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{{Adams
#REDIRECT [[IR 05000443/2014002]]
| number = ML14127A376
| issue date = 05/06/2014
| title = IR 05000443-14-002; 01/01/2014 - 03/31/2014; Seabrook Station, Unit No. 1; Operability Determinations and Functionality Assessments
| author name = Dentel G
| author affiliation = NRC/RGN-I/DRP/PB3
| addressee name = Ossing M, Walsh K
| addressee affiliation = NextEra Energy Seabrook, LLC
| docket = 05000443
| license number = NPF-086
| contact person = Dentel G
| document report number = IR 14-002
| document type = Inspection Report, Letter
| page count = 37
}}
See also: [[see also::IR 05000443/2014002]]
 
=Text=
{{#Wiki_filter:K. Walsh
                                              UNITED STATES
                                    NUCLEAR REGULATORY COMMISSION
                                                  REGION I
                                    2100 RENAISSANCE BLVD., SUITE 100
                                      KING OF PRUSSIA, PA 19406-2713
                                                May 6, 2014
      Mr. Kevin Walsh
      Site Vice President
      Seabrook Nuclear Power Plant
      NextEra Energy Seabrook, LLC
      c/o Mr. Michael Ossing
      P.O. Box 300
      Seabrook, NH 03874
      SUBJECT:        SEABROOK STATION, UNIT NO. 1 - NRC INTEGRATED INSPECTION
                      REPORT 05000443/2014002
      Dear Mr. Walsh:
      On March 31, 2014, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection
      at Seabrook Station, Unit No. 1. The enclosed inspection report documents the inspection
      results, which were discussed on April 10, 2014, with you and other members of your staff.
      The inspection examined activities conducted under your license as they relate to safety and
      compliance with the Commissions rules and regulations and with the conditions of your license.
      The inspectors reviewed selected procedures and records, observed activities, and interviewed
      personnel.
      This report documents one NRC-identified finding of very low safety significance (Green). This
      finding was determined to involve a violation of NRC requirements. However, because of the
      very low safety significance, and because it was entered into your corrective action program
      (CAP), the NRC is treating the finding as a non-cited violation (NCV), consistent with Section
      2.3.2.a of the NRC Enforcement Policy. If you contest the subject or severity of any NCV in this
      report, you should provide a response within 30 days of the date of this inspection report, with
      the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control
      Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region I; the
      Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington,
      DC 20555-0001; and the NRC Resident Inspector at Seabrook Station. In addition, if you
      disagree with the cross-cutting aspect assigned to the finding in this report, you should provide
      a response within 30 days of the date of this inspection report, with the basis for your
      disagreement, to the Regional Administrator, Region I, and the NRC Resident Inspector at
      Seabrook Station.
      Additionally, as we informed you in the most recent NRC integrated inspection report, cross-
      cutting aspects identified in the last six months of 2013 using the previous terminology were
      being converted in accordance with the cross-reference in Inspection Manual Chapter (IMC)
      0310. Section 4OA5 of the enclosed report documents the conversion of these cross-cutting
 
K. Walsh                                            2
aspects which will be evaluated for cross-cutting themes and potential substantive cross-cutting
issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment review. If you
disagree with the cross-cutting aspect assigned, you should provide a response within 30 days
of the date of this inspection report, with the basis for your disagreement, to the Regional
Administrator, Region I, and the NRC Resident Inspector at Seabrook Station.
In accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390 of the NRCs Rules
of Practice, a copy of this letter, its enclosure, and your response (if any) will be available
electronically for public inspection in the NRCs Public Document Room or from the Publicly
Available Records component of the NRCs Agencywide Documents Access Management
System (ADAMS). ADAMS is accessible from the NRC website at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room).
                                                Sincerely,
                                                  /RA/
                                                Glenn T. Dentel, Chief
                                                Reactor Projects Branch 3
                                                Division of Reactor Projects
Docket No.      50-443
License No.    NPF-86
Enclosure:      Inspection Report No. 05000443/2014002
                w/ Attachment: Supplemental Information
cc w/encl:      Distribution via ListServ
 
 
ML14127A376
                                                Non-Sensitive                              Publicly Available
    SUNSI Review
                                                Sensitive                                  Non-Publicly Available
OFFICE          RI/DRP                  RI/DRP                  RI/DRP
                PCataldo/GTD for per    RBarkley/GTD for per
NAME                                                            GDentel/GTD
                email                  email
DATE            05/05/14                05/06/14                05/06/14
                                             
                                      1
              U.S. NUCLEAR REGULATORY COMMISSION
                                  REGION I
Docket No.:  50-443
License No.: NPF-86
Report No.:  05000443/2014002
Licensee:    NextEra Energy Seabrook, LLC
Facility:    Seabrook Station, Unit No.1
Location:    Seabrook, New Hampshire 03874
Dates:      January 1, 2014 through March 31, 2014
Inspectors:  P. Cataldo, Senior Resident Inspector
            C. Newport, Resident Inspector
            E. Burket, Emergency Preparedness Inspector
            T. Burns, Reactor Inspector
            B. Dionne, Health Physicist
            W. Cook, Senior Reactor Analyst
Approved by: Glenn T. Dentel, Chief
            Reactor Projects Branch 3
            Division of Reactor Projects
                                                        Enclosure
 
                                                                2
                                              TABLE OF CONTENTS
SUMMARY .................................................................................................................................... 3
REPORT DETAILS ....................................................................................................................... 4
1.  REACTOR SAFETY .............................................................................................................. 4
  1R01  Adverse Weather Protection ....................................................................................... 4
  1R04  Equipment Alignment .................................................................................................. 4
  1R05  Fire Protection ............................................................................................................. 5
  1R06  Flood Protection Measures ........................................................................................ 6
  1R11  Licensed Operator Requalification Program ............................................................... 6
  1R12  Maintenance Effectiveness ......................................................................................... 7
  1R13  Maintenance Risk Assessments and Emergent Work Control .................................. 7
  1R15  Operability Determinations and Functionality Assessments ....................................... 8
  1R18  Plant Modifications ................................................................................................... 10
  1R19  Post-Maintenance Testing ....................................................................................... 11
  1R22  Surveillance Testing ................................................................................................. 11
  1EP4  Emergency Action Level and Emergency Plan Changes ......................................... 12
  1EP6  Drill Evaluation ......................................................................................................... 12
2.  RADIATION SAFETY .......................................................................................................... 13
  2RS1  Radiological Hazard Assessment and Exposure Controls ....................................... 13
  2RS2  Occupational ALARA Planning and Controls ........................................................... 16
4.  OTHER ACTIVITIES ............................................................................................................ 17
  4OA1  Performance Indicator Verification ........................................................................... 17
  4OA2  Problem Identification and Resolution ..................................................................... 18
  4OA3  Follow-Up of Events and Notices of Enforcement Discretion ................................... 22
  4OA5  Other Activities .......................................................................................................... 23
  4OA6  Meetings, Including Exit ............................................................................................ 24
ATTACHMENT: SUPPLEMENTARY INFORMATION................................................................ 24
SUPPLEMENTARY INFORMATION ........................................................................................ A-1
KEY POINTS OF CONTACT .................................................................................................... A-1
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED ..................................... A-1
LIST OF DOCUMENTS REVIEWED ........................................................................................ A-1
LIST OF ACRONYMS ............................................................................................................. A-10
                                                                                                                                Enclosure
 
                                                    3
                                              SUMMARY
IR 05000443/2014002; 01/01/2014-03/31/2014; Seabrook Station, Unit No. 1; Operability
Determinations and Functionality Assessments.
This report covered a three-month period of inspection by resident inspectors and announced
inspections performed by regional inspectors. Inspectors identified one finding of very low
safety significance (Green), which was an NCV. The significance of most findings is indicated
by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using IMC
0609, Significance Determination Process (SDP), dated June 2, 2011. Cross-cutting aspects
are determined using IMC 0310, Components Within Cross-Cutting Areas, dated December
19, 2013. All violations of NRC requirements are dispositioned in accordance with the NRCs
Enforcement Policy, dated June 7, 2012. The NRCs program for overseeing the safe operation
of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight
Process, Revision 4.
Cornerstone: Mitigating Systems
  Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V,
    Procedures, because NextEra did not ensure adequate separation was maintained
    between temporary scaffolding and safety-related equipment. Specifically, six instances
    of scaffolding installed in the plant were identified with less than the minimum standoff
    distance to safety-related equipment specified in NextEra procedures and no corresponding
    engineering evaluation to support these deviations. NextEra entered this NCV into their
    CAP as AR 01933827 and assessed the six deviations for any impact on the associated
    safety-related systems.
    This performance deficiency was considered more than minor because it affected the
    protection against external factors attribute of the Mitigating Systems cornerstone and its
    objective to ensure the availability, reliability, and capability of systems that respond to
    initiating events to prevent undesirable consequences. Specifically, NextEra did not
    evaluate scaffolding installations when insufficient separation to safety-related equipment
    existed after procedural requirements were revised to a more restrictive value. Additionally,
    it was similar to example 4.a in IMC 0612, Appendix E, Examples of Minor Issues, which
    states that the issue of failing to appropriately evaluate scaffold installation as required by
    procedures is more than minor if the licensee routinely failed to perform engineering
    evaluations. The issue was evaluated in accordance with IMC 0609, Appendix A, The
    Significance Determination Process for Findings At-Power and determined to be of very low
    safety significance (Green), because it did not involve the loss or degradation of equipment
    or function specifically designed to mitigate a seismic event. This finding has a cross-cutting
    aspect in the area of Problem Identification and Resolution, Evaluation, because NextEra
    personnel did not perform an adequate extent of condition review after revision of their
    erection of scaffold procedure. This performance deficiency directly contributed to multiple
    instances of scaffold members erected within two inches of safety-related equipment without
    an engineering evaluation [P.2]. (Section 1R15)
                                                                                              Enclosure
 
