultant change in bending

moment was evaluated to show that the reduction in capacity was minimal and the

stresses on the steel and concrete remain below the design stress limits with margin.

NextEra's evaluation of the condition concluded that a change in the dynamic seismic

response of the structure would be minor, and the CEB remains capable of performing

its design function.

The prompt operability determination for the CEB (ARs 1644074 and 1664399)

evaluated how the reduced modulus would affect the structure by analysis of locally

impacted sections. The evaluation did not address the effects of reduced modulus on

the changes to the natural frequencies of the structure and the global response of the

structurelo seismic loads. The inspectors requested further information on the effects of

the reduced modulus on stresses and strain in the concrete and rebar system for which

NextEra will complete additional analyses.

The prompt operability determination (POD) for the Control Building (AR581434) as well

as for other ASR impacted structures (AR1664399) evaluated the effects of reduced

modulus on portions of the below grade structures and the components housed within

them. The evaluations lacked details to explain the effects of the reduced modulus on

structuralflexure as related to components attached to the structures, such as pipe

supports and cable trays. Similarly, the evaluations lacked details with regard to the

structure's response to seismic events as related to structure rigidity and changes in the

natural frequency, and the bases to use the ground response spectra. Further, the

evaluations lacked details to explain how the function of support anchor bolts would not

be adversely impacted by reduced concrete compressive strength in the CBIET.

This item is unresolved pending further NRC review of the above issues, action by

NextEra to complete additional analysis of the CEB conditions and subsequent NRC

Region I review of that analysis, and the completion of reviews by the NRC Office of

Nuclear Reactor Regulation specified in the associated task interface agreement

(ADAMS No. ML111610530). The result of these reviews will determine whether there

is a performance deficiency associated with this item. (URl 05000443/2011003-03,

Operability Evaluation for Degraded Goncrete in ASR Affected Plant Structures).

Enclosure
16

Untimelv Operabilitv Determination - AR 1664399

lntroduction. The inspectors identified a Green non-cited violation (NCV) of Technical

Specification (TS) 6.7.1.a that requires written procedures be established and

implemented, including administrative procedures that define authorities and

responsibilities for safe operation. NextEra identified a degraded and nonconforming

condition related to reduced modulus of elasticity for buildings housing safety related

equipment on May 27,2011, but did not complete an operability assessment until June

27, 2011, when AR1664399 and EC250348 were issued. The delayed entry of the issue

into the corrective action process to assess operability was contrary to Section 4.3 of

EN-AA-203-1001 that requires an operability determinations (OD) be completed in a

time frame commensurate with the safety significance of the issue (in most cases within

8 hours).

Description. Procedure EN-AA-203-1001, "OperabiIity Determinations/Functional

Assessments," provides requirements for evaluation of degraded conditions and

nonconforming conditions and requires: the Shift Manager (a licensed Senior Reactor

Operator) make an OD for each condition that involves equipment issues related to the

ability of an SSC to perform its TS function (Section 3.2.1); degraded or nonconforming

conditions be entered in to the corrective action program (Section a.2.1); an immediate

OD be performed following the discovery of a degraded or nonconforming condition

(Section 4.1.7); and, the immediate operability determination shall be completed in a

manner commensurate with the safety significance (in most cases during the shift when

a concern was generated or 8 hours) and consider all plant conditions (Section 4.3.1).

On April 21,2011, NextEra issued AR 1644074 upon receipt of initial test results

showing the modulus of elasticity for concrete core samples taken from the containment

enclosure building (CEB) was below the American Concrete lnstitute ACI-318 design

value 3.62 E+03 ksi. A measured modulus as low as 2.16E+03 ksi (60% of the design

value) was a degraded and nonconforming condition with respect to the properties of

concrete in Category I structures as described in Section 3.8 of the UFSAR. A reduced

modulus impacts the flexural capacity of the impacted walls and thus the function of the

building. The Shift Manager, with input from Engineering, documented the basis for an

immediate operability determination for the CEB in AR 1644074. NextEra issued

EC250348 and Calculation C-S-1-10156 on April 25, 2011 , which evaluated the integrity

of the CEB with consideration of the reduced modulus to disposition the CEB as

operable. The evaluations supporting EC250348 relied upon Calculation C-S-110150,

completed for the Control Building on September 23,2010, to show that the reduced

modulus had minimal impact on flexure and bending moment capacity of the building

walls that are heavily reinforced with steel. NextEra also initiated core sampling in

several buildings including - the equipment vault, the emergency feedwater the

emergency diesel generator buildings as part of the extent of condition review for the

issues identified in the control building.

The initial report for the results of the additional testing performed identified reduced

modulus of elasticity in all of the buildings in the expanded scope (equipment vault,

emergency feedwater and emergency diesel generator buildings). The information was

provided to the responsible engineer and made available in a draft report to NextEra on

May 27, 2011, but subject to further review and comment with the vendor for final

acceptance. On June 27, 2011, NextEra issued AR1664399 with an immediate

operability determination to address the same reduced modulus condition described in

Enclosure
.4

17

AR1644074. A Prompt Operability Determination (POD) was issued on June 28,2011,

to disposition the degraded condition for all impacted buildings which determined the

structures were fully operable with margins. The inspectors identified that, although the

data for the 3 buitdings was preliminary (final reports were not issued by the vendor until

July 1 and 27), NextEra should have initiated a condition report on May 27 ,2011 to

establish an immediate operability determination for the buildings since the reduced

concrete modulus was a degraded and nonconforming condition as described in UFSAR

Section 3.8. The failure to initiate a timely operability determination on May 27 was

contrary to Sections 4.2.1 and 4.3.1 of EN-AA-203-1001. The failure to promptly enter a

degraded condition into the corrective actions process to allow the Shift Manager to

make timely operability evaluations was a performance deficiency.

Analvsis. The inspectors determined that the failure to properly implement procedure

EN-AA-203-1001 for the degraded and nonconforming condition discussed above was a

performance deficiency. This performance deficiency was considered more than minor

based on a comparison with Examples 3.j and 3.k of Appendix E of IMC 0612.

Specifically, the performance deficiency was more than minor because a reasonable

doubt of operability for the affected concrete structures existed until further engineering

evaluations were completed to demonstrate the structures and systems that they housed

would remain functional under design and licensing basis conditions. The finding

affected the Mitigating Systems cornerstone objective to ensure the availability, reliability

and capability of systems that respond to initiating events in order to prevent core

damage. The issue was evaluated using IMC 0609, "Significance Determination

Process" (SDP), and was determined to be of very low safety significance (Green)

because the finding was not a design or qualification deficiency, did not result in an

actual loss of safety function, was not a loss of a barrier function, and was not potentially

risk significant for external events. The finding had a cross cutting aspect in the area of

problem identification and resolution, P.1(a), because NextEra did not enter identified

degraded concrete conditions for several site buildings into the corrective actions

process in a timely manner that would have ensured the shift manager completed timely

operability evaluations for the affected structures.

Enforcement. Technical Specification 6.7.1.a, Procedures and Programs, requires that

procedures be established and implemented covering administrative procedures that

define authorities and responsibilities for safe operation. Procedure EN-AA-203-1001

defines responsibilities and requirements for r:naking immediate ODs to establish the

acceptability of continued plant operation when SSCs are found to be degraded or

nonconforming. Contrary to the above, NextEra did not make an immediate operability

determinations to establish the acceptability of continued plant operation when the

reduced concrete modulus for several plant structures was identified on May 27,2011.

Because this failure to make timely operability determinations is of very low safety

significance and was entered into NextEra's Corrective Action Program (CR1673102),

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

Enforcement Policy. (NCV 05000443/20 1 1 003-04, U ntimely Operabil ity Determi nation

for Degraded Goncrete Structures Housing Safety-Related Equipment)

Inadequate Operabilitv Determination - AR 1664708

lntroduction. The inspectors identified a Green non-cited violation (NCV) of Technical

Specification (TS) 6.7.1.a that requires that written procedures be established and

implemented, including administrative procedures that define authorities and

Enclosure
18

responsibilities for safe operation. NextEra identified a degraded condition related to

reduced service water (SW) flow to the B emergency diesel generator (EDG) heat

exchanger (HX) on June 28, 2Q11, but did not fully evaluate the reduced flow under all

plant conditions as required by NextEra procedure EN-AA-203-1001.

Description. Procedure EN-AA-203- 1 00'1, "Operabi l ity Determ inations/Functional

Assessments," provides requirements for evaluation of degraded conditions and

nonconforming conditions and requires: the Shift Manager (a licensed Senior Reactor

Operator) make an OD for each condition that involves equipment issues related to the

abif ity of an SSC to perform its TS function (Section 3.2.1); degraded or nonconforming

conditions be entered in to the corrective action program (Section 4.2.1); an immediate

OD be performed following the discovery of a degraded or nonconforming condition

(Section a.1.7); and, the immediate operability determination shall be completed in a

manner commensurate with the safety significance (in most cases during the shift when

a concern was generated or 8 hours) and consider all plant conditions (Section 4.3.1).

On June 28,2011, NextEra issued AR1664708 when operators observed reduced SW

flow through the B EDG heat exchanger during weekly testing of valve SW-V18. The

flow initially varied from 500 to 1000 gpm, but increased to 1400 gpm with additional

valve strokes. The operators questioned the adequacy of flow indication from SW-FE6191 , but the operability evaluation documented in AR1664708 accepted the flow with

"no operability issues noted" and with plans to monitor the condition. NextEra observed

"normal" flow during a subsequent valve stroke.

