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{{Adams
#REDIRECT [[IR 05000443/2011003]]
| number = ML112241543
| issue date = 08/12/2011
| title = IR 05000443/2011003; on 04/01/2011-06/30/2011; Seabrook Station, Unit No. 1; Routine Integrated Report; Fire Protection; Operability Evaluations
| author name = Burritt A
| author affiliation = NRC/RGN-I/DRP/PB3
| addressee name = Freeman P
| addressee affiliation = NextEra Energy Seabrook, LLC
| docket = 05000443
| license number = NPF-086
| contact person = Burritt A  RGN-I/DRP/PB3/610-337-5069
| document report number = IR-11-003
| document type = Inspection Report, Letter
| page count = 49
}}
See also: [[see also::IR 05000443/2011003]]
 
=Text=
{{#Wiki_filter: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 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.
  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 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 .
.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 (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
                                                                                      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
 
                                              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)
.4 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 ity Determ inations/Functional
                                                        l
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-FE-
6191 , 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-AA-
      203-1001 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 post-
    modification testing was adequate to ensure the SSCs would function properly. The
                                                                                          Enclosure
 
                                              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.
  b.  Findinqs
      No findings were identified.
.2  Temporarv Modification    - EC 272290: lnstall Varistor in Panel for CBA-CP-177
  a.  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.
  b.  Findinos
    No findings were identified.
.3  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.
                                                                                      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-CS-
          FCV-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-in-
depth 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)
      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.
b.  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
 
                                                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.
  b.  Findinos
      No findings were identified.
4.    OTHER ACTIVITIES
4c.A2 ldentification and Resolution of Problems (71152    - 2 sample)
.1    Review of ltems Entered into the Corrective Action Prooram
a.  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.
  b.  Findinqs
      No findings were identified.
.2    Semi-Annual Review to ldentifv Trends
a.  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
                                                                                        Enclosure
 
                                            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 Pl-AA-207, "Trend Coding and
  Analysis."
b. 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 RC-
  V147, 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
 
                                                33
40A5 Other Activities
.1  (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.
.2  (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).
4046 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,
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      URI        lnadequate 50.59 Screening for Design Change
                                      EC 272057
05000443/201 1 003-03      URI        Operability Evaluation for Degraded
                                      Concrete in ASR Affected Plant Structures
Opened and Closed:
05000443/201 1003-01      NCV        Inadequate Control of Combustible Materials
05000443/2011003-04        NCV        U nti mely Operability Determ nation for Deg raded
                                                                    i
                                      Concrete Structures Housing Safety-Related
                                      Equipment
05000443/201 1 003-05      NCV        I nadeq uate Operabil ity Determ i nation for Red uced
                                      EDG HX Cooling Water Flow
Closed:
05000443/25151183          TI        Followup to the Fukushima Daiichi Nuclear Station
                                      Fuel Damage Event (Section 4OA5.1)
05000443/25151184          TI        Availability and Readiness Inspection of
                                      Severe Accident Management Guidelines
                                      (Section 4OA5.2)
Discussed:
None
                                                                                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 l 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
ODl.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 and Repair, 515111
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
Engineering Evaluation EE-1 1-02, OR14 Outage Schedule Initial Shutdown Risk Evaluation. 3118111
Foreign Print98727 - Reactor Vessel Outlet Nozzle DM Weld Flaw Evaluation in the Post
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
                                                                                    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
                            LIST OF ACRONYMS
ACI  American Concrete Institute
ADAMS Agency-wide Documents Access and Management System
ALARA As Low As is Reasonably Achievable
AMS  Airborne Monitoring System
AP    ALARA Plans
AR    Action Request
ASME  American Society of Mechanical Engineers
ASR  Alkali-silica Reaction
BACC  Boric Acid Corrosion Control (Program)
CAP  Corrective Action Program
CB/ET Control Building/Electric Tunnel
CEB  Containment Enclosure Building
CEDE  Committed Effective Dose Equivalent
CR    Condition Report
DG    Diesel Generator
DPW  Declared Pregnant Workers
ECT  Eddy Current Testing
EDEX  External Effective Dose Equivalent
EDG  Emergency Diesel Generator
EFW  Emergency Feedwater
FBL  Fire Brigade Leader
FHB  Fuel Handling Building
FPP  Fire Protection Program
GTAW  Gas Tungsten Arc Welding
HEPA  High Efficiency Particulate Air
IMC  Inspection Manual Chapter
IP    Inspection Procedure
tsl  In-service Inspection
LHRA  Locked High Radiation Areas
MR    Maintenance Rule
MSIP  Mechanical Stress lmprovement Process
MT    Magnetic Particle Test
NCV  Non-cited Violation
NDE  Non-Destructive Examination
NFPA  National Fire Protection Association
NRC  U.S. Nuclear Regulatory Commission
NVLAP National Voluntary Laboratory Accreditation Program
OR    Outage for Refueling
OD    Operability Deficiency
ODs  Operability Determinations
OM    Operations Management
PAB  Primary Auxiliary Building
PARS  Publicly Available Records
PCCW  Primary Component Cooling Water
PDI  Performance Demonstration Initiative
PMT  Post-maintenance Testing
POD  Prompt Operability Determination
                                                          Attachment
 
                                    A-12
PQR  Procedure Qualification Record
PT    Penetrant Test
PWR  Pressurized Water Reactor
RCP  Reactor Coolant Pump
RCS  Reactor Coolant System
RHR  Residual Heat Removal
RPV  Reactor Pressure Vessel
RSC  Radiation Safety Committee
RWP  Radiation Work Permit
SAMG  Severe Accident Management Guidelines
SBO  Station Blackout
SDP  Significance Determination Process
SFP  Spent Fuel Pool
SG    Steam Generator
SM    Shift Manager
SMAW  Shielded Metal Arc Welding
SPF  Spent Fuel Pool
SPM  Scintillation Portal Monitor
SRWM  Subcritical Rod Worth Measurement
SSC  Structures, Systems or Components
SW    Service Water
SWP  Service Water Pump
TI    Temporary Instruction
TLD  Thermolum inescent Dosimeter
TS    Technical Specifications
UFSAR Updated Final Safety Analysis Report
UT    Ultrasonic Testing
VHRA  Very High Radiation Area
VT    Visual Test
W-I-P Work-ln-Progress
WO    Work Order
WPS  Weld Procedure Specification
WR    Work Request
                                            Attachment
}}

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