                                                  4
                                          REPORT DETAILS
Summary of Plant Status
Seabrook operated essentially at full power for the entire assessment period, with the exception
of minor downpowers for turbine control valve testing. However, on March 31, 2014, plant load
was reduced to approximately 15% for turbine generator testing prior to a shutdown and entry
into refueling outage No. 16 at midnight, March 31, 2014. Documents reviewed for each section
of this inspection report are listed in the Attachment.
1.      REACTOR SAFETY
        Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01 - 1 sample)
        Readiness for Impending Adverse Weather Conditions
    a. Inspection Scope
        The inspectors reviewed NextEras preparations for the onset of cold weather and snow
        on February 5, 2014. The inspectors reviewed the implementation of adverse weather
        preparation procedures before the onset of and during this adverse weather
        condition. The inspectors verified that operator actions defined in NextEras adverse
        weather procedure maintained the readiness of essential systems. The inspectors
        discussed readiness and staff availability for adverse weather response with operations
        and work control personnel.
    b. Findings
        No findings were identified.
1R04 Equipment Alignment
        Partial System Walkdowns (71111.04Q - 3 samples)
    a. Inspection Scope
        The inspectors performed partial walkdowns of the following systems:
            'A' emergency diesel generator (EDG) while 'B' EDG was out of service (OOS) for
            annual maintenance on January 27, 2014
            Supplemental emergency power system (SEPS) while 'A' EDG was OOS for annual
            maintenance on February 12, 2014
            'A' emergency feedwater (EFW) pump return to service on March 26, 2014
        The inspectors selected these systems based on their risk-significance relative to the
        reactor safety cornerstones at the time they were inspected. The inspectors reviewed
        applicable operating procedures, system diagrams, the Updated Final Safety Analysis
                                                                                          Enclosure
 
                                                5
      Report (UFSAR), technical specifications (TSs), work orders (WOs), condition reports,
      and the impact of ongoing work activities on redundant trains of equipment in order to
      identify conditions that could have impacted system performance of their intended safety
      functions. The inspectors also performed field walkdowns of accessible portions of the
      systems to verify system components and support equipment were aligned correctly and
      were operable. The inspectors examined the material condition of the components and
      observed operating parameters of equipment to verify that there were no deficiencies.
      The inspectors also reviewed whether NextEra staff had properly identified equipment
      issues and entered them into the CAP for resolution with the appropriate significance
      characterization.
  b. Findings
      No findings were identified.
1R05 Fire Protection
      Resident Inspector Quarterly Walkdowns (71111.05Q - 5 samples)
  a. Inspection Scope
      The inspectors conducted tours of the areas listed below to assess the material
      condition and operational status of fire protection features. The inspectors verified
      that NextEra controlled combustible materials and ignition sources in accordance with
      administrative procedures. The inspectors verified that fire protection and suppression
      equipment was available for use as specified in the area pre-fire plan, and passive fire
      barriers were maintained in good material condition. The inspectors also verified that
      station personnel implemented compensatory measures for OOS, degraded, or
      inoperable fire protection equipment, as applicable, in accordance with procedures.
        Residual heat removal (RHR) containment spray safety injection (CSSI) equipment
          vault train 'A' RHR-F-1B-Z, RHR-F-2B-Z, RHR-3B-Z, RHR-F-4B-Z, RHR-F-4B-Z1,
          RHR-F-4B-Z2 on January 15, 2014
        RHR CSSI equipment vault train 'B' RHR-F-1A-Z, RHR-F-2A-Z, RHR-3A-Z,
          RHR-F-4A-Z, RHR-F-4A-Z1, RHR-F-4A-Z2 on January 16, 2014
        SEPS-F-1-0 on January 21, 2014
        Circulating water pump room SW-F-1A-Z on March 11, 2014
        Primary auxiliary building (PAB) piping penetration area PAB-F-1A-Z, PAB-F-1J-Z
          on March 17, 2014
  b. Findings
      No findings were identified.
                                                                                        Enclosure
 
                                                6
1R06 Flood Protection Measures (71111.06 - 1 sample)
      Internal Flooding Review
  a. Inspection Scope
      The inspectors reviewed the UFSAR, the site flooding analysis, and plant procedures to
      assess susceptibilities involving internal flooding. The inspectors also reviewed the CAP
      to determine if NextEra identified and corrected flooding problems and whether operator
      actions for coping with flooding were adequate. The inspectors also focused on the B
      EDG building to verify the adequacy of equipment seals located below the flood line,
      floor and water penetration seals, watertight door seals, common drain lines and sumps,
      sump pumps, level alarms, control circuits, and temporary or removable flood barriers.
  b. Findings
      No findings were identified.
1R11 Licensed Operator Requalification Program (71111.11 - 2 samples)
.1    Quarterly Review of Licensed Operator Requalification Testing and Training
  a. Inspection Scope
      The inspectors observed licensed operator simulator training on January 23, 2014,
      which included simulated degraded equipment and subsequent equipment failures
      and initiators, which resulted in escalating degraded plant conditions that ensured
      implementation of emergency operating procedures by the operating crew, as well as
      implementation of the emergency plan. This emergency plan implementation included
      classification of specific events that warranted an Alert Event Declaration. The
      inspectors evaluated operator performance during the simulated event and verified
      completion of risk significant operator actions, including the use of abnormal and
      emergency operating procedures. The inspectors assessed the clarity and effectiveness
      of communications, implementation of actions in response to alarms and degrading plant
      conditions, and the oversight and direction provided by the control room supervisor. The
      inspectors verified the accuracy and timeliness of the emergency classification made by
      the shift manager and the TS action statements entered by the control room supervisor.
      Additionally, the inspectors assessed the ability of the crew and training staff to identify
      and document crew performance problems.
  b. Findings
      No findings were identified.
.2    Quarterly Review of Licensed Operator Performance in the Main Control Room
  a. Inspection Scope
      The inspectors observed general control room activities, including alarm response and
      control room shift turnovers, conducted on January 13, 2014, March 14, 2014 and
                                                                                          Enclosure
 
                                              7
      March 27, 2014. Additionally the inspectors observed turbine control valve testing on
      January 17, 2014, engineered safety features actuation system (ESFAS) relay testing
      on January 27, 2014, operator response to a failed open B steam generator feed
      regulating bypass valve on March 14, 2014, and restoration from enclosure building
      exhaust fan EAH-FN-4A testing on March 19, 2014. The inspectors observed test
      performance to verify that procedure use, crew communications, and coordination of
      activities between work groups similarly met established expectations and standards.
  b. Findings
      No findings were identified.
1R12 Maintenance Effectiveness (71111.12 - 2 samples)
  a. Inspection Scope
      The inspectors reviewed the samples listed below to assess the effectiveness of
      maintenance activities on structure, system, or component (SSC) performance and
      reliability. The inspectors reviewed system health reports, CAP documents,
      maintenance WOs, and maintenance rule (MR) basis documents to ensure that
      NextEra was identifying and properly evaluating performance problems within the scope
      of the MR. For each sample selected, the inspectors verified that the SSC was properly
      scoped into the MR in accordance with 10 CFR 50.65 and verified that the (a)(2)
      performance criteria established by NextEra staff was reasonable. As applicable, for
      SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective
      actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that NextEra
      staff was identifying and addressing common cause failures that occurred within and
      across MR system boundaries.
          SW pump P-41C increased vibration trending in January 2014
          ED/EDE 120 VAC electrical distribution systems on February 18, 2014
  b. Findings
      No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 5 samples)
  a. Inspection Scope
      The inspectors reviewed station evaluation and management of plant risk for the
      maintenance and emergent work activities listed below to verify that NextEra performed
      the appropriate risk assessments prior to removing equipment for work. The inspectors
      selected these activities based on potential risk significance relative to the reactor safety
      cornerstones. As applicable for each activity, the inspectors verified that NextEra
      personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the
      assessments were accurate and complete. When NextEra performed emergent work,
      the inspectors verified that operations personnel promptly assessed and managed plant
      risk. The inspectors reviewed the scope of maintenance work and discussed the results
      of the assessment with the stations probabilistic risk analyst to verify plant conditions
                                                                                          Enclosure
 
                                              8
      were consistent with the risk assessment. The inspectors also reviewed the TS
      requirements and inspected portions of redundant safety systems, when applicable,
      to verify risk analysis assumptions were valid and applicable requirements were met.
          RHR system valve maintenance and testing on January 14, 2014
          ESFAS relay testing on January 15, 2014
          Inverter 1B corrective maintenance following internal transformer failure on
          February 19, 2014
          Planned SW cooling tower switchover on March 20, 2014
          'B' feedwater regulating bypass valve M/A station replacement on March 27, 2014
  b. Findings
      No findings were identified.
1R15 Operability Determinations and Functionality Assessments (71111.15 - 7 samples)
  a. Inspection Scope
      The inspectors reviewed operability determinations for the following degraded or non-
      conforming conditions:
        East and west pipe chase low temperature impact on feedwater isolation valve
          operability on January 3, 2014
        Operability of safety-related equipment in close proximity to temporary scaffolding on
          January 16, 2014
        Service water pumphouse seismic monitor non-functional following monthly testing
          on January 24, 2014
        Containment enclosure ventilation area seal gaps identified on January 28, 2014
        'A' EDG did not trip on overspeed during return to service testing on February 14, 2014
        Reactor coolant system (RCS) leakage into the RHR system on March 13, 2014
        Turbine-driven EFW pump P-37B oil leak identified during testing on March 19, 2014
      The inspectors selected these issues based on the risk significance of the associated
      components and systems. The inspectors evaluated the technical adequacy of the
      operability determinations to assess whether TS operability was properly justified and
      the subject component or system remained available such that no unrecognized
      increase in risk occurred. The inspectors compared the operability and design criteria in
      the appropriate sections of the TSs and UFSAR to NextEras evaluations to determine
      whether the components or systems were operable. Where compensatory measures
      were required to maintain operability, the inspectors determined whether the measures
      in place would function as intended and were properly controlled by NextEra. The
      inspectors determined, where appropriate, compliance with bounding limitations
      associated with the evaluations.
  b. Findings
      Introduction. The inspectors identified a Green NCV of 10 CFR 50, Appendix B,
      Criterion V, Procedures, because NextEra did not ensure adequate separation was
      maintained between temporary scaffolding and safety-related equipment. Specifically,
                                                                                        Enclosure
 