The operators again observed reduced and erratic SW flows (800-1500 gpm) during the

next valve test on July 9, 2011. The B EDG was declared inoperable, AR1667857 was

written and a POD per Section 4.3.1 .C of EN-AA-203-1001 was requested at that time.

A more detailed investigation and evaluation determined that flow element SW-FE-6191

was partially plugged contributing to variability in flow indication (but not the reduction),

and that marine groMh inside the SW pipe (line 1806-1-1 53-16) upstream of diesel heat

exchanger DG-E428 was causing intermittent fouling of the heat exchanger tubes and

reduced flow. Corrective actions were initiated to address potential heat exchanger

fouling in both EDG SW cooling loops. The engineering evaluation associated with the

POD confirmed that the "8" EDG and cooling subsystem was and had been fully

operable under design basis conditions, including cooling water temperatures at the

environmental extremes for operation on the ocean or the cooling tower.

The inspectors noted that diesel operating procedure OS 1026.09 requires a minimum of

900 gpm when SW cooling is provided on the ocean, and 1800 gpm when SW cooling is

provided on the cooling tower. The inspectors noted further that normal SW flow

through the heat exchanger varies from 1500 to 1900 gpm depending on ocean level at

the intake. The inspectors determined that the June 28 operability evaluation

documented in AR 1664708lacked sufficient basis to explain the reduced flow at 500-

1400 gpm, and failed to fully evaluate the EDG cooling function. The June 28 evaluation

did not fully evaluate flow under all plant conditions as specified by Section 4.3.1.A of

EN-AA-203-1001, namely, whether flow was adequate relative to the 1800 gpm needed

for operation on the cooling tower. NextEra should have fully investigated the flow

anomaly on June 28 and evaluated EDG operability under all design basis conditions

per Section 4.3.1.C of EN-AA-203-1001 , since reduced SW flow in the range of 500 to

1000 gpm was a degraded condition that impacted engine cooling. When EDG cooling

was further investigated on July 9, the presumption on June 28 that the flow anomaly

Enclosure
19

was caused by erratic indication was proven wrong. The initial incomplete operability

evaluation resulted in the delayed identification, assessment and mitigation of a

degraded condition impacting the functionality of the EDG. The failure to promptly and

thoroughly evaluate degraded conditions for operability was a performance deficiency.

Analvsis: The inspectors determined that not properly implementing procedure EN-AA203-1001 for the degraded condition discussed above was a performance deficiency.

This performance deficiency was considered more than minor based on a comparison

with Examples 3.j and 3.k of Appendix E of IMC 0612. Specifically, the performance

deficiency was more than minor because a reasonable doubt of operability existed until

further engineering evaluations were completed to demonstrate adequate service water

flow to the B EDG HX existed and that the B EDG remained functional under design and

licensing basis conditions. The finding affected the Mitigating Systems cornerstone

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

initiating events in order to prevent core damage. The issue was evaluated using IMC

0609, "significance Determination Process" (SDP), and was determined to be of very

low safety significance (Green) because the finding was not a design or qualification

deficiency, did not result in an actual loss of safety function, was not a loss of a barrier

function, and was not potentially risk significant for external events. The finding had a

cross cutting aspect in the area of problem identification and resolution, P.1(c), because

NextEra personnel did not adequately evaluate operability to ensure that EDG cooling

flow was acceptable under all operating conditions and assure appropriate corrective

actions were timely comPleted.

Enforcement. Technical Specification 6.7.1.a, Procedures and Programs, requires that

procedures be established and implemented covering administrative procedures that

define authorities and responsibilities for safe operation. Procedure EN-AA-203-1001

defines responsibilities and requirements for making immediate ODs to establish the

acceptability of continued plant operation when SSCs are found to be degraded.

Contrary to the above, NextEra did not fully evaluate degraded service water flow on

June 28, 2011, resulting in the delayed identification, assessment and correction of a

condition that impacted B EDG cooling. Because the finding is of very low safety

significance and was entered into NextEra's corrective action program (CR1673102),

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

Enforcement Pol icy. (NCV 05000443/201 1 003-05, I nadequate Operabil ity

Determination for Reduced EDG HX Gooling Water Flow)

1R18 Plant Modifications (71111.18 - 3 samples)

.1 Permanent Modification - EC 145280: Proiect 52 SY Upqrade

a. Inspection Scope

The inspectors completed one permanent modification inspection sample. The

inspectors reviewed modification package EC145280 that completed changes in the 345

kV switchyard to enhance reliability. The modifications included the installation of new

breakers and bus sections to connect the Seabrook generator and unit auxiliary

transformer to the grid. The review was completed to verify that the design bases and

performance capability of the system was not degraded. The inspectors verified the new

configuration was accurately reflected in the design documentation, and that the postmodification testing was adequate to ensure the SSCs would function properly. The

Enclosure
b.

.2

a.

20

inspectors interviewed plant staff, and reviewed issues entered into the corrective action

program to verify that NextEra was effective at identifying and resolving problems

associated with temporary modifications. The documents reviewed are listed in the

Attachment.

Findinqs

No findings were identified.

Temporarv Modification - EC 272290: lnstall Varistor in Panel for CBA-CP-177

Inspection Scope

The inspectors completed one temporary modification inspection sample. The

inspectors reviewed modification package EC 272290 associated with operation of the A

EDG. The modification installed a varistor in the unit sub panel for CBA-CP-177. The

varistor was installed across relay coils 52X and 52Y to minimize electrical transients

when the anti-pump and breaker closing relays operated. The purpose of the varister is

to suppress induced voltages in the emergency power sequencer logic circuits that can

negatively affect proper operation of the A emergency diesel generator (CR1645405).

The review was completed to confirm that the design bases and performance capability

of the system were not degraded. The inspectors verified the new configuration was

accurately reflected in the design documentation (reference Drawing 310926 Sheet

AC4b), and that the poslmodification testing was adequate to ensure that affected

SSCs would function properly. The inspectors also interviewed plant staff, and reviewed

issues entered into the corrective action program to verify that NextEra was effective at

identifying and resolving problems associated with temporary modifications. The

documents reviewed are listed in the Attachment.

Findinos

No findings were identified.

Temporarv Modification - EC 272512: Leak Sealinq of Sl-V82

Inspection Scope

The inspectors completed one temporary modification inspection sample. The

inspectors reviewed modification package EC272512 that installed a mechanical clamp

and seal on safety injection system check valve Sl-V82. The leak seal was installed to

minimize gasket leak at the body to bonnet flange identified during plant startup to begin

operating cycle 15. The review was completed to confirm that the design bases and

performance capability of the system were not degraded. The inspectors verified the

new configuration was accurately reflected in the plant documentation and that the

clamp installation and sealing process would not adversely affect the check valve design

functions. The inspectors also interviewed plant staff, and reviewed issues entered into

the corrective action program to verify that NextEra was effective at identifying and

resolving problems associated with temporary modifications. The documents reviewed

are listed in the Attachment.

b.

.3

Enclosure
21

b. Findinqs

No findings were identified.

1R19 Post-Maintenance Testinq (71111.19 - 7 samples)

a. Inspection Scooe

The inspectors completed seven post-maintenance testing (PMT) inspection samples.

The inspectors observed portions of PMT activities in the field to verify the tests were

performed in accordance with the approved procedures. The inspectors assessed the

test adequacy by comparing the test methodology to the scope of the maintenance work

performed. The inspectors evaluated the test acceptance criteria to verify that the test

procedure ensured that the affected systems and components satisfied applicable

design, licensing bases and TS requirements. The inspectors also reviewed recorded

test data to confirm all acceptance criteria were satisfied during testing. The documents

reviewed are listed in the Attachment. The activities reviewed are listed below:

. Retest of main steam to emergency feedwater pump turbine steam supply valve 1-

MS-V-393 on May 18,2A11, following overhaul per WO 40065448.

. Retest of chemical and volume control system charging flow control valve 1-CSFCV-121 on May 18, 2011, following overhaul per WO 1 199620'

o Retest of "8" train charging pump 1-CS-P-28 on April 26,2011, following motor

replacement per WO 627385.

o Retest of reactor coolant loop 1 residual heat removal pump suction isolation valve

1-RC-V-22 on April 11,2Q11, following overhaul per WO 1196480'

o Retest of "A" train RHR pump 1-RH-P-8A following replacement on May 16,2011,

per WO 40085334.

o Retest of "A" emergency diesel generator on May 9, 2011, following failure of the

emergency power sequencer during testing per WO 40082703 (CR1645405).

. Retest of Sl check valve Sl-V-82 in accordance with OX1 401.04 on May 22, 2011,

following leak seal repair per WO40086371.

b. Findinqs

No findings were identified.

1R20 Refuelinq and Outaqe Activities (71111.20 - 1 sample)

.1 Refuelinq Outaqe OR14

a. Inspection Scope

The inspectors completed one refueling and outage activities inspection sample. The

inspectors reviewed the operational, maintenance, and testing activities for the

fourteenth refueling outage (OR14) starting on April 1, 2011. The documents reviewed

are listed in the Attachment.

Enclosure
22

Review of Outaqe Plan

The inspectors reviewed the outage plans to evaluate NextEra's ability to assess and

manage the outage risk. The inspectors reviewed the outage risk assessment provided

in Engineering Evaluation EE-11-02, "OR14 Outage Schedule Initial Shutdown Risk

Review."

