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IR 05000443/2011003; on 04/01/2011-06/30/2011; Seabrook Station, Unit No. 1; Routine Integrated Report; Fire Protection; Operability Evaluations
	ML112241543
Person / Time
Site:	Seabrook
Issue date:	08/12/2011
From:	Arthur Burritt; Reactor Projects Branch 3
To:	Freeman P; NextEra Energy Seabrook
	Burritt A RGN-I/DRP/PB3/610-337-5069
References
	IR-11-003
	Download: ML112241543 (49)
	v • d • e

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 (RIORAMAlL Resource)

D. Lew, DRA (RIORAMAlL Resource)

D. Roberts, DRP (RIDRPMAlL Resource)

J. Clifford, DRP (RIDRPMAlL 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

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

NPF-86

Report No.: 05000443/201 1 003

NextEra Energy Seabrook, LLC

Facility: Seabrook Station, Unit No.1

Location: Seabrook, New Hampshire 03874

Dates: April 1 ,2011through June 30, 2011

Inspectors: 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

Approved by: 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 ACTIVlTIES.............. ......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 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ). 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.

b. Findinqs

No findings were identified.

,2 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

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).

b. Findinqs

1. 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:

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)

a. 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.

b. Findinos

No findings were identified.

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 non-

destructive 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 pre-

engineered 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 SK-

EC145189-2000. 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

14

. AR1641413, evaluation of containment shellwith craze cracking in concrete,

April 20, 2011.

. AR1644074, operability of containment enclosure 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

b. 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 581434581434in 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 .

.1 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 (AR581434581434. 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

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 (AR581434581434 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 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ).

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 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ) 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

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