                                          9
six instances of scaffolding installed in the plant were identified with less than the
minimum standoff distance to safety-related equipment specified in NextEra procedures
and no corresponding engineering evaluation to support these deviations.
Description. 10 CFR 50, Appendix B, Criterion V, requires that activities affecting quality
be prescribed by documented procedures and be accomplished in accordance with
those procedures. When used in the plant, the design and installation of temporary
scaffolding must be controlled to ensure that it is not installed too close to safety-related
equipment. During a seismic event, scaffolding installed too close to safety-related
equipment can come into contact with that equipment, cause damage to it, and affect its
safety function. NextEra procedures control the installation of temporary scaffolding at
Seabrook by specifying a minimum separation between scaffolding and safety-related
equipment, and by requiring an engineering evaluation when the minimum separation
cannot be met.
NextEra mechanical maintenance procedure, MS0599.47, Erection of Scaffolding,
Revision 2, states that members of scaffolding erected adjacent to operable safety-
related equipment shall not be less than two inches unless justified by an Engineering
Evaluation. MS0599.47 was revised in February 2013, and the requirement for scaffold
separation from operable safety-related equipment was changed from ...should not be
less than 2 inches and in no case less than 1/2 inch without an engineering evaluation to
...shall not be less than 2 inches unless justified by engineering evaluation.
While performing a plant walkdown on January 15, 2014, the inspectors identified
temporary scaffold members installed less than two inches from the A Containment
Building Spray (CBS) pump discharge and suction lines. The A CBS pump and its
associated piping are classified as safety-related equipment and were operable at the
time. The identified scaffold did not include an engineering evaluation that provided
acceptance of separation of less than two inches. Subsequent plant walkdowns by
NextEra personnel identified five additional instances of scaffolding installed less than
two inches from operable safety-related equipment without an associated engineering
evaluation. NextEra personnel determined that an inadequate extent of condition review
following the February 2013 revision of MS0599.47, resulted in scaffolding being staged
in the plant at less than the new, more restrictive scaffold separation requirement of two
inches. Having identified multiple instances where NextEra personnel had not complied
with the separation requirement of the scaffolding procedure, the inspectors concluded
that NextEra had not been adequately controlling the design and installation of
temporary scaffolding.
NextEra entered the additional instances of inadequate separation identified during their
independent walkdowns into the CAP. All discrepancies were corrected and assessed
for any potential impact to the operability or functionality of affected systems. The
inspectors reviewed the CRs and determined that the safety function of each system
potentially impacted by temporary scaffolding, including those identified by the
inspectors and NextEra, would not have been affected during a seismic event.
Analysis. The inspectors determined that not providing adequate separation between
temporary scaffolding and safety-related equipment without an engineering basis was
a performance deficiency within NextEras ability to foresee and correct. Specifically,
several scaffold members were observed within two inches of safety-related equipment
without an engineering evaluation as specified by current procedural requirements.
                                                                                    Enclosure
 
                                                10
      This performance deficiency was considered more than minor because it affected the
      protection against external factors attribute of the Mitigating System cornerstone and its
      objective to ensure the availability, reliability, and capability of systems that respond to
      initiating events to prevent undesirable consequences. Specifically, NextEra did not
      evaluate scaffolding installation when insufficient separation to safety-related equipment
      existed after procedural requirements were revised to a more restrictive value.
      Additionally, it was similar to example 4.a in IMC 0612, Appendix E, Examples of Minor
      Issues, which states that the issue of failing to appropriately evaluate scaffold
      installation as required by procedures is more than minor if the licensee routinely failed
      to perform engineering evaluations. The issue was evaluated in accordance with IMC
      0609, Appendix A, The Significance Determination Process for Findings At-Power and
      determined to be of very low safety significance (Green) since it did not involve the loss
      or degradation of equipment or function specifically designed to mitigate a seismic event.
      This finding is related to the cross-cutting area of Problem Identification and Resolution-
      Evaluation, because NextEra did not thoroughly evaluate issues to ensure that
      resolutions address causes and extent of conditions commensurate with their safety
      significance (P.2). Specifically, NextEra personnel did not perform an adequate extent
      of condition review after revision of their erection of scaffolding procedure. This
      performance deficiency directly contributed to multiple instances of scaffolding members
      erected within two inches of safety-related equipment without an engineering evaluation.
      Enforcement. 10 CFR 50, Appendix B, Criterion V, requires, in part, that activities
      affecting quality shall be prescribed by documented procedures and shall be
      accomplished in accordance with those procedures. NextEra mechanical maintenance
      procedure, MS0599.47, Erection of Scaffolding, Revision 2, states that members of
      scaffolding erected adjacent to operable safety-related equipment shall not be less than
      two inches from the equipment unless justified by an Engineering Evaluation. Contrary
      to the above, on January 15, 2014, the inspectors identified that certain activities
      affecting quality at Seabrook were not accomplished in accordance with documented
      procedures. Specifically, following a revision of the minimum scaffolding separation
      requirement in February 2013, multiple instances of scaffolding outside of the new
      requirements were left uncorrected and engineering evaluations were not completed.
      Installation of temporary scaffolding in the vicinity of safety-related equipment has the
      potential to adversely affect that equipments performance during a seismic event
      because it was installed with insufficient standoff distance. After the issue was identified
      by the inspectors, NextEra performed independent walkdowns of all scaffolding,
      identifying five additional instances of inadequate separation distance. All identified
      discrepancies were corrected or evaluated as adequate. Because this violation is of
      very low safety significance (Green) and NextEra entered this into their CAP (AR
      01933827), this violation is being treated as an NCV consistent with the Enforcement
      Policy. (NCV 05000443/2014002-01, Scaffolding Installed with Insufficient
      Separation to Safety Related Equipment)
1R18 Plant Modifications (71111.18 - 1 sample)
      Permanent Modifications
  a. Inspection Scope
      The inspectors evaluated a modification that replaced the SW flow element to the A
      EDG jacket water cooler implemented under engineering change EC280824, and
                                                                                            Enclosure
 
                                              11
      completed on February 13, 2014. The inspectors verified that the design and licensing
      bases, as well as the performance capability of the affected SW train and associated
      components were not degraded by the modification. The inspectors reviewed
      associated modification documents, which included topic notes, implementation work
      order instructions, equivalent change revisions, applicable interface documents (for
      example, drawings), and applicable post-modification testing.
  b. Findings
      No findings were identified.
1R19 Post-Maintenance Testing (71111.19 - 6 samples)
  a. Inspection Scope
      The inspectors reviewed the post-maintenance tests for the maintenance activities listed
      below to verify that procedures and test activities ensured system operability and
      functional capability. The inspectors reviewed the test procedure to verify that the
      procedure adequately tested the safety functions that may have been affected by the
      maintenance activity, that the acceptance criteria in the procedure was consistent with
      the information in the applicable licensing basis and/or design basis documents, and that
      the procedure had been properly reviewed and approved. The inspectors also
      witnessed the test or reviewed test data to verify that the test results adequately
      demonstrated restoration of the affected safety functions.
          Thermal barrier cooling pump monthly surveillance following electrical breaker
          testing on January 11, 2014
          SEPS diesel generator DG-2A maintenance on January 21, 2014
          Portable diesel driven pump B.5.b functional test following repairs on February 3, 2014
          EDG A exhaust valve Belleville washer replacement on February 11, 2014
          1B vital inverter Ferro-Resonant transformer replacement on February 19, 2014
          Service water pump P-41C following shaft sleeve replacement on March 6, 2014
  b. Findings
      No findings were identified.
1R22 Surveillance Testing (71111.22 - 7 samples)
  a. Inspection Scope
      The inspectors observed performance of surveillance tests and/or reviewed test data of
      selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR,
      and NextEra procedure requirements. The inspectors verified that test acceptance
      criteria were clear, tests demonstrated operational readiness and were consistent with
      design documentation, test instrumentation had current calibrations and the range and
      accuracy for the application, tests were performed as written, and applicable test
      prerequisites were satisfied. Upon test completion, the inspectors considered whether
      the test results supported that equipment was capable of performing the required safety
      functions. The inspectors reviewed the following surveillance tests:
                                                                                        Enclosure
 