Monitorinq of Shutdown Activities

The inspectors reviewed activities to shut the plant down in accordance with plant

procedures. The inspectors observed completion of various activities specified to place

the plant in a cold shutdown condition to assess operator performance, communications,

command and control and procedure adherence. The inspectors reviewed operator

adherence to TS specified cooldown limits. The inspectors performed inspection tours

of plant areas not normally accessible during plant power operations to verify the

integrity of structures, piping and supports, and to confirm that systems appeared

functional.

Refuelinq Activities and Reactivitv Control

The inspectors verified that refueling activities were performed in accordance with

procedures OS1000.09 and RS0721 . The inspectors independently verified on a

sampling basis that requirements for core alteration were met. The inspectors observed

NextEra actions during core alterations to assure core reactivity was controlled. The

inspectors observed activities from the control room, the reactor cavity and the spent fuel

pool at various times. The inspectors verified that fuel movement was tracked in

accordance with the fuel movement schedule. The inspectors verified NextEra action to

meet the requirements of TS 3.9 for refueling operations, including the requirements for

boron concentration and core monitoring using the source range monitors. The

inspectors observed communications and coordination of activities between the control

room and the refueling stations while fuel handling activities were in progress. The

inspectors verified reactivity was controlled in accordance with the requirements of

Technical Specification 3.9.

Control of Outaoe Risk and Activities

The inspectors reviewed daily shutdown risk assessments during OR14 to verify that

NextEra addressed the outage impact on defense-in-depth for the critical safety

functions: electrical power availability, inventory control, decay heat removal, reactivity

control, and containment. The inspectors reviewed how NextEra provided defense-indepth for each safety function and implemented the planned contingencies in order to

minimize overall risk where redundancy was limited or not available. The inspectors

periodically reviewed risk updates accounting for schedule changes and unplanned

activities. The inspectors reviewed management controls to manage fatigue by reviewing

waiver requests and assessments.

Controlof Heaw Loads

The inspectors reviewed NextEra's activities to control the lift of heavy loads in

accordance with plant procedures and the commitments to NUREG 0612. The

inspectors observed the lift preparations and lift activities to verify adherence to

established procedures and controls. The inspectors used an operating experience

smart sample as a reference for this review.

Enclosure
23

Clearance Activities and Confiquration Control

The inspectors reviewed a sample of risk significant clearance activities and verified tags

were properly hung and/or removed, equipment was appropriately configured per the

clearance requirement, and that the clearance did not impact equipment credited to

meet the shutdown critical safety functions.

Inventorv Control

The inspectors reviewed NextEra actions to establish, monitor and maintain the proper

water inventory in the reactor during the outage, and in the reactor and spent fuel pool

after flooding the reactor cavity for refueling activities. The inspectors reviewed the plant

system flow paths and configurations established for reactor makeup and reactivity

control, and verified the configurations were consistent with the outage plan.

Reduced Inventorv and Mid-Loop Conditions

The inspectors reviewed NextEra's procedures to implement commitments from Generic

Letter 88-17 and confirmed that controls for mid-loop operations were in place. The

inspectors verified reactor coolant system instrumentation was installed and configured

to provide accurate indication. The inspectors reviewed outage activities that were

performed during periods when there was a short time-to-boil to assure adequate

controls were in place. Periodically, during the decreased inventory conditions, the

inspectors verified that the configurations of the plant systems were in accordance with

the commitments. During reduced inventory operations, the inspectors observed

NextEra's control of distractions to assure the operator could maintain the specified

reactor vessel level.

Foreiqn Material Exclusion

The inspectors reviewed the implementation of Seabrook procedures for foreign material

exclusion control (FME) for the open reactor vessel, reactor cavity and spent fuel pool.

The inspectors reviewed NextEra actions to verify that FME issues were documented

and resolved.

Electrical Power

Thl inspectors verified that the status of electrical systems met TS requirements and the

outage risk control plan. The inspectors verified that compensatory measures were

implemented when electrical power supplies were impacted by outage work activities

and that credited backup power supplies were available.

RHR Svstem Monitorinq

The inspectors observed spent fuel pool (SFP) and reactor decay heat removal system

status and operating parameters to verify that the cooling systems operated properly.

The review included periodic review of SFP and reactor cavity level, temperature, and

RHR flow. The inspectors reviewed system status to verify the proper system alignment

was established for vessel and cavity level measurement.

Containment Control

The inspectors reviewed NextEra activities during the outage to control primary

containment closure and integrity, and to prepare the containment for closure prior to

plant restart. The inspectors performed tours of all levels in the containment throughout

the outage and prior to plant startup per procedure OS101 5.18 to review NextEra's

cleanup and demobilization controls in areas where work was completed to assure that

tools. materials and debris were removed. This review focused on the control of

Enclosure
24

transient combustibles and the removal of debris that could impact the performance of

safety systems.

Monitorinq Plant Heat up. Approach to Critical and Startup

The inspectors observed operator performance during the plant startup activities

performed between April 28, 2011 and May 26, 2011. The inspection consisted of

control room observations, plant tours and a review of the operator logs, plant computer

information, and station procedures. The inspectors observed pre-job briefs for key

evolutions. The inspectors reviewed the preparations for changes in operating modes.

The reactor was taken critical on May 23,2011 at 03:08 a.m., and completed power

ascension to 100% FP on May 26, 2011. The inspectors verified, on a sampling basis,

that TS, license conditions, and other requirements for mode changes were met. The

inspectors verified RCS integrity throughout the restart process by periodically reviewing

RCS leakage calculations and by review of systems that monitor conditions inside the

containment.

Problem ldentification and Resolution

The inspectors reviewed NextEra actions to identify outage related issues and enter

them into the corrective action program. This inspection included a review of the

corrective actions for Condition Report 1640003. The inspectors reviewed a sample of

the corrective actions to verify they were appropriate to resolve the identified issues.

b. Findinos

No findings were identified.

1R22 Surveillance Testinq (71111.22 - 7 samples)

Inspection Scope

The inspectors completed seven surveillance testing inspection samples. The

inspectors observed portions of surveillance testing activities for safety-related systems

to verify that the system and components were capable of performing their intended

safety function, to verify operational readiness, and to ensure compliance with specified

TS and surveillance procedures. The inspectors attended selected pre-evolution

briefings, performed system and control room walk downs, observed operators and

technicians perform test evolutions, reviewed system parameters, and interviewed the

system engineers and field operators. The test data recorded was compared to

procedural and TS requirements, and to prior tests to identify any adverse trends. The

documents reviewed are listed in the Attachment. The following surveillance activities

were reviewed:

. EX 1804.033, Containment Spray System 10 Year Air Flow Test, April 8,2011

(wo1 209232 11209233).

. OX1426.34, Diesel Generator 1A 18 Month Operability Surveillance, April 26,2011,

May 2, 2011 and May 1 1, 2011 (WO 40077892).

. OX1413.08, Residual Heat Removal Pump 8A Comprehensive Test (lST),

April 18,2011 (WO 01203773).

. RS1748, Subcritical Physics Testing Using SRWM, May 17 ,2011'

. OX1 426.32, Diesel Generator 1B 18 Month Operability Surveillance, April 24,2011

Enclosure
25

through April 25, 2011 (WO 40076902).

. EX1803.003, Local Leakage Rate Testing of FP-V-588 and FP-V-592 (LLRT),

April 1, 2011 (WO01209198).

. EX1803.003, Local Leakage Rate Testing of Penetration X358, Pressurizer Sample

Line (LLRT), April 5, 2011 (WO01209191 ).

The inspectors reviewed deficiencies related to surveillance testing and verified that the

issues were entered into the corrective action program. The documents reviewed are

listed in the Attachment.

b. Findinqs

No findings were identified.

2. RADIATION SAFEW

Cornerstone: Occupational Radiation Safety

2RS01 Radioloqical Hazard Assessment and Exoosure Controls (71124.01)

a. Inspection Scope

During the period May 9, 2011 through May 12,2011, the inspector performed the

following activities to verify that NextEra was evaluating, monitoring, and controlling

radiological hazards for work performed during the OR-14 refueling outage in locked

high radiation areas (LHRA) and other radiological controlled areas. lmplementation of

these controls was reviewed against the criteria contained in 10 CFR Part20, Technical

Specifications, and NextEra's procedures. The documents reviewed are listed in the

Attachment.

Radioloqical Hazards Control and Work Coveraqe

The inspector identified work performed in radiological controlled areas and evaluated

NextEra's assessment of the radiological hazards. The inspector evaluated the survey

maps, exposure control evaluations, electronic dosimeter dose/dose rate alarm set

points, and radiation work permits (RWP), associated with these areas, to determine if

the exposure controls were acceptable. Specific work activities evaluated included

transferring the 8A residual heat removal (RHR) pump into the decay heat vault (RWP

65) and hydrolazing the spent fuel pool (SFP) leak-off lines (RWP 61). For these tasks,

the inspector attended the pre-job briefings, reviewed relevant documents, and

discussed the job assignments with the workers. Radiation protection technicians were

questioned regarding their knowledge of plant radiological conditions for these jobs, and

the associated controls.

The inspector reviewed the air sample records for samples taken prior to installing steam

generator (SG) nozzle dams, to determine if the samples collected were representative

of the breathing air zone and analyzed/recorded in accordance with established

procedures. During plant tours, the inspector verified that continuous air monitors were

strategically located to assure that potential airborne contamination could be identified in

a timely manner and that the monitors were located in low background areas.

Enclosure
26

The inspector toured accessible radiological controlled areas located in the primary

auxiliary building, fuel handling building, decay heat vaults, and waste processing

building. With the assistance of a radiation protection technician, independent radiation

surveys were performed of selected areas to confirm the accuracy of survey data, and

the adequacy of postings.