                                              12
        Rod control testing and verification of proper operation of digital rod position
          indication on January 9, 2014
        Portable diesel driven pump B.5.b annual functional test on January 14, 2014
        'A' cooling tower pump comprehensive inservice test on January 16, 2014 (IST)
        RCS leak rate surveillance test on March 5, 2014 (RCS leak rate)
        RCS pump seal monthly controlled leakage surveillance on March 12, 2014
        Containment enclosure exhaust fan EAH-FN-4A monthly testing on March 19, 2014
        'B' EDG operability surveillance and ESFAS slave relay testing March 25, 2014
  b. Findings
      No findings were identified.
      Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04 - 1 sample)
      Emergency Preparedness Drill Observation
  a. Inspection Scope
      NextEra implemented various changes to the Seabrook Emergency Action Levels
      (EALs), Emergency Plan, and Implementing Procedures. NextEra had determined that,
      in accordance with 10 CFR 50.54(q)(3), any change made to the EALs, Emergency
      Plan, and its lower-tier implementing procedures, had not resulted in any reduction in
      effectiveness of the Plan, and that the revised Plan continued to meet the standards in
      50.47(b) and the requirements of 10 CFR 50, Appendix E.
      The inspectors performed an in-office review of all EAL and Emergency Plan changes
      submitted by NextEra as required by 10 CFR 50.54(q)(5), including the changes to
      lower-tier emergency plan implementing procedures, to evaluate for any potential
      reductions in effectiveness of the Emergency Plan. This review by the inspectors was
      not documented in an NRC Safety Evaluation Report and does not constitute formal
      NRC approval of the changes. Therefore, these changes remain subject to future NRC
      inspection in their entirety. The requirements in 10 CFR 50.54(q) were used as
      reference criteria.
  b. Findings
      No findings were identified.
1EP6 Drill Evaluation (71114.06 - 2 samples)
.1    Emergency Preparedness Drill Observation
  a. Inspection Scope
      The inspectors evaluated a routine NextEra emergency drill on March 12, 2014, to
      identify any weaknesses and deficiencies in the event classification and notification
                                                                                          Enclosure
 
                                                  13
      activities. The inspectors observed emergency response operations in the simulator and
      technical support center to determine whether the event classification and notification
      activities were performed in accordance with procedures. The inspectors also attended
      the individual facility drill critique to compare inspector observations with those identified
      by NextEra staff, to evaluate NextEras critique and to verify whether NextEra staff was
      properly identifying weaknesses and entering them into the CAP.
  b. Findings
      No findings were identified.
.2    Emergency Preparedness Training Observations
  a. Inspection Scope
      The inspectors observed a simulator training evolution for Unit 1 licensed operators on
      January 23, 2014, which involved simulated emergency plan implementation by an
      operations crew. NextEra planned for this evolution to be evaluated and included in
      performance indicator data regarding drill and exercise performance. The inspectors
      observed event classification and notification activities performed by the crew. The
      inspectors also attended the post-evolution critique for the scenario. The focus of the
      inspectors activities was to note any weaknesses and deficiencies in the crews
      performance and ensure that NextEra evaluators noted the same issues and entered
      them into the CAP.
  b. Findings
      No findings were identified.
2.    RADIATION SAFETY
      Cornerstone: Public Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)
  a. Inspection Scope
      From March 4 to March 8, 2014, the inspectors reviewed NextEras performance in
      assessing the radiological hazards and exposure control in the workplace. The
      inspectors used the requirements in 10 CFR Part 20 and guidance in Regulatory Guide
      (RG) 8.38, Control of Access to High and Very High Radiation Areas for Nuclear Plants,
      TSs, and the NextEra procedures required by TSs, as criteria for determining
      compliance.
      Inspection Planning
      The inspectors reviewed 2013 NextEra performance indicators for the occupational
      exposure cornerstone for Seabrook Nuclear Station. The inspectors reviewed the
      results of radiation protection program audits. The inspectors reviewed any reports of
      operational occurrences related to occupational radiation safety since the last inspection.
                                                                                          Enclosure
 
                                          14
Radiological Hazard Assessment
The inspectors determined if there have been changes to plant operations since the last
inspection that may result in significant new radiological hazards. The inspectors
evaluated whether NextEra assessed the potential impact of these changes and has
implemented periodic monitoring, as appropriate, to detect and quantify the radiological
hazard.
The inspectors reviewed the last two radiological surveys from the Fuel Transfer Canal
and Letdown Line in the Demineralizer Alley. The inspectors evaluated whether the
thoroughness and frequency of the surveys were appropriate for the given new
radiological hazard.
The inspectors conducted walk-downs and independent radiation measurements in the
facility, including radioactive waste processing, storage, and handling areas to evaluate
material and radiological conditions.
The inspectors reviewed one risk-significant work activity that involved exposure to
radiation. This activity was the initial entry, survey and decontamination of the fuel
transfer canal following the second dry spent fuel storage cask loading campaign. For
this work activity, the inspectors assessed whether the pre-work surveys performed were
appropriate to identify and quantify the radiological hazard and to establish adequate
protective measures. The inspectors evaluated the radiological survey program to
determine if radiological hazards were properly identified (e.g., discrete radioactive hot
particles, transuranics and hard to detect nuclides in air samples, transient dose rates
and large gradients in radiation dose rates).
The inspectors observed work in potential airborne radioactivity areas, and evaluated
whether the air samples from the fuel transfer canal air sample locations were
representative of the breathing air zone and were properly evaluated.
Instructions to Workers
The inspectors selected three containers of radioactive materials and assessed whether
the containers were labeled and controlled in accordance with 10 CFR Part 20
requirements.
The inspectors reviewed the following radiation work permits (RWP) used to access high
radiation areas (HRA) and evaluated if the specified work control instructions and control
barriers were consistent with TS requirements for HRA.
  RWP 14-0015 High Integrity Container/Liner Shipping Preparation to include
    Capping, Weighing and Transfer to Waste Processing Building, January 1, 2014
  RWP 14-0022 Inspect CS Valves inside Letdown Valve Room at Power,
    January 15, 2014
  RWP 14-0027 Primary Auxiliary Building Demineralizer Alley Work/Entry,
    December 31, 2013
  RWP 14-0058 Fuel Storage Building Transfer Canal Radiation Protection Survey,
    Decontamination and Maintenance Support Activities, March 2, 2014
                                                                                  Enclosure
 
                                          15
For these RWPs, the inspectors assessed whether allowable stay-times or permissible
dose for radiologically significant work under each RWP were clearly identified. The
inspectors evaluated whether electronic personnel dosimeter (EPD) alarm set-points
were in conformance with survey indications and plant procedural requirements.
The inspectors reviewed two occurrences where a workers EPD malfunctioned or
alarmed. The inspectors evaluated whether workers responded appropriately. The
inspectors assessed whether the issue was included in the corrective action program
and whether compensatory dose evaluations were conducted as appropriate.
For work activities that could suddenly increase radiological conditions, the inspectors
assessed the NextEra means to inform workers of these changes.
Contamination and Radioactive Material Control
The inspectors reviewed NextEras criteria for the survey and release of potentially
contaminated material. The inspectors evaluated whether there was sufficient
procedural guidance on alarm response.
The inspectors reviewed NextEras procedures and records to verify that the radiation
detection instrumentation was used at an appropriate sensitivity level. The inspectors
selected six sealed sources from the NextEra inventory records and reviewed whether
the sources were accounted for and were tested for loose surface contamination.
The inspectors evaluated whether any recent transactions involving nationally tracked
sources were reported in accordance with 10 CFR Part 20 requirements.
Radiological Hazards Control and Work Coverage
The inspectors evaluated ambient radiological conditions and performed independent
radiation measurements during walk-downs of the facility. The inspectors assessed
whether the conditions were consistent with applicable posted surveys, RWPs, and
associated worker briefings.
The inspectors examined NextEra physical and programmatic controls for highly
activated or contaminated materials stored within spent fuel and other storage pools.
The inspectors assessed whether appropriate controls were in place to preclude
inadvertent removal of these materials from the pool.
The inspectors examined the posting and physical controls for selected HRAs, locked
high radiation area (LHRA) and very high radiation areas (VHRA) to verify conformance
with the occupational performance indicator.
Risk-Significant HRA and VHRA Controls
The inspectors discussed with the Radiation Protection Manager (RPM) the controls and
procedures for high-risk HRAs and VHRAs. The inspectors assessed whether any
changes to NextEra relevant procedures reduce the effectiveness of worker protection.
                                                                                  Enclosure
 
                                              16
      The inspectors evaluated NextEra controls for VHRAs and areas with the potential to
      become a VHRA to ensure that an individual was not able to gain unauthorized access
      to these areas.
      Problem Identification and Resolution
      The inspectors evaluated whether problems associated with radiation monitoring and
      exposure control were being identified by NextEra at an appropriate threshold and were
      properly addressed for resolution in the licensees corrective action program. The
      inspectors assessed the appropriateness of the corrective actions for problems
      documented by NextEra that involve radiation monitoring and exposure controls. The
      inspectors assessed NextEra processes for applying operating experience to their plant.
  b. Findings
      No findings were identified.
2RS2 Occupational ALARA Planning and Controls (71124.02)
  a. Inspection Scope
      The inspectors assessed performance with respect to maintaining occupational
      individual and collective radiation exposures as low as is reasonably achievable
      (ALARA). The inspectors used the requirements in 10 CFR Part 20, RG 8.8,
      Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear
      Power Plants will be As Low As Is Reasonably Achievable, RG 8.10, Operating
      Philosophy for Maintaining Occupational Radiation Exposure As Low as Is Reasonably
      Achievable, TSs, and NextEra procedures required by TSs, as criteria for determining
      compliance.
      Inspection Planning
      The inspectors reviewed information regarding Seabrooks collective dose history,
      current exposure trends, and ongoing or planned activities in order to assess current
      performance and exposure challenges. The inspectors reviewed the plants three year
      rolling average collective radiation exposure.
      The inspectors compared the site-specific trends in collective exposures against the
      industry average values and from similar nuclear power plants. In addition, the
      inspectors reviewed any changes in the radioactive source term by reviewing the trend
      in average contact dose rate with reactor coolant piping and steam generator primary
      channel head space and manways. The inspectors reviewed site-specific procedures
      associated with maintaining occupational exposures ALARA, which included a review of
      processes used to estimate and track exposures from specific work activities.
      Radiological Work Planning
      The inspectors assessed whether NextEra planning identified appropriate dose
      reduction techniques; considered alternate dose reduction features; and estimated
      reasonable dose goals. The inspectors evaluated whether NextEras ALARA
      assessment had taken into account decreased worker efficiency from use of respiratory
      protective devices and/or heat stress mitigation equipment. The inspectors determined
                                                                                      Enclosure
 
                                                17
      whether NextEra work planning considered the use of remote technologies as a means
      to reduce dose and the use of dose reduction insights from industry operating
      experience and plant-specific lessons learned. The inspectors assessed the integration
      of ALARA requirements into work procedure and RWP documents.
      Source Term Reduction and Control
      The inspectors used licensee records to determine the historical trends and current
      status of plant source term known to contribute to elevated facility collective dose. The
      inspectors assessed whether the licensee had developed contingency plans for
      expected changes in the source term as the result of changes in plant fuel performance
      issues or changes in plant primary chemistry.
      Problem Identification and Resolution
      The inspectors evaluated whether problems associated with ALARA planning and
      controls are being identified by the licensee at an appropriate threshold and were
      properly addressed for resolution in the licensees corrective action program. The
      inspectors assessed NextEras process for applying operating experience to their plant.
  b. Findings
      No findings were identified.
4.    OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
      Unplanned Scrams, Unplanned Power Changes, and Unplanned Scrams with
      Complications (3 samples)
  a. Inspection Scope
      The inspectors reviewed NextEras submittals for the following Initiating Events
      Cornerstone performance indicators for the period of January 1, 2013 through
      December 31, 2013.
        Unplanned scrams per 7000 critical hours
        Unplanned scrams with complications
        Unplanned power changes per 7000 critical hours
      To determine the accuracy of the performance indicator data reported during those
      periods, inspectors used definitions and guidance contained in Nuclear Energy Institute
      (NEI) Document 99-02, Regulatory Assessment Performance Indicator Guideline,
      Revision 6. The inspectors reviewed NextEras operator narrative logs, event reports,
      and NRC integrated inspection reports to validate the accuracy of the submittals.
  b. Findings
      No findings were identified.
                                                                                        Enclosure
 
                                                18
4OA2 Problem Identification and Resolution (71152 - 2 samples)
.1    Routine Review of Problem Identification and Resolution Activities
  a. Inspection Scope
      As required by Inspection Procedure 71152, Problem Identification and Resolution, the
      inspectors routinely reviewed issues during baseline inspection activities and plant
      status reviews to verify that NextEra entered issues into the CAP at an appropriate
      threshold, gave adequate attention to timely corrective actions, and identified and
      addressed adverse trends. In order to assist with the identification of repetitive
      equipment failures and specific human performance issues for follow-up, the inspectors
      performed a daily screening of items entered into the CAP and periodically attended
      condition report screening meetings.
  b. Findings
      No findings were identified.
.2    Annual Sample: Increasing Frequency of Leaks in Service Water Piping in the Vicinity of
      Installation/Fabrication Welds
  a. Inspection Scope
      During the period January 27 to January 31, 2014, inspectors reviewed a root cause
      evaluation (RCE AR 16379222) completed by NextEra staff for a service water pipe leak
      that occurred in August 2013. This problem was described in a licensee event report
      submitted to the NRC dated December 23, 2013. The inspectors determined the
      effectiveness of actions by NextEra staff to identify, characterize, correct and prevent
      reoccurrence of SW system leaks.
      The inspectors assessed problem identification threshold, apparent cause analysis,
      extent of condition reviews, and timeliness of corrective actions. The inspectors
      reviewed documents listed in the Attachment to this report and interviewed NextEra
      engineering personnel to assess the effectiveness of the planned, scheduled, and
      completed corrective actions to resolve the identified deficiency.
      The inspectors reviewed non-destructive test procedures, procedure qualifications
      including test personnel qualifications to determine compliance with the applicable
      American Society of Mechanical Engineers codes and standards. Also, the inspectors
      reviewed system health reports, work orders, procurement documents, drawings and
      photographs to determine if the nonconforming condition was appropriately identified,
      documented, characterized and entered into NextEras corrective action process.
      The inspectors reviewed root cause evaluation AR 16379222 and interviewed members
      of the evaluation team. The inspectors interviewed the qualified non-destructive test
      examiner to evaluate the ultrasonic test method used. Test results were reviewed with
      the test examiner to assess the remaining wall thickness for continued operation without
      encroaching on minimum wall requirements.
                                                                                        Enclosure
 
                                                  19
  b. Findings and Observations
      No findings were identified. The root cause evaluation and corrective actions were
      reasonable, appropriate and timely.
      NextEras root cause evaluation addressed a history of SW degradation (corrosion/
      erosion) resulting in wall thinning and pressure boundary penetration and leakage. The
      areas where wall thinning and leakage occurred was determined to be associated with
      the loss of protective coating and/or liner failure at fabrication/installation welds which
      typically results in turbulent fluid flow. This turbulent flow was particularly aggressive in
      the attack of base metals and protective coatings at these weld locations and
      configuration changes. The inspectors assessed the root cause determination, results of
      the extent of condition investigation of other locations within the SW system and other
      fluid (circulating water) systems with similar piping materials, operating parameters and
      configurations.
      The inspectors noted that examination using ultrasonic testing was performed at
      selected locations with known change in flow patterns and velocity changes. The results
      of this testing identified areas exhibiting variable wear rates. An evaluation of these test
      results was made to determine pipe structural and pressure retaining integrity.
      The inspectors visually examined several portions of previous SW pipe and fittings that
      had been removed from the SW system in prior outages due to identified leaks. The
      removed samples provided confirmatory evidence of corrosive/erosive attack from
      turbulent flow at root locations of field welds, configuration changes and pipe to fitting
      intersections. These locations revealed characteristic pin hole leaks at weld locations
      and a general wastage of pipe and fitting interior diameters. These locations were
      evaluated for compliance with minimum wall thickness requirements. Those locations
      which were identified as active leaks at weld locations or, where areas exhibiting loss
      of wall thickness and were encroaching on minimum wall requirements, were
      dispositioned for repair/replacement in the CAP.
      The inspectors determined that this issue received appropriate management attention as
      indicated by the corrective action that was taken to perform a temporary leak repair by
      the installation of a weldolet encapsulating the leak location. At the next outage, (OR16)
      the weldolet will be removed and replaced with a more suitable flush patch. The patch
      will be coated internally with a corrosion/erosion resistant material. The inspectors
      discussed the licensee plans to systematically remove and replace the SW piping with a
      base metal that is significantly more resistant to erosion/corrosion attack. The
      inspectors examined numerous lengths of pipe and fittings which were staged for
      replacement in the plant during the April 2014 refueling outage and subsequent outages.
.3    Annual Sample: Review of Activities Associated with Alkali Silica Reaction Affected
      Structures
  a. Inspection Scope
      March 12 to 13, NRC inspectors from Region I and a structural engineer from the
      Division of License Renewal, NRR, witnessed testing conducted at the Ferguson
      Structural Engineering Laboratory (FSEL) at the University of Texas - Austin. The
      testing was conducted in support of the Seabrook Alkali-Silica Reaction (ASR)
                                                                                            Enclosure
 
                                            20
  Integrated Corrective Action Plan. Specifically, the inspectors witnessed load testing of
  the control beam for reinforcement anchorage (lap-splice) capacity. The testing was
  performed in accordance with MPR Project 0326-0063, Procedure 5-7, Structural
  Testing of Shear and Anchorage Specimens, Revision 1.
  The inspectors also reviewed the results of the December 2013, Combined Crack
  Indexing (CCI) measurements and the supporting engineering analysis. Proprietary data
  sheets and associated evaluations were made available for inspector review.
  Additionally, the inspectors review included discussions with the responsible Seabrook
  engineers, as well as petrography specialists consulting for the University of Texas
  Ferguson Structural Engineering Laboratory
  Lastly, the inspectors reviewed NextEras revised Prompt Operability Determinations
  (PODs) that address additional Seabrook structures identified as being affected by ASR
  via the Phase 3 ASR walkdown program.
b. Findings and Observations
  The inspectors identified no findings.
  Review of CCI and Crack Width Measurements
  The inspectors examined the December 2013 CCI measurement results documented in
  Foreign Print (FP) 100847 and FP 100848, dated January 30, 2014. As documented in
  these NextEra reports, there are 32 areas currently being monitored for ASR progression
  using the CCI and crack width methodology. As of December 2013, 26 of the 32 areas
  have been monitored on a six-month basis for approximately two years. Based upon the
  data collected to date, NextEra has concluded the following: 1) the data suggests a slow
  increasing trend in CCI and crack width over the past two years; 2) at 14 interior ASR
  locations, the horizontal and vertical CCI data indicates an overall upward trend (an
  average increase of 0.04 mm/m, with a measurement tolerance of 0.05 mm); 3) at nine
  exterior locations, the horizontal and vertical CCI data indicates no significant change
  over the two-year period; 4) the six floor/ceiling/roof locations indicate flat to upward
  trends early in the period, but no change later in the two-year period; and, 5) some
  fluctuation in the measured CCI and crack width values have been observed. The
  fluctuations may be attributed to thermal effects, cyclic or constant moisture exposure,
  measurement device accuracy, or the condition of the measured surfaces as impacted
  by cleaning/preparation and weathering effects. Independent inspector review of CCI
  and crack width measurement data and photographs of selected areas confirmed
  NextEras conclusions. As stated in FP 100847 and FP 100848, NextEra will continue
  the six-month data collection to comply with the Structures Monitoring Program and
  validate these observed ASR progression trends.
  Review of Operability Determinations
  Based on the result of recent Phase 3 walkdowns, NextEra identified six additional areas
  with CCI > 1.0 millimeter per meter (mm/m). In accordance with the Seabrook
  Structures Monitoring Program (SMP), NextEra staff completed evaluations of the
  affected structures to assess the potential impact of ASR on continued operability. The
  six additional areas identified with a CCI value greater than 1.0 mm/m were: cooling
  tower exterior (elevation 25, reference CTE-02); primary auxiliary building (PAB)
                                                                                        Enclosure
 