Additionally the inspector reviewed the RWPs developed for other work performed

during OR-14 including installation of temporary shielding and scaffolding. ln particular,

the inspector reviewed the electronic dosimeter dose/dose rate alarm set points, stated

on the RWP, to determine if the set points were consistent with the survey indications

and plant policy.

lnstructions to Workers

By attending pre-job briefings, the inspector determined that workers, performing

radiological significant tasks, were properly informed of electronic dosimeter alarm set

points, low dose waiting areas, stay times, and work site radiological conditions. By

observing work-in-progress, the inspector determined that stay times were appropriately

monitored by supervision to assure no procedural limit was exceeded. Jobs observed

included transferring the 8A RHR pump into the decay heat vault and hydrolazing SFP

leak off lines.

During plant tours, the inspector determined that locked high radiation areas (LHRA) and

a very high radiation area (VHRA) had the appropriate warning signs and were properly

secured.

The inspector inventoried the keys to LHRAs to determine if the keys were appropriately

controlled, as specified by procedure. The inspector discussed with radiation protection

supervision the procedural controls for accessing LHRAs and VHRAs and determined

that no changes have been made to reduce the effectiveness and level of worker

protection.

Contamination and Radioactive Material Control

During plant tours the inspector confirmed that contaminated materials were properly

bagged, surveyed/labeled, and segregated from work areas. The inspector observed

workers using contamination monitors to determine if various tools/equipment were

potentially contaminated and met criteria for releasing the materials from the RCA.

Radioloqical Hazards Control and Work Coveraqe

By observing preparations for installing the 8A RHR pump and for hydrolazing the SFP

leakoff lines, the inspector determined that workers wore the appropriate protective

equipment, had dosimetry properly located on their bodies, and were under the positive

control of radiation protection personnel. Supervisory personnel specified the roles and

responsibilities of each worker and reviewed the potentialjob hazards to assure that

exposure was minimized and that industrial safety measures were implemented.

Radiation Worker Performance

During job performance observations, the inspector determined that workers complied

with RWP requirements and were aware of radiological conditions at the work site.

Additionally, the inspector determined that radiation protection technicians were aware of

RWP controls/limits applied to various tasks and provided positive control of workers to

reduce the potential of unplanned exposure and personnel contaminations.

Enclosure
27

Problem ldentification and Resolution

A review of Nuclear Oversight field observations (OR-14 Daily Quality Summaries)

reports, dose/dose rate alarm reports, personnel contamination event reports and

associated condition reports, was performed to determine if identified problems and

negative performance trends were entered into the corrective action program and

evaluated for resolution and to determine if an observable pattern traceable to a similar

cause was evident.

Relevant condition reports (CR), associated with radiation protection control access and

radiological hazard assessment, initiated between January and May 2Q11, were

reviewed and discussed with NextEra staff to determine if the follow up activities were

being performed in an effective and timely manner, commensurate with their safety

significance.

b. Findinqs

No findings were identified.

2RS02 Occupational ALARA Plannino and Controls (71124.02)

lnspection Scope

During the period May 9, 2011, through May 12,2011, the inspector performed the

following activities to verify that NextEra was properly implementing operational,

engineering, and administrative controls to maintain personnel exposure as low as is

reasonably achievable (ALARA) for tasks performed during the OR-14. lmplementation

of this program was reviewed against the criteria contained in the 10 CFR ParL 20,

applicable industry standards, and NextEra's procedures. The documents reviewed are

listed in the Attachment.

Radioloqical Work Planninq

The inspector reviewed pertinent information regarding site cumulative exposure history,

current exposure trends, and the ongoing exposure challenges for the outage. The

inspector reviewed various OR-14 ALARA plans.

The inspector reviewed the exposure status for tasks performed during the outage and

compared actual exposure with forecasted estimates contained in various project

ALARA plans (AP). In particular, the inspector evaluated the effectiveness of ALARA

controls for alljobs that were estimated to exceed 5 person-rem. These jobs included

reactor vessel disassembly/reassembly (AP 11-01), steam generator (S/G) eddy current

testing (ECT) (AP 1 1-02), and reactor vessel nozzle walk downs (AP 11-13).

The inspector reviewed the ALARA plans and associated Work-ln-Progress (W-l-P)

ALARA reviews for those jobs whose actual dose approached the forecasted estimate.

Included in this review were the W-l-P's for cavity decontamination, reactor coolant

pump seal replacement/motor maintenance, and scaffolding installation.

The inspector evaluated the departmental interfaces between radiation protection,

operations, maintenance crafts, and engineering to identify missing ALARA program

elements and interface problems. The evaluation was accomplished by interviewing site

Enclosure
28

staff, reviewing outage W-l-P reviews, and reviewing recent station radiation safety

committee (RSC) meeting minutes. Included in this review were the actions taken by the

RSC to lower outage pro.yect dose goals, as a result of lowering the plant's source term

by an effective primary system cleanup.

Verification of Dose Estimates

The inspector reviewed the assumptions and basis for the OR-14 ALARA forecasted

exposure. The inspector also reviewed the revisions made to various outage project

dose estimates due to a reduced source term (i.e., lower dose rates); including reactor

disassembly/reassembly activities, reactor coolant pump (RCP) maintenance, and steam

generator maintenance.

The inspector evaluated the implementation of the NextEra's procedures associated with

monitoring and re-evaluating dose estimates and allocations when the forecasted

cumulative exposure for tasks exceeded the actual exposure. Included in the review

were W-l-P reports, that evaluated the effectiveness of ALARA measures, including

source term controls, and actions by the RSC to subsequently lower dose goals from the

original estimates.

Additionally, the inspector reviewed the exposures for the ten (10) workers receiving the

highest doses for 2Q11 to confirm that no individual exceeded the regulatory limits or

performance indicator thresholds.

Source Term Reduction and Control

The inspector reviewed the status and historical trends for the source term. Through

review of survey maps and interviews with the Radiation Protection Manager, the

inspector evaluated recent source term measurements and control strategies. Specific

strategies being employed included use of macro-porous clean up resin, use of

submersible ion exchange filters in the reactor cavity, and installation of

permanenUtemporary shielding.

The inspector reviewed the effectiveness of temporary shielding by reviewing pre/post

installation radiation surveys for selected components having elevated dose rates.

Shielding packages reviewed included those placed on the RHR piping, pressurizer

spray piping, steam generator penetrations, and RCP piping.

Job Site Inspections

During plant tours, the inspector assessed the implementation of ALARA controls

specified in APs and RWPs, performed during OR-'14. These activities include work on

the 8A RHR pump (AP 11-019) and hydrolazing SFP leak off lines. Workers were

questioned regarding their knowledge of job site radiological conditions and ALARA

measures applied to their tasks.

Problem ldentification and Resolution

The inspector reviewed elements of NextEra's corrective action program related to

implementing the ALARA program to determine if problems were being entered into the

program for timely resolution, the comprehensiveness of the cause evaluation, and the

effectiveness of the corrective actions. Specifically, recent condition reports related to

programmatic dose challenges, personnel contaminations, dose/dose rate alarms, and

the effectiveness in predicting and controlling worker exposure were reviewed.

Enclosure
29

b. Findinqs

No findings were identified.

2RS03 ln-Plant Airborne Radioactivitv Control and Mitioation (7 1 124.03)

b.

Inspection Scope

During the period May 9, 2011 through May 12,2011, the inspector performed the

following activities to verify that in-plant airborne concentrations of radioactive materials

are being controlled and monitored, and to verify that respiratory protection devices are

properly selected and used by qualified personnel. lmplementation of these programs

was evaluated against the criteria contained in 10 CFR Parl20, applicable industry

standards, and NextEra's procedures. The documents reviewed are listed in the

Attachment.

Enqineerinq Controls

The inspector evaluated the use of portable HEPA ventilation systems installed in

various plant areas during the OR-14 outage. The inspector determined that the

ventilation systems were located at work locations; e.9., steam generators, and the 8A

RHR pump cubicle where airborne contamination could potentially occur. The inspector

reviewed testing records for portable HEPA ventilation systems to determine that

procedural performance criteria were met.

Respiratorv Protection

The inspector reviewed the use of respiratory protection devices worn by workers. The

inspector reviewed initial radiation survey and air sampling records for S/G nozzle dam

installations in the A through D hot and cold legs, associated RWPs, and APs to

determine if the use of respiratory protection devices was commensurate with the

potential externaldose that may be received when wearing these devices. Additionally,

the inspector evaluated the use of respiratory protection; i.e., Delta Suits, for other

outage tasks, including cavity decontamination.

Problem ldentification and Resolution

The inspector reviewed elements of NextEra's corrective action program related to

implementing the airborne monitoring program to determine if problems were being

entered into the program for timely resolution, the comprehensiveness of the cause

evaluation, and the effectiveness of the corrective actions. Specifically, condition reports

related to monitoring challenges, personnel contaminations, dose assessments, and the

reliability of monitoring equipment were reviewed.

Findinqs

No findings were identified.

Enclosure
30

2RS04 Occupational Dose Assessment (71124.04)

Inspection Scope

During the period May 9, 2011 through May 12,2011, the inspector performed the

following activities to verify the accuracy and operability of personal monitoring

equipment and the effectiveness in determining a worker's total effective dose

equivalent. lmplementation of these programs was evaluated against the criteria

contained in 10 CFR Part.2O, applicable industry standards, and NextEra's procedures.

The documents reviewed are listed in the Attachment.