                                          21
penetration area (elevation -26, reference MF105-01); west pipe chase (elevation 12,
reference MF202-02); and three areas in electrical manholes (below grade elevations,
reference CI-W03-Wall, CI-W05-Wall, and CI-W11-Wall). The structural evaluations
were documented as Supplement IV and V to FP 100716, Seabrook Station: Impact of
ASR on Concrete Structures and Attachments, and utilized the same design capacity
versus calculated demand margin analysis approach as the previously completed PODs
(reference Section 9.1, NRC Inspection Report 05000443/2012010). Inspector review of
the six additional ASR-affected area PODs concluded that the impacted structures have
adequate strength margin available and are fully capable of performing their safety
functions.
Control Beam Testing Observations
The inspectors witnessed the performance of load testing of the first control beam (a
specimen that has not undergone ASR aging). The beam (A-7, reinforcement
anchorage control specimen) was tested in accordance with MPR Project 0326-0063,
Procedure 5-7, Structural Testing of Shear and Anchorage Specimens, Revision 1, on
March 13, 2014. Beam failure occurred at a load of approximately 251,000 pounds, as
compared to the estimated failure load of 214,000 pounds. The failure load was slightly
higher than the estimated design capacity, but within the accuracy of the design
calculations. No test anomalies were identified. The results of the control beam test will
be used to compare subsequent ASR-affected specimen tests to evaluate the impact of
ASR degradation on structural performance. The inspectors observed proper procedural
adherence, good test coordination and proper communications and safety practices
exhibited by the testing staff, supervisory personnel and quality assurance overseers.
The inspectors verified proper testing preparations and quality control oversight as
specified by MPR Project 0326-0063, Procedure 5-6, General Preparation of Test
Facilities and Specimens, Revision 2, and MPR Procedure 0326-0062-46, Procedure
for In-process Inspections of FSEL Reinforcement Anchorage Test Setup for Seabrook
Station, Revision 0.
Initial Test Specimen ASR Expansion Results
NextEra and the UT-Austin FSEL staff have observed in the large-scale test specimens
that the X- and Y-direction deep pin expansion measurements (comparable to the
Seabrook vertical and horizontal wall surface CCI measurements) do not appear to
correlate with the through-wall (e.g., out-of-plane, or Z-direction) deep pin expansion
measurements after the initial phase of ASR expansion. X- and Y-direction expansion
appears to plateau while the Z-direction expansion continues to trend upward (increase).
All large-scale reinforcement anchorage and shear specimens have demonstrated this
expansion trend. The Z-direction expansion in the test specimens has been observed to
be 10 times greater than the X- and Y- expansions after approximately one year.
The preliminary implication of these test specimen expansion measurement trends is
that the X- and Y- expansion measurement methods (CCI and crack width) currently
used for monitoring the progression of ASR on Seabrook Station structure surfaces (per
the Structures Monitoring Program) may not provide alone, an adequate means to
monitor (1) ASR progression and (2) by inference (pending the completion of the testing
program), the ASR impact on the affected buildings structural performance. The
validation of the use of the CCI and crack width measurements for monitoring the
structural impact of ASR has been an objective of the large specimen testing program.
                                                                                  Enclosure
 
                                              22
    In considering these initial test program results, NextEra staff initiated an Action Report
    (No. 01952162) to address this issue. In addition to evaluating the future impact of using
    CCI to monitor ASR progression on affected structures, NextEra staff conducted a
    preliminary assessment of this test data for impact to their PODs completed for ASR-
    affected Seabrook structures. NextEra staff concluded this initial test program data does
    not adversely impact the POD margins analyses, principally because the PODs are not
    dependent upon measured CCI values for assessing the level of ASR degradation.
    Instead, those structures demonstrating the most significant ASR progression (as
    assumed by CCI and crack width measurements of >1.0 mm/m and >1.0 mm,
    respectively) were evaluated using conservative and bounding degradation values
    derived from published research and testing data developed from non-reinforced
    concrete specimens. At the close of the inspection period, NextEra staff had initiated
    actions to re-evaluate the use of CCI and crack width for the SMP and to re-evaluate
    their methods for monitoring and assessing ASR progression of test specimens in order
    to correlate test data to Seabrook Station.
    NextEra staff communicated with the inspectors their plans to fabricate an additional
    large-scale test specimen to instrument with strain gages and allow ASR progression in
    order to validate the use of strain gages for through-wall (Z-direction) expansion
    monitoring. The purpose of this activity is to validate the use of one or more strain gage
    designs that can subsequently be installed in Seabrook structures to accurately monitor
    through-wall expansion. In addition, NextEra staff described plans to increase the core
    sampling of control and ASR-affected large scale test specimens in order to more
    accurately measure ASR impact on concrete compressive and tensile strength and
    modulus of elasticity. NextEra staff further communicated plans to conduct petrographic
    examination of through-wall core samples from the test specimens. This additional
    concrete material property testing and data collection is intended to be used to support
    the correlation of testing program structural performance data to Seabrook structures
    (along with additional, but not yet defined, core sampling of Seabrook ASR-affected
    structures).
    In summary, the inspectors concluded that the PODs completed for ASR-affected
    Seabrook structures remain unaffected by the X-,Y- and Z-direction expansion data
    measured, to date, in the test specimens. Actions planned by NextEra to assess the
    adequacy of the SMP structural evaluation criteria and modify the ASR testing program
    were viewed appropriate by the inspectors and the Seabrook ASR Issue Technical
    Team, at this time. As stated above, the PODs use bounding assumptions not
    dependent on the degree of expansion measured. Additional inspections are planned by
    the NRC to evaluate NextEras ongoing corrective actions to resolve the non-
    conformance related to ASR-affected structures at the Seabrook Station.
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 1 sample)
    Plant Events
  a. Inspection Scope
    For the degraded plant equipment transient listed below, the inspectors reviewed and/or
    observed plant parameters, reviewed personnel performance, and evaluated
    performance of mitigating systems. The inspectors communicated the plant events to
                                                                                          Enclosure
 
                                                23
      appropriate regional personnel, and compared the event details with criteria contained in
      IMC 0309, Reactive Inspection Decision Basis for Reactors, for consideration of
      potential reactive inspection activities. The inspectors reviewed NextEras follow-up
      actions related to the events to assure that NextEra implemented appropriate corrective
      actions commensurate with their safety significance.
      B steam generator feedwater bypass valve failed to open on March 14, 2014
  b. Findings
      No findings were identified.
4OA5 Other Activities
.1    Cross-Cutting Aspects
      The table below provides a cross-reference from the 2013 and earlier findings and
      associated cross-cutting aspects to the new cross-cutting aspects resulting from the
      common language initiative. These aspects and any others identified since January
      2014 will be evaluated for cross-cutting themes and potential substantive cross-cutting
      issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment
      review.
              Finding                                    Old Cross-        New Cross-
                                                        Cutting Aspect    Cutting Aspect
        05000443/2013004-01, Inadequate                  H.1(b)            H.14
        Operability Determination Regarding Service
        Water Leakage and Associated TS Violation
.2    Buried Piping, TI-2515/182, Phase 2 (1 sample)
  a. Inspection Scope
      The licensees buried piping and underground piping and tanks program was inspected
      in accordance with paragraphs 03.02.a of the Temporary Instruction (TI) 2515/182, and
      it was confirmed that activities which correspond to the completion dates specified in the
      program, that have passed since the Phase 1 inspection was conducted, have been
      completed.
      The licensees buried piping and underground piping and tanks program was inspected
      in accordance with paragraph 03.02.b of the TI and responses to specific questions
      found in http:www.nrc.gov/reactors/operating/ops-experience/buried-pipe-ti-phase-2-
      insp-req-2011-11-16.pdf were submitted to NRC headquarters staff.
  b. Findings
      No findings were identified.
                                                                                        Enclosure
 
                                              24
4OA6 Meetings, Including Exit
    On April 10, 2014, the inspectors presented the inspection results to Mr. Kevin Walsh,
    Site Vice President, and other members of the Seabrook Station staff. The inspectors
    verified that no proprietary information was retained by the inspectors or documented in
    this report.
ATTACHMENT: SUPPLEMENTARY INFORMATION
                                                                                    Enclosure
 
                                            A-1
                              SUPPLEMENTARY INFORMATION
                                  KEY POINTS OF CONTACT
Licensee Personnel
K. Walsh, Site Vice President
T. Vehec, Plant General Manager
V. Brown, Senior Licensing Engineer
M. Chevalier, RP Supervisor
J. Connolly, Site Engineering Director
D. Currier, Emergency Planning Manager
K. Douglas, Maintenance Director
D. Flahardy, Radiation Protection Manager
M. Ossing, Licensing Manager
V. Pascucci, Nuclear Oversight Manager
D. Robinson, Chemistry Manager
T. Waechter, Nuclear Plant Shift Manager
              LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened/Closed
05000443/2014002-01              NCV      Scaffolding Installed with Insufficient Separation to
                                          Safety Related Equipment (Section 1R15)
                              LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
ON1090.13, Response to Natural Phenomena Affecting Plant Operations, Revision 1
OS1200.03, Severe Weather Conditions, Revision 20
OS1090.09, Station Cold Weather Operations, Revision 2
Section 1R04: Equipment Alignment
Procedures
OS1026.02, Operating the DG 1A Lube Oil System, Revision 14
OS1026.03, Operating DG 1A Jacket Water Cooling System, Revision 11
OS1026.04, Operating DG 1A Starting Air System, Revision 12
OS1026.05, Operating the DG 1A Fuel Oil System, Revision 14
OS1026.06, Operating the DG 1A Air Intake, Exhaust and Vacuum System, Revision 9
OS1036.01, Aligning the Emergency Feedwater System for Automatic Operation, Revision 17
OX1461.03, SEPS Operational Readiness Status Surveillance, Revision 1
                                                                                    Attachment
 