External Dosimetrv

The inspector verified that NextEra's dosimetry processor was accredited by the

National Voluntary Laboratory Accreditation Program (NVLAP). The inspector verified

that the approved dosimeter irradiation categories were consistent with the types and

energies of the site's source term. The inspector reviewed NextEra's semi-annual

quality control evaluation; i.e., TLD blind spiking, of the dosimetry processor.

The inspector confirmed that NextEra has developed "correction factors" to address the

response differences of electronic dosimeters as compared to thermoluminescent

dosimeters.

Internal Dosimetrv

The inspector evaluated the equipment and methods used to assess worker dose

resulting from the uptake of radioactive materials. Included in this review were bioassay

procedures, whole body counting equipment (FastScan, Chair counter, portal

contamination monitors) calibration checks and operating procedures, and the analytical

results for 10 CFR Part 61 samples.

The inspector determined that the procedural methods include techniques to distinguish

internally deposited radioisotopes from external contamination, methods to assess dose

from hard-to-measure radioisotopes, and methods to distinguish ingestion pathways

from inhalation pathways.

The inspector reviewed the results from two whole body counts to assess the adequacy

of the counting time, background radiation contribution, and the nuclide library used for

assessing deposition. No individual exposure exceeded a committed effective dose

equivalent (CEDE) of 10 mrem.

Special Dosimetric Situations

Declared Preqnant Workers

The inspector reviewed the procedural controls, and associated records, for managing

declared pregnant workers (DPW) and determined that no DPWs were employed during

the outage. The inspector reviewed the procedural controls to assure compliance with

10 CFR Part20.

Multi-Dosimetrv Methods

The inspector reviewed NextEra's procedures for monitoring external dose where

significant dose gradients exist at the work site. For OR-14, external effective dose

Enclosure
b.

31

equivalent (EDEX) methods were used to evaluate personnel exposure for

installing/removing steam generator nozzle dams. The inspector reviewed the

dosimetric results for these jobs. The inspector confirmed that in addition to the TLDs

worn, workers also wore electronic dosimeters, equipped with telemetry, to assure that

dose fields were promptly monitored by radiation protection technicians.

Problem ldentification and Resolution

The inspector reviewed elements of NextEra's corrective action program related to

implementing the dosimetry program to determine if problems were being entered into

the program for timely resolution, the comprehensiveness of the cause evaluation, and

the effectiveness of the corrective actions. Specifically, condition reports related to dose

assessments, personnel contaminations, and dose/dose rate alarms were reviewed.

Findinos

No findings were identified.

OTHER ACTIVITIES

ldentification and Resolution of Problems (71152 - 2 sample)

Review of ltems Entered into the Corrective Action Prooram

Inspection Scope

As specified by Inspection ProcedureTll52, "ldentification and Resolution of Problems,"

and in order to help identify repetitive equipment failures or specific human performance

issues for follow-up, the inspectors performed a daily screening of items entered into the

Seabrook corrective action program (CAP). This review was accomplished by accessing

NextEra's computerized database. The documents reviewed are listed in the

Attachment.

Findinqs

No findings were identified.

Semi-Annual Review to ldentifv Trends

lnspection Scope

As specified by Inspection Procedure 71152, "Problem ldentification and Resolution," the

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

indicate the existence of more significant safety issues. The inspection included a

review of repetitive or closely-related issues documented by NextEra outside of the

corrective action program, such as assessment reports, trend reports, performance

indicators, major equipment problem lists, system health reports, and maintenance or

corrective action program backlogs. The inspectors reviewed the Seabrook corrective

action program database for the first and second quarters of 2011, to assess CRs

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

as well as individual issues identified during the NRCs daily CR review (Section

4OA2.1). The inspectors reviewed the 2011 First Quarter trend reports by the

4.

4c.A2

.1

a.

b.

.2

a.

Enclosure
b.

32

operations, security and nuclear projects departments, together with the Fourth Quarter

2010 trend report to verify that NextEra was appropriately evaluating and trending

adverse conditions in accordance with procedure P|-AA-207, "Trend Coding and

Analysis."

Assessment and Observations

No findings were identified. The inspectors did not identify any trends that NextEra had

not identified. The inspectors reviewed a sample of issues and events that occurred over

the past two quarters that were documented in the corrective action program. The

inspectors verified that NextEra appropriately considered identified issues as emerging

trends, and in some cases, verified the adequacy of the actions completed or planned to

address the identified trends.

NextEra noted the need for continued focus on human performance. NextEra completed

a common cause evaluation for an adverse trend in human performance in Operations

(CR594198) with improvements noted in the first quarter of 2Q11. During periodic

meetings with station management, the inspectors discussed NRC observations related

to human performance. One example included the inadvertent loss of 345KV Line 394

(CR 1640003) that was caused by a combination of inadequate work package

instructions and inadequate worker knowledge of tagout conditions. Another example

included the inadequate performance of a reactor coolant system (RCS) leakage

surveillance per Technical Specification 4.6.2.1.e (CR1663219), in which valve RCV147, whose position is indicated on the main control board, remained closed for thirty

(30) days. While the procedures used for the RCS leakage surveillance could be

enhanced, the cause of the issue was the failure to use fundamental operator skills

during the performance of routine duties. NextEra corrective actions include a renewed

emphasis on operator fundamentals in the operator training program. NextEra continues

to address human performance site wide through procedure enhancements,

management observations and a focus on procedure compliance in continuing training

sessions.

NextEra continued to focus on equipment performance and reliability. Performance

problems with secondary plant equipment continue to challenge operators and have

resulted in the need to reduce plant power or take the turbine offline three times in three

quarters (CRs 591828,1616988,1657622), as reflected in an adverse trend in the NRC

Performance indicator for Unplanned Power changes. During periodic meetings with

station management, the inspectors discussed emergent equipment issues that

impacted safety system availability [e.9., service water system corrosion (CR1633034),

EDG sequencer failure (CR1645405), A RHR pump seal leakage (CR1647943)lor

impacted the primary system boundary [e.9., Sl check valve leakage (CR1652573) and

safety valve RC-V117 leakage (CR1662418). NextEra continues to use the preventive

maintenance optimization process and the plant health committee reviews of system

health reports to focus on equipment issues. Self-assessments have been effective to

identify the need for additional actions to address service water system piping

degradation (CR 1 637 922).

Enclosure
.1

40A5

4046

33

Other Activities

(Closed) NRC Temporarv Instruction 2515/183, "Follow up to the Fukushima Daiichi

Nuclear Station Fuel Damage Event"

The inspectors assessed the activities and actions taken by NextEra to assess its

readiness to respond to an event similar to the Fukushima Daiichi nuclear plant fuel

damage event. This included (i) an assessment of NextEra's capability to mitigate

conditions that may result from beyond design basis events, with a particular emphasis

on strategies related to the spent fuel pool, as required by NRC Security Order Section

8.5.b issued February 25, 2002, as committed to in severe accident management

guidelines, and as specified by 10 CFR 50.54(hh); (ii) an assessment of NextEra's

capability to mitigate station blackout (SBO) conditions, as required by 10 CFR 50.63

and station design bases; (iii) an assessment of NextEra's capability to mitigate internal

and external flooding events, as specified by station design bases; and (iv) an

assessment of the thoroughness of the walkdowns and inspections of important

equipment needed to mitigate fire and flood events, which were performed by NextEra to

identify any potential loss of function of this equipment during seismic events possible for

the site. lnspection Report 05000443/201 1009 (ML1 1 1300174) documented detailed

results of this inspection activity.

(Closed) NRC Temporarv lnstruction 2515/184. "Availabilitv and Readiness Inspection of

Severe Accident Manaqement Guidelines (SAMGS)"

On May 20,2011, the inspectors completed a review of NextEra's severe accident

management guidelines (SAMG), implemented as a voluntary industry initiative in the

1990's, to determine (i) whether the SAMGs were available and updated, (ii) whether

NextEra had procedures and processes in place to control and update its SAMGS, (iii)

the nature and extent of NextEra's training of personnel on the use of SAMGS, and (iv)

licensee personnel's familiarity with SAMG implementation. The results of this review

were provided to the NRC task force chartered by the Executive Director for Operations

to conduct a near-term evaluation of the need for agency actions following the

Fukushima Daiichifuel damage event in Japan. Plant-specific results for Seabrook

Station were provided in an Attachment to a memorandum to the Chief, Reactor

lnspection Branch, Division of Inspection and Regional Support, dated May 27,2011

(M1111470361).

Meetinqs. Includinq Exit

On July 13,2011, the resident inspectors presented the results of the second quarter

routine integrated inspections to Mr. E. Metcalf and Seabrook Station staff. The

inspectors also confirmed with NextEra that no proprietary information was reviewed by

inspectors during the course of the inspection,

.2

ATTACHMENT: SUPPLEMENTAL INFORMATION

Enclosure
A-1

SU PPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

NextEra Personnel

J. Ball. Maintenance Rule Coordinator

K. Boehl, Health Physics Analyst

B. Brown, Supervisor, Civil Engineering

V. Brown, Senior Licensing Analyst

K. Browne, Operations Manager

M. Collins, Manager, Design Engineering

W. Cox, Radiological Engineer

R. Gutherie, Plant System Engineer

F. Haniffy, Senior Radiation Protection Analyst

L. Hansen, Plant Engineering

N. Levesque, Plant Engineering

E. Metcalf, Plant General Manager

W. Meyer, Radiation Protection Manager

M. O'Keefe, Licensing Manager

M. Nadeau, System Engineer, Control Building Air Handling

D. Perkins, Supervisor, Radiation Protection Technical Services

M. Scannell, Radiation Protection Technical Specialist

R. Sterritt, ALARA Coordinator

T. Vassallo, Principal Engineer - Nuclear

J. Walsh, Nuclear Steam Supply System, Supervisor

Attachment
A-2

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened:

05000443/201 1003-02

05000443/201 1 003-03

Opened and Closed:

05000443/201 1003-01

05000443/2011003-04

05000443/201 1 003-05

Closed:

05000443/25151183

05000443/25151184

Discussed:

None

URI

URI

NCV

NCV

NCV

lnadequate 50.59 Screening for Design Change

EC 272057

Operability Evaluation for Degraded

Concrete in ASR Affected Plant Structures

Inadequate Control of Combustible Materials

U nti mely Operability Determ i nation for Deg raded

Concrete Structures Housing Safety-Related

Equipment

I nadeq uate Operabil ity Determ i nation for Red uced

EDG HX Cooling Water Flow

Followup to the Fukushima Daiichi Nuclear Station

Fuel Damage Event (Section 4OA5.1)

Availability and Readiness Inspection of

Severe Accident Management Guidelines

(Section 4OA5.2)

TI

TI

Attachment
A-3

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

OP-AA-1 02-1002, Seasonal Readiness, Revision 0

OAl.42, Operations Department - Severe Weather Plan lmplementation

OS1200.03, Severe Weather Conditions, Revision 18

NM11800, Hazardous Condition Response and Recovery Plan

ON1490.09, Summer Readiness surveillance, Revision 5

ON0443.59, Yard Hydrant SemiAnnual Inspection, Revision 5

2011 Summer Readiness Site Certification

SBK 1 1-018, Nuclear Oversight Report - Summer Readiness

Condition Report; AR1655329, 1653764, 1607562

Work Order: 40083993, 40038324, 1384685, 40038436

ODI 61, Redeclaration / Joint Owners & NDDO Notification Guideline, Revision 47

ODI 90, 345kV Etectrical Disturbance Communication, Analysis, & Reporting Guideline,

Revision 6

051246.02, Degraded VitalAC Power, Revision 10

Seasonal Readiness Review - System Engineering

ER1.1, Classification of Emergencies, Revision 49

Operations Department Turnover Report

Daily Status Report

Station Operating Logs - various

Section 1R04: Equipment Aliqnment

OS 1412.09, Rev. 7, PCCW Monthly Flow Check

OX 1412.05, Rev. 8, Monthly PCCW Loop A Valve Verification

D rawi n g s 1 -CC-820205, 1 -CC- B 20206, 1 -CC-820207

UFSAR Section 9.2.2 Cooling for Reactor Auxiliaries

Work Orders 40040600, 40073132

OX1416.01, Monthly Service Water Valve Verification

OX1416.06, Service Water Discharge Valves Quarterly Test and 18 Month Position Verification

System Health Report - Service Water System

Operations Logs - various

PID: 1 -SW-820795, 1 -SW-820794, 1 -NHY-20247 6

UFSAR Section 9.2,7.3

Technical Specifications 3.7.4 Service Water System/Ultimate Heat Sink

Detailed System Text - Service Water System

Plant Engineering Action Plan Register

Operations Logs - various

OS1013.03, Residual Heat Removal System Train A Startup and Operation, Revision2l

OS1013.04, Residual Heat Removal System Train B Startup and Operation, Revision22

OS1001 .1 1, Reactor Coolant System Shutdown Level, Revision 5

OS1016.03, Service Water Train A Operation, Revision 11

OS1016.04, Service Water Train B Operation, Revision 13

OS1016.05, Service Water Cooling Tower Operation, Revision 19

Section 1R05: Fire Protection

Fire Protection Pre Fire Strategies

Fire lmpairment List

Technical Requirement 11 Fire Rated Assemblies

Technical Requirement 12 Fire Detection Instrumentation

Attachment
A-4

UFSAR Section 9.5.1 Fire Protection Systems

UFSAR Section 13.2.2.9 Fire Protection Personnel

OS1200.004, Fire Hazards Analysis for Affected Area I Zone - Appendix A

OS1200.00, Response to Fire or Fire Alarm Actuation, Revision 15

NUREG 1805 Chapter 8

FP 2.2, Control of Combustible Materials, Revision 13 (draft)

Response to NRC Fire Protection issue 1-SS-CP-1668

Fire Zone W-F-1A, 1B-Z & W-F-5-0

Station Operating Logs - various

Section 1R08: Inservice Inspection

ES1807.002 Rev 9, Liquid Penetrant Examination - Solvent Removable

ES1807.003 Rev 8, Magnetic Particle Examination

ES1807.001 Rev 7 CH 2, Visual Examination Procedure for Welding

ES03-01 -27 Rev 2, PDI Generic Procedure for Manual Ultrasonic through Wall and Length

Sizing of Ultrasonic lndications in Reactor Pressure VesselWelds (PDt-UT-7)

ES10-01-32 Rev 00, Remote lnservice Examination of Reactor Vessel Nozzle to Safe End,

Nozzle to Pipe, and Safe End to Pipe Welds Using the Nozzle Scanner

(PDl-lSl-254-SE-NB, Rev 1 )

ES1807.025 Rev 5, lnservice Inspection (lSl) Visual Examination Procedure (W-2)

ES1807.012 Rev 6, Ultrasonic Thickness Measurements

MA 10.3 Rev 5, Boric Acid Corrosion Control Program

P1-AA-102 Rev 3, Non-Safety Operating Experience Program

P1-AA-102-1001 Rev 4, Operating Experience Program Screening and Responding (tncoming)

AR 00569156, 81 Boric Acid Leak in Outlet lsolation Valve Packing Area

AR 01636221, Medium to Heavy Boric Acid Leakage from RHR Pump Suction Packing

AR 00210637, Boric Acid Leak at Packing on Valve FCV121

AR 0021 9427, Boric Acid Leak from Transmitter Fitting RC-FT-415

AR 00213435, Boric Acid Leak at Packing Charging Header Vent Valve 1-CS-V-836

AR 01640609, Reactor Vessel Hot Leg Post MSIP Exam (158 degree nozzle)

Examination Reports

1198488, Liquid Penetrant of SW-1814 Joint F0104, dwg SKEC145189-2000

1198488, Magnetic Particle Exam of SW-1814 Joint F0104, dwg SK-EC145189-2000

1 1 98488, Ultrasonic Examination of SW1 81 4-1 -156-24 Thickness Report

1-SW-1814-001 ,!/i'-2 Visual Examination Form, Service Water System

40055977-01, Magnetic Particle Exam Data Sheet (SW) Weld F0105, 106 and 107

01209165, Visual Examinat5ion (W-2) of Pressurizer Heater Sleeves

1208874, Remote Visual (W-2) Examination of RPV Bare Metal Upper Head

SP-SWOL-DS01, Ultrasonic Exam of Pressurizer Spray Nozzle (Phased Array)

S-SWOL-DS01, Ultrasonic Exam of Pressurizer Surge Nozzle (Phased Array)

Work Orders

WO 01 199620 01, 1-CS-F?V-121 Overhaul Valve Replace Valve Trim

WO 01202400 01, CS-V-836-B3 (Wet) Boric Acid Leak at Packing

WO 01198488 02, Weld Repair of Salt Service Water Line lnstall Repair Cap

WO 40055977 01, Fabrication of Salt Service Welded Pipe Replacement Spool Piece

Work Requests

WR 94002854, Boric Acid Leak at Charging Flow Control Valve FCV 121

Attachment
A-5

WR 940d3420, Charging Header Vent Valve Boric Acid Leak CS-V-836

WR 94002533. Fabricate and Install Reducer in Line 1814-01Salt Service Water

Weldinq Procedures (WPS) and Procedure Qualification Records (PQR)

WPS ES0815.004, Manual Gas tungsten (GTAW) and Shielded Metal Arc (SMAW) Welding

of Carbon Steel to Carbon Steel (Pl to Pl )

WPS ES0815.004, Manual SMAW of carbon steel to carbon steel PQR SBKI-8'15.004-1

Weld Procedure Qualification Record GTAW/SMAW of P1 to P1 with Post Weld Heat

Treatment (PWHT)

PQR SBKI -815.004-2 WPS for P1 to P1 without PWHT

UC 371 & 391, Welder Performance Qualification Record Review to use ES0815.004-1

Drawinos

SK-EC270505-2000, Installation Detail Service Water Piping Repair (SW 1814)

SK-EC270505-2001, Fabrication Detail Service Water Piping Repair (SW 1814)