                                              A-2
Condition Reports
1880681        1904097      1936382        1934562        1952234
Section 1R05: Fire Protection
Procedures
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, PAB-F-1A-Z, PAB-F-1J-Z
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, RHR-F-1B-Z, RHR-F-2B-Z,
      RHR-3B-Z, RHR-F-4B-Z, RHR-F-4B-Z1, RHR-F-4B-Z2
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, RHR-F-1A-Z, RHR-F-2A-Z,
      RHR-3A-Z, RHR-F-4A-Z, RHR-F-4A-Z1, RHR-F-4A-Z2
Seabrook Station Fire Protection Pre-Fire Strategies, Volume II, SEPS-F-1-0 21-0
Seabrook Station Fire Protection Pre-Fire Strategies, Volume II, SW-F-1A-Z
Section 1R06: Flood Protection Measures
Condition Reports
00158490      01939967
Miscellaneous
Seabrook Station Moderate Energy Line Break Study
Section 1R11: Licensed Operator Requalification Program
Procedures
ER 1.1, Classification of Emergencies, Revision 53
Section 1R12: Maintenance Effectiveness
Procedures
ER-AA-201-2001, System and Program Health Reporting, Revision 4
PEG-40, Scoping Changes and Program Interfaces, Revision 5
PEG-45, Maintenance Rule Program Monitoring Activities, Revision 17
Condition Reports
1625261        1927188      1927781        1932096        1932711      1933065
1936449        1945056
Miscellaneous
ED/EDE 120 VAC System Health Report
EE-10-010, Maintenance Rule-PRA Basis Document PRA Risk Ranking and performance
      Criteria based on SSPSS-2009, Revision 1
NEI-99-02, Revision 7
SW-P-41C In-Service Testing Pump Data Sheet and Data Logs
Service Water System Health Report
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
OS1046.24, Removing EDE-I-1B from Service during Power Operation, Revision 3
Miscellaneous
Maintenance Rule a(4) Risk Assessment Report for Workweek 1407-12
PRA-301, MR (a)(4) Process for On-Line Maintenance Group Instruction, Revision 0
WM-AA-100-1000, Work Activity Risk Management, Revision 0
                                                                                  Attachment
 
                                              A-3
Maintenance Orders/Work Orders
40248951        40295189
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
IX1670.905, Seismic Monitoring Data Retrieval Following a Seismic Event, Revision 6
IX1670.919, SWPH Seismic Monitor Calibration, Revision 6
MA 4.8, Control of Scaffolding, Revision 10
MA 5.7, Station Barriers, Penetration Seals, and Fire Barrier Wrap, Revision 17
MS0599.16, Construction, Repair and Rework of Silicone Base Penetration Seals, Revision 7
MS0599.47, Erection of Scaffolding, Revision 2
MX0539.50, Emergency Diesel Generator Engine 34-Month Preventative Maintenance,
      Revision 6
MX0599.02, 18 Month Inspection of Technical Requirement Fire Rated Assembly Penetration
      Seals, Revision 2
OS1005.05, Safety Injection System Operation, Revision 25
OX1413.01, A Train RHR Quarterly Flow and Valve Stroke Tests, and 18 Month Valve Stroke
      Observation, Revision 20
OX1413.03, B Train RHR Quarterly Flow and Valve Stroke Tests, and 18 Month Valve Stroke
      Observation, Revision 11
OX1426.34, Diesel Generator 1A 18 Month Operability Surveillance, Revision 10
OX1436.02, Turbine Driven Emergency Feedwater Pump Quarterly and Monthly Valve
      Alignment, Revision 20
Condition Reports
0182029        0216388      0214364        0221647        1612785      1804255
1833819        1914234      1918208        1928775        1930569      1930855
1933827        1934585      1935442        1936576        1937513      1937679
1941086        1941153      1942147        1945355        1945771      1946400
1947827        1949876
Maintenance Orders/Work Orders
40153374        40191167      40199273      40228515      40245357      40250128
40291172        40291812      40291925      40292272
Miscellaneous
Calculation C-S-1-61035, Allowable CEVA Penetration Seal Opening Size, Revision 3
Colt-Pielstick PC2V Engine Vendor Manual
EE-98-019, Control of Temporary Loads in Seismic Areas
Engineering Evaluation 14-001, Scaffold & Temporary Equipment Engineering Evaluation
FP22849, Terry Turbine Instruction Manual, Revision 1
Preventive Maintenance Activity MS-CAT-12-SEAL-INSP
Penetration Seal Design, Seal No. PB-021-EV101-7502, FP4490R-01
Penetration Seal Design, Seal No. PB-021-EV101-7504, FP4492R-01
Tech Spec and Commitment Logs dated January 2-3, 2014
Technical Requirement TR21-4.3.3.3.1
Drawings
1-RH-B20662, Residual Heal Removal Sys. Train A Detail, Revision 22
                                                                                  Attachment
 
                                              A-4
Section 1R18: Plant Modifications
Procedures
IS1672.141, SW-F-6181, DG-E-42A Jacket Water Cooler Service Water Outlet Flow Calibration,
        Revision 6
EN-AA-100, Design Control Program, Revision 1
EN-AA-100-1003, Control of Design Interfaces, Revision 1
EN-AA-205-1100, Design Change Packages, Revision 9
EN-AA-205-1103, Equivalent Design Package, Revision 0
Condition Reports
1860416        1881903        1939926
Maintenance Orders/Work Orders
40233615
Miscellaneous
Specification 9763-006-174-1D, Data Sheets for Electronic Transmitters (Non-Class 1E),
        Revision 14
Drawings
1-SW-D20795, Service Water System Nuclear Detail, Revision 43
Section 1R19: Post-Maintenance Testing
Procedures
LS0556.08, Routine Preventative Maintenance 7.5 KVA Westinghouse Inverter, Revision 8
LS0556.09, Replacement of Ferro-Resonant Transformers and Capacitors in Westinghouse 7.5
        KVA Inverters, Revision 5
MM-AA-100, Conduct of Maintenance, Revision 4
MX0539.63, Emergency Diesel Generator Exhaust Valve Removal, Replacement, and Belleville
        Washer Replacement, Revision 2
ON0443.113, Portable Diesel Driven Pump Annual Functional Test, Revision 5
OS1046.24, Removing EDE-I-1B from Service during Power Operation, Revision 2
OS1047.01, Vital Inverter Operation, Revision 14
OS1247.01, Loss of a 120VAC Instrument Panel, Revision 17
OS1412.10, Thermal Barrier Cooling Water Pump Monthly Rotation, Revision 6
OX1416.04, Service Water Quarterly Pump and Discharge Valve Test and Comprehensive
        Pump Test, Revision 19
OX1446.03, Electrical Bus Weekly Operability, Revision 12
OX1456.86, Operability Testing of IST Pumps, Revision 10
SAG-9, PDDP and Hose Trailer Deployment, Revision 5
Condition Reports
0221649        1931807        1932393      1932711      1933020      1933808
1934499        1934512        1934562      1936449      1936703      1936858
1940192        1940751        1941180      1945142      1946434      1946440
1946653        1947234        1947394
Maintenance Orders/Work Orders
01168951      01207733      01210186      0305866      0320197      40196428
40228208      40239216      40259353      40290809      40295189      40298167
40298355
                                                                                  Attachment
 
                                              A-5
Miscellaneous
Calculation C-S-1-86208, Extreme Damage Mitigating Strategy Flow Capability, Revision 3
Calculation C-S-1-50014, SW Pumps (SW-P-41A thru D) IST Uncertainties, Revision 0
Calculation 9763-3-ED-00-34-F, AC Ground Detection System
FP35465, SEPS Generator Set Technical Manual
FP500076, Godwin PDDP Instruction Manual, Revision 6
NASA TN D-8177, Apollo Experience Report-Detection and Minimization of Ignition Hazards
        From Water/Glycol Contamination of Silver-Clad Electrical Circuitry
Standing Order SOO-14-001, B.5.b Pump Status, dated January 28, 2014
Westinghouse Technical Bulletin NSID-TB-87-09
Drawings
4950C70, Sheet 4, Inverter Schematic
1-NHY-310105, Sheet E02a, UPS 1-I-1B Vital Instrument Distribution Panel 1-PP-1B
1-NHY-310231, Sheet I20a, Motor/Load List Motor Control Center 1-EDE-MCC-615, Revision 7
1-NHY-310895, Sheet B4Qa, Thermal Barrier PCCW Recirc. Pump P-322B Schematic
        Diagram, Revision 2
Section 1R22: Surveillance Testing
Procedures
EN-AA-205-1102, Temporary Configuration Changes, Revision 5
MA-AA-100-1011, Equipment Troubleshooting, Revision 0
ON0443.113, Portable Diesel Driven Pump Annual Functional Test, Revisions 1 and 5
ON0443.114, 18 Month B.5.b Equipment Inventory Surveillance, Revision 10
OS1001.04, RCS Unidentified Leak Rate Action Level Exceedence, Revision 0
OS1007.01, Automatic and Manual Rod Control, Revision 12
OX1408.06, Controlled Leakage Monthly Surveillance, Revision 6
OX1416.06, Service Water Cooling Tower Pumps Quarterly and 2 Year Comprehensive Test,
        Revision 21
OX1423.07, Monthly Testing of Train A Enclosure Emergency Exhaust, Revision 8
OX1426.05, DG 1B Monthly Operability Surveillance, Revision 28
OX1426.19, Aligning DG 1B Controls for Auto Start, Revision 3
OX1456.46, Train B ESFAS Slave Relay K608 Quarterly Go Test, Revision 7
Condition Reports
1822620        1929096      1933872        1934512      1941467          1949526
1949825
Maintenance Orders/Work Orders
40149909      40178159      40235580      40244087      40245450        40259353
40264853      40246660      40246721      40246722      40287931        40290041
40290042
Miscellaneous
1-SW-OT-031 IST Pump Data Log
1-SW-OT-011 IST Pump Data Log
Activity 1-CP-CP-113-CRDM-1, CRDM Current Command Trace Acquisition, Revision 6
ASME OM CODE-2004
Calc 88-002, IST Calculation of Total Developed Head for Service Water and Cooling Tower
        Pumps
Engineering Evaluation SS-EV-98006, Revision 1
                                                                                    Attachment
 