Miscellaneous

AR 220564, Self Assessment - Boric Acid Corrosion Control Program

2010 3rd Qtr, Program Health Report - Boric Acid Corrosion Control Program

2010 4th Qtr, Program Health Report - Boric Acid Corrosion Control Program

CR 05-1 1634, Engineering Evaluation for 1-CS-FCV-121

CRO1636130, UT results of SW Piping Indicates Wall Thinning

CR (AR 00213435), Boric Acid Corrosion Evaluation (EDl 30560) Valve 1-CS-V-836

CR (AR 210637), Boric Acid Corrosion Control ASME Bolting Evaluation 4-10-2011

MSE#:05-040, Maintenance Support Evalfor Valves CS-FCV-121 and 1-CS-HCV-182

EC145189, ASME Xl Repair/Replacement Plan Traveler Component SW-1814

EC 271779 R0, Temp Installation for Repair of Section of SW-1814-001

EDI 30560, Boric Acid Corrosion Evaluation of Valve 1-CS-V-836 Vent Valve

SllR, Inservice Inspection Program Plan for 3'o Ten Year lnterval

Section 1R11: Licensed Operator Requalification Proqram

OS1000.02, Plant Startup from Hot Standby to Minimum Load, Revision 20

OS1000.05, Power Increase, Revision 16

OS1000.07, Approach to Critical, Revision 10

OS1007.01, Automatic and Manual Rod Control, Revision 10

OS1056.03, Containment Penetrations, Revision 6

OS1213.01, Loss of RHR While in Reduced Inventory, Revision

ON1O31 .02, Starting and Phasing the Turbine Generator, Revision 26

ON1031.13, Post Maintenance Turbine Startup, Revision 12

RS1735, Reactivity Calculations, Revision 4

ODt.101, Guarded Equipment Recommendations for Refueling Outages, Revision 5

ODl.82, Mode Change Notice, Revision 15

Section 1 Rl2: Maintenance Effectiveness

System Health Report - RHR system

Maintenance Rule Performance and Scope Report

UFSAR Section

Condition Reports 1612061, 1632409, 1633034, 1636533

PODs for CR1 61 2061 | 16324091 1633034

Drawing 1-SW-820795

OR14 Service Water Inspections / Results

Attachment
A-6

OR14 Service Water Piping Assessment

AR1939781 - SW Train B Pump House Inspection

Work Orders 40080265, 0062557 1 02, 4007 8949

EC272058, SW Pipe Repairs

Design Engineering Review for Service Water Pipe SW-1814 83 Day UT Results

AR1637922 - DQS of Service Water Corrective Actions

System Health Reports - Service Water System

Plant Engineering Action Register

Condition Reports 2010-201 1

Work Requests 2O1O-201 1

Station Operating Logs - various

Section 1R13: Maintenance Risk and Emerqent Work

OR14 Outage Schedule Initial Shutdown Risk Review Rev. 0

OR14 SW Extent of Condition Inspection Matrix 411312011

WM-AA-1000 Work Activity Risk Management Rev. 6

OS 1016.1 1 Contingency Ocean Pump Restoration for SW Work Activities with Ocean

Service Water Pumps not in Service. Revision 01

UFSAR 9.2.5 Ultimate Heat Sink

Drawings 1 -SW-820795, 1 -SW-B -8.20794

M-Rule a(4), Risk Assessment Reports

Station Operating Logs - various

AR 1 640932, 1610327, 1639921, 1 631 769, 1631776, 1640932, 1640932

wo 1 199040, 1205038, 1203446, 249348, 00626035

Lift Plan and Rigging Evaluation - A RHR Pump Roof Plug and RHR motor

TS - Various

RHR leak rate summary

Plant Engineering Register

EX1801.002, Leakage Reduction Program Surveillance, Revision 9

MS0523.24, Ingersoll-Rand Residual Heat Removal Pump Maintenance, Revision 7

OS1213.02, Loss of RHR while Operating at Reduced lnventory or Midloop Conditions,

Revision 12

OS1215.05, Loss of Refueling Cavity Water, Revision 15

OS1213.01, Loss of RHR During Shutdown Cooling, Revision 15

OS1056.03, Containment Penetrations, Revision 6

OD|.103, Conduct of Infrequently Performed Tests or Evolutions, Revision 0

OD1.101, Guarded Equipment Recommendations for Refueling Outages, Revision 5

Work Orders 40086371 Tasks 1, 2, 3 and 4, WO 1382815

Adverse Condition Monitoring Plan for Sl-V82 dated 618111

Sl-V-82 Operational Decision Making

MS0526.09, On Stream Leak Repairs, Revision 4

Insulation Removal Evaluation for Sl-V-82

lN93-90, Unisolatable Reactor Coolant System Leak Following Repeated Applications of

Leak Sealant

Section 1 Rl5: Operabilitv Evaluations

ODM, Operational Decision Making for RC-V-117 Leak (AR 1633034)

Station Operating Logs - various

Adverse Condition Monitoring Plan

Plant Engineering Register -

IMC ggo0, Operability Determinations and Functionality Assessments for Resolution of

Degraded or Nonconforming Conditions Adverse to Quality or Safety

Attachment
A-7

Seabrook 1 0CFR5059 Resource Manual

NEI 96-07, Guidelines for 10 CFR 50.59 lmplementation, Revision 1

EN-AA-203-1001 , Operability Determinations / Functionality Assessments, Revision 5

Prompt Operability Determinations for AR 581434, 1664399

Condition Reports 581434, 1641413, 1644074, 1644399, 1629282, 1664708

Calculations C-S-1-10156, C-S-1-101 50, C-S-1-10155

Design Change EC250348 , 272057,

CFR 50.59 Screen for EC272057

OD/FA-11-0005, Reduced Concrete Modulus in Below Grade Walls in CEB, EHE EV, EFWPH,

B DG FOST Room

Section 1R18: Plant Modifications

Permanent ModificationEC12T3S, Seabrook Substation Reliability Upgrade Project

Permanent Modification EC145280, Seabrook Substation Reliability Upgrade Project

Phase ll

Foreign Print 100606

5059 Screen for EC1 45280

EC145280 Procedure and Training Needs

Temporary Modification EC272512,Team Inc Repair for Sl-V-82

Temporary Modification EC 272290, Install Varistor in Panel for CBA-CP-177

Switchyard Work Orders 01 384069-12, 01384069-1 4

EN 10-01 -20,345KV Bus#6 System 1 Testing

EN 10-01-22,345KV Switchyard Circuit Load Test

LN0561.45, 345KV Bus Primary and Line Cable Differential Relay Testing & Calibration

Condition Report 1640003 Apparent Cause Report - Breaker 294 Open During Switchyard

Modifications

SORC Meeting #11-015

Condition Report 1652573, 1 652598

IMC 9900, On-Line leak Sealing Guidelines for ASME Code Class I and 2 Components

EPRI Technical Report NP-6523-D, On-Line Leak Sealing: A guide for Nuclear Power Plant

Maintenance Personnel

Boric Acid Corrosion Control ASME SA 453 Grade 660 Bolting Evaluation for Sl-V-82,511912011

Form NPV-1 Manufacturer's Data Report BS-72305-AR6-AR1 for 6" 1655 Swing check Valve,

9t26t78

Westinghouse Certification of Vendor Test Results Valve Studs & Nuts, 9128177

Work Order 40086371

Work Request 91W004461

Team lndustrial Services Work Order 237-05131

Drawing 1 -Sl-D20446, 1 -S l-25 1 - 1 3, D-048 08-837 4D 48

Foreign Print 52914, 100630,

Calculation C-S-1 -45864, Piping Qualification for Leak Repair of Sl-V-82

Calculation C-S-1-10158, Sl-V-82 Injection Pressure Bolting Evaluation, Revision 0,5119111

Calculation C-S-1-10158, Sl-V-82 Injection Pressure Bolting Evaluation, Revision 1,5120111

Section 1Rl9: Post Maintenance Testinq

OX1413.01, A Train RHR Quarterly Flow and Valve Stroke Test and 18 Month Valve

Stoke Observation, Revision 16

OX1456.86, Operability Testing of IST Pumps, Revision 4

wo 40083875, 1205107, 1205112, 1205043, 40084983, 1203622, 40068999 , 620087 ,

1 1 94007, 12037 97, 12037 98

Attachment
A-B

AR 1 65591 0, 1 656350, 1660228, 1 660236, 164187 5, 1642125, 1 631 81 1, 1647 943,

1647949, 1647983, 1646546, 1645417, 1633233, 1649428,

EC24938, 27 2291, 27 2303, 0002466

08MSE055

PtD D20662

Technical Specification - various

Plant Engineering Action Plan Register -

Station Operating Logs - various

WO 40082703 Task 6.40082746

Foreign Prints 31 417, 31 425, 31 61 0

Foreign Print 31919, Emergency Power Sequencing System

EPS Logic Drawing 2948-1020, Sheets 1, 3, 4, 6, 10

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

OX1401 .04, Reactor Coolant system Pressure lsolation Valve Leakage Rate Tests, Revision 0

DCR 98-039, CBA Replacement Project

Condition Report 1645405 - DG A EPS Did Not Fully Sequence

Failure Investigation Process for DG A EPS (AR1645405)

Seabrook Train A Emergency Power Sequencer Troubleshooting

Section 1R20: Refuelinq and Outaqe Activities

Action Request 1640003

Clearances MTO, 1 -CC-V-1 1 12, 1-CC-V-1092

Control Room Narrative Logs

Condition Reports

and Repair, 515111

Engineering Evaluation EE-1 1-02, OR14 Outage Schedule Initial Shutdown Risk

Foreign Print98727 - Reactor Vessel Outlet Nozzle DM Weld Flaw Evaluation in

Main Control board and MPCS Plant Parameter Displays and Trends

MSIP Configuration (AR1 644106), 4121 11 1

Mode Change Report Mode 6 to Mode 5

Mode Change Report Mode 5 to Mode 4

Mode Change Report for Modes 3,2, 1

Open Condition Reports and Actions with Mode Restrictions

Operations Component Deviation Log - various dates

Outage and Operations Department Turnover Sheets

MS0504.15, Reactor Vessel Upper Internals Assembly Installation, Revision 12

MS0504.16, Upper Internals Installation, Revision 11

OD1.101, Guarded Equipment Recommendations for Refueling Outages, Revision 5

ODl.82, Mode Change Notice, Revision 15

OM-AA-O4, OR14 Scope Change Meeting Report

OM-AA-04, Plant Readiness for Operations, Revision 2

ON1031.02, Starting and Phasing the Turbine Generator, Revision 26

ON1031.13, Post Maintenance Turbine Startup, Revision 12

OP-AA-103-1000, Reactivity Management, Revision 0

OR14 Mode Hold / Milestone Report, 412512011

OS1000.02, Plant Startup from Hot Standby to Minimum Load, Revision 20

OS1000.03, Plant Shutdown from Minimum Load to Hot Standby, Revision 18

OS1000.04, Plant Cooldown from Hot Standby to Cold Shutdown, Revision 30

OS1000.05, Power Increase, Revision 16

OS1000.06, Power Decrease, Revision 15

OS1000.07, Approach to Critical, Revision 10

OS1000.09, Refueling Operation, Revision 14

Evaluation. 3118111

the Post

Attachment
A-9

OS1000.12, Operation with RCS at Reduced Inventory/Midloop Conditions, Revision 9

OS1000.14, Reactor Coolant system Evacuation and Fill, Revision 10

OS1007.01, Automatic and Manual Rod Control, Revision 10

OS1001 .1 1 , Reactor Coolant System Shutdown Level, Revision 5

OS1013.03, Residual Heat Removal System Train A Startup and Operation, Revision 21

OS101 4.02, Operation of Spent Fuel Cooling and Purification System, Revision '15

OS1015.05, FuelTransfer System and Upender Operation, Revision 7

OS1015.07, Spent Fuel Bridge Assembly Operation, Revision 16

OS1015.1 8, Setting Containment Integrity for Mode lV Entry, Revision 6

OS1056.03, Containment Penetrations, Revision 6

OS1213.01, Loss of RHR While in Reduced Inventory, Revision

RD0717, Automated EXCEL Core Offload Tracking, Revision 0

RS0721, Refueling Administrative Control, Revision 9

RS1735, Reactivity Calculations, Revision 4

Technical Specification 3.9

Technical Specification and Commitment Logs

WO 01203652, Containment and Containment Spray Recirculation Sump Surveillance, Slllll

Work Order 1209198,

Station Operating Logs - various

Section 1 R22: Surveillance Testinq

EE 11-003, Containment Spray System Spray Nozzle Test Surveillance Frequency

Modification

wo 40082703, 40077896, 40078 102, 40077894, 40077892, 40049329, 40049337 ,

01 21 051 23, 01209191, 01 2091 90, 01 2091 99, 01 2091 98, 01203722

Calc C-S-1-50006,

Specification 9763.006-238-5 Primary Component Cooling Water Pumps

AR 1645405,

DBD-ESF-1, Engineered Safety Features Response Times, Revision 1

Technical Specifications 3.4.6.2.f ,4.5.2.e and 4.0.5

OR14 Local Leak Rate Test Summary, 412812011

RE1707-B-R, Shutdown Margin Verification , 5117111

Subcritical Rod Worth Measurement Data Analysis System Results 511712011

Westinghouse Letter NAH-1 1-42, Cycle 15 Subcritical Physics Testing, 611612011

Westinghouse Letter LTR-NRC-O8-13, SER Compliance with WCAP-16260-P-A, 4115l20OB

WCAP-16260-P-A, The Spatially Corrected Inverse Count Rate (SCICR) Method for Subcritical

Reactivity Measurement, September 2005

Station Operating Logs - various

Section 2RS01: Radioloqical Hazard Assessment and Exposure Gontrols

HD0958.03, Personnel Survey and Decontamination Techniques

HD0958.04, Posting of Radiologically Controlled Areas

HD0958.17, Performance of Routine Radiological Surveys

HN0958.25, High Radiation Area Controls

HD0958.30, Inventory and Control of Locked or Very High Radiation Area Keys and Locksets

Condition Reports 1638564, 1640938, 1644445, 1640938, 1626367 , 1612661 , 1640268,

1650347 , 1618932, 1623142, 1629512, 1649794, 1604791, 1642643, 1 626363, 1 639705,

1 651 585, 1651072, 1 651 584

Attachment
A-10

Section 2RS02: Occupational ALARA Planninq and Controls

RP-AA-1 04, ALARA Program

RP-AA-1 04-1 000, ALARA lmplementing Procedure

RP-AA-1 01-2004, Method for Monitoring and Assigning Effective Dose Equivalent for

High Dose Gradient Work

Condition Reports: See Section 2RS01

Section 2RS03: In-Plant Airborne Radioactivitv Gontrol and Mitiqation

HD0955.01, Analysis of Smears and Air Samples

HD0958.01, Air Sampling

HD0965.12, Respiratory Protection lssue and Use

Condition Reports: See Section 2RS01

Section 2RS04: Occupational Dose Assessment

HD0955.54, Operation of the TSA Model SPM-906 Portal Monitor

HD0955.62, Use of the Argos 4AlB

HD0958.1 9, Evaluation of Dosimetry Abnormalities

HD0958.27, Dose Assessment for Personnel Contaminations

HN0958.39, Multi-Badge Control & Exposure Tracking

HD0958.41, Blind Spiking of TLDs

HD0958.42, Determination and Controlof Dose to an Embryo/Fetus

HD0958.49, Response Protocols for Whole Body Counting and Personnel Contamination

Monitoring

HD0961.29, Internal Dosimetry Assessment

HD0963.28, Calibration and Troubleshooting of MGP Instruments DMC 2000 Dosimeters

HD0992.02, lssuance and Control of Personnel Monitoring Devices

RP-AA-1 01-2004, Method for Monitoring and Assigning Effective Dose Equivalent for

High Dose Gradient Work

Condition Reports: See Section 2RS01

Miscellaneous Documents:

NVLAP Certification Records, Personnel Dosimetry Performance Testing

Annual Review Report of the 2010 10 CFR Part 61 Radionuclide Analysis

Electronic Dosimeter Dose/Dose Rate Alarm Reports, January - May 2011

Top Ten Individual Exposure Records for 2Q11

Portable HEPA Inventory & Test Records

EPRI Standard Radiation Monitoring Program Data Summary for primary piping

Reactor Coolant System OR-14 Clean Up Data

Nuclear Oversight Field Observation OR-14 Daily Quality Summary Reports

HPSTID 09-01 1, Use of Effective Dose Equivalent for Steam Generator Nozzle Dam Work

HPSTID 08-13, Calibration of the FastScan WBC System

OR-14 ALARA Plans (AP)fuVork-ln-Prooress (WlP) Reviews:

AR 1 1-01, reactor vessel disassembly/re-assembly

AR 1 1-02, steam generator (S/G) eddy current testing/tube plugging

AR 11-03, S/G secondary side maintenance

AP 1 1-11, scaffolding Installation/Removal

AP 1 1-13, reactor vessel bare metal visual inspections

Attachment
A-11

ACI

ADAMS

ALARA

AMS

AP

AR

ASME

ASR

BACC

CAP

CB/ET

CEB

CEDE

CR

DG

DPW

ECT

EDEX

EDG

EFW

FBL

FHB

FPP

GTAW

HEPA

IMC

IP

tsl

LHRA

MR

MSIP

MT

NCV

NDE

NFPA

NRC

NVLAP

OR

OD

ODs

OM

PAB

PARS

PCCW

PDI

PMT

POD

LIST OF ACRONYMS

American Concrete Institute

Agency-wide Documents Access and Management System

As Low As is Reasonably Achievable

Airborne Monitoring System

ALARA P|ans

Action Request

American Society of Mechanical Engineers

Alkali-silica Reaction

Boric Acid Corrosion Control (Program)

Corrective Action Program

Control Building/Electric Tunnel

Containment Enclosure Building

Committed Effective Dose Equivalent

Condition Report

Diesel Generator

Declared Pregnant Workers

Eddy Current Testing

External Effective Dose Equivalent

Emergency Diesel Generator

Emergency Feedwater

Fire Brigade Leader

Fuel Handling Building

Fire Protection Program

Gas Tungsten Arc Welding

High Efficiency Particulate Air

Inspection Manual Chapter

Inspection Procedure

In-service Inspection

Locked High Radiation Areas

Maintenance Rule

Mechanical Stress lmprovement Process

Magnetic Particle Test

Non-cited Violation

Non-Destructive Examination

National Fire Protection Association

U.S. Nuclear Regulatory Commission

National Voluntary Laboratory Accreditation Program

Outage for Refueling

Operability Deficiency

Operability Determinations

Operations Management

Primary Auxiliary Building

Publicly Available Records

Primary Component Cooling Water

Performance Demonstration Initiative

Post-maintenance Testing

Prompt Operability Determination

Attachment
PQR

PT

A-12

Procedure Qualification Record

Penetrant Test

Pressurized Water Reactor

Reactor Coolant Pump

Reactor Coolant System

Residual Heat Removal

Reactor Pressure Vessel

Radiation Safety Committee

Radiation Work Permit

Severe Accident Management Guidelines

Station Blackout

Significance Determination Process

Spent Fuel Pool

Steam Generator

Shift Manager

Shielded Metal Arc Welding

Spent Fuel Pool

Scintillation Portal Monitor

Subcritical Rod Worth Measurement

Structures, Systems or Components

Service Water

Service Water Pump

Temporary Instruction

Thermolum inescent Dosimeter

Technical Specifications

Updated Final Safety Analysis Report

Ultrasonic Testing

Very High Radiation Area

Visual Test

Work-ln-Progress

Work Order

Weld Procedure Specification

Work Request

PWR

RCP

RCS

RHR

RPV

RSC

RWP

SAMG

SBO

SDP

SFP

SG

SM

SMAW

SPF

SPM

SRWM

SSC

SW

SWP

TI

TLD

TS

UFSAR

UT

VHRA

VT

W-I-P

WO

WPS

WR

Attachment