                                              A-6
Drawings
1-CS-B20725, Chemical & Volume Control Sys. Seal Water Detail, Revision 20
1-NHY-310932, Cntmnt Encl Emer Exh Fan 1-FN-4A Schematic Diagram, SH-BB3a, Revision 9
1-NHY-310932, Cntmnt Encl Emer Exh Fan 1-FN-4A Legend & SW Development, SH-BB3b,
      Revision 10
1-NHY-503515, EAH - Contn. Encl. Emer Exh Fltr Fan Logic Diagram, Revision 7
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Procedures
ER 3.3, Emergency Operations Facility Operations, Revision 51
Section 1EP6: Drill Evaluation
Procedures
EP-AA-101-1000, Nuclear Division Drill and Exercise Procedure, Revision 5
ER 1.1, Classification of Emergencies, Revision 52
ER 1.2, Emergency Action Plan Activation, Revision 61
ER 3.1, Technical Support Center Operations, Revision 53
Condition Reports
1948051        1910629
Miscellaneous
ER 2.0B, Seabrook Station State Notification Fact Sheet, Revision 31
Form EPDP-03A, EP Cornerstone Reporting and Information Form, Revision 23
Section 2RS1: Radiation Hazard Assessment and Exposure Control
Procedures
HD0958.04, Posting of Radiologically Controlled Areas, Revision 33
HD0958.03, Personnel Survey and Decontamination Techniques, Revision 24
HD0958.13, Generation and Control of Radiation Work Permits, Revision 39
HD095817, Performance of Routine Radiological Surveys, Revision 12
HD0958.19, Evaluation of Dosimetry Abnormalities, Revision 37
HD0958.38, Evaluation of Isotopic Mix, Revision 29
HN0958.25, High Radiation Area Control, Revision 37
HN0958.30, Inventory and Control of LHRA or VHRA Keys and Locksets, Revision 26
HX0958.23, Radioactive Source Control, Revision 20
RP-AA-100-1001, RP Conduct of Ops, Revision 3
RP-AA-100-1002, Radworker Instructions and Responsibilities, Revision 1
RP-AA-101, Personnel Monitoring Program, Revision 0
RP-AA-101-1001, Radiation Protection Conduct of Operations, Revision 3
RP-AA-101-1002, Dosimetry Data Processes for Sentinel Software, Revision 3
RP-AA-101-2004, Method for Monitoring and Assigning Effective Dose Equivalent for High Dose
      Gradient Work, Revision 3
RP-AA-102-1001, Area Rad Surveys, Revision 0
RP-AA-102-1000, Alpha Monitoring, Revision 0
RP-AA-102-1002, Dosimetry Data Process for Sentinel, Revision 3
RP-AA-103-1001, Posting Requirements, Revision 1
RP-AA-103-1002, High Rad Controls, Revision 1
RP-AA-107-1003, Unconditional and Conditional Release of Material, Revision 1
                                                                                Attachment
 
                                              A-7
Audits, Self-Assessments, and Surveillances
Seabrook Station Radiation Protection Department Self Evaluation and Trend Analysis
        Report for 4th Quarter 2013, January 31, 2014
Quick Hit Assessment Report 1928716, NRC 71124.01 and .02 Radiological Hazard
Assessment and ALARA Planning and Control, February 3, 2014
Seabrook Nuclear Oversight Report SBK-14-001, Radiation Protection and Radwaste
Programs, February 24, 2014
Condition Report
01836289        01855852      01898310      01903346    01906680      01934952
01940807        01941338      01943228      01945353    01945687
Miscellaneous
Seabrook Updated Final Safety Analysis Report
Seabrook Survey M-20140304-5, Initial Entry into Fuel Transfer Canal, March 4, 2014
Seabrook Survey M-20140122-3, HSYQ066A FSB-21-FB202 Quarterly, January 22, 2014
Seabrook Survey M-20140203-2, HSYQ-082B WPC-(-26)-MF-106 Quarterly, February 3, 2014
Seabrook Survey M-20130904-3, HSXQ-079A PAB-7-PB309 Quarterly, September 4, 2013
Seabrook Survey M-20131123-2, HSYQ-083A WPC-(-20)-MF102 Quarterly, November 23, 2013
Seabrook RWP 14-0015 High Integrity Container/Liner shipping Preparation to Include Capping,
        Weighing and Transfer to WPB, January 1, 2014
Seabrook RWP 14-0022 Inspect CS Valves inside Letdown Valve Room at Power, January 15, 2014
Seabrook RWP 14-0027 PAB Demin Alley Work/Entry, December 31, 2013
Seabrook RWP 14-0058 FSB Transfer Canal RP Survey, Decon and Maintenance Support
        Activities, March 2, 2014
Seabrook Non Exempt Source Inventory and Leak Test, September 14, 2013
Seabrook Exempt Source Index, July, 22, 2013
Seabrook Dosimetry Abnormality Occurrence Report CR 01904744, November 18, 2013
Seabrook Dosimetry Abnormality Occurrence Report CR 01906680, December 4, 2013
Seabrook 2014 Air Sample Log, March 7, 2014
Seabrook Lesson Plan HP1188C Alpha Monitoring Course, April 15, April 29 and May 13, 2013
Seabrook Log of VHRA and LHRA Access Points, March 5, 2014
Seabrook LHRA In Service Key Box Log, January 31, 2014
Seabrook LHRA/VHRA Key Issue Log, March 4, 2014
Seabrook HRA/LHRA Briefing Acknowledgement Form, March 5, 2014
Work Order
40235669
Section 2RS2: Occupational ALARA Planning and Controls
Procedures
RP-AA-104 ALARA Program, Revision 2
RP-AA-104-1000, ALARA Implementing Procedure, Revision 5
Audits, Self-Assessments, and Surveillances
Seabrook Station Radiation Protection Department Self Evaluation and Trend Analysis
        Report for 4th Quarter 2013, January 31, 2014
Quick Hit Assessment Report 1928716, NRC 71124.01 and .02 Radiological Hazard
Assessment and ALARA Planning and Control, February 3, 2014
Seabrook Nuclear Oversight Report SBK-14-001, Radiation Protection and Radwaste
        Programs, February 24, 2014
                                                                                  Attachment
 
                                              A-8
Corrective Action Document
01836312      01843713      01856278      01867573      01872019      01883752
01890162      01893578      01896323      01904259      01930630      01944341
Miscellaneous
EPRI Standard Radiation Monitoring Program Results through OR 15, September 25, 2012
Seabrook Updated Final Safety Analysis Report
Seabrook Post Outage Critique: ALARA and Station Dose Performance, June 2013
Seabrook Station Nuclear Plant 5-Year ALARA Plan 2013-2017, July 31, 2013
Seabrook ALARA Review Board Meeting 13-04, December 11, 2013
Seabrook ALARA Review Board Meeting 14-01, March 3, 2014
Seabrook Temporary Shielding Log for OR 16, March 2014
Pre-Job ALARA Review Package: 13-01 Dry Fuel Transfer from Pool to Pad and Associated
      Tasks for 8 ISFSI Casks, June 27, 2013
Post-Job ALARA Review: 13-01 Dry Fuel Transfer from Pool to Pad and Associated Tasks
      for 8 ISFSI Casks, December 4, 2013
Pre-Job ALARA Review Package 14-01 OR 16 Reactor Dissassembly and Reassembly,
      December 26, 2013
Pre-Job ALARA Review Package 14-02 OR 16 Steam Generator Eddy Current Testing and
      Tube Plugging, Febuary 25, 2014
Pre-Job ALARA Review Package 14-03 OR 16 In Service Inspection, Febuary 25, 2014
Pre-Job ALARA Review Package 14-07 OR 16 Fuel Handling Project, Febuary 25, 2014
Pre-Job ALARA Review Package 14-09 OR 16 RCP Seal Replacement, Febuary 25, 2014
Pre-Job ALARA Review Package 14-10 OR 16 Scaffolding, Febuary 25, 2014
Pre-Job ALARA Review Package 14-13 Replace Rx Ventillation Ducting Under Vessel with
      New Design, Febuary 25, 2014
Post Project Critique Dry Fuel Storage, Seabrook Station 2nd Loading Campaign, October 2013:
Section 4OA1: Performance Indicator Verification
Procedures
NAP-206, NRC Performance Indicators, Revision 6
Miscellaneous
LIC-13017, Documentation Supporting the Seabrook Station NRC 1st Quarter 2013
      Performance Indicator Submittal
LIC-13036, Documentation Supporting the Seabrook Station NRC 2nd Quarter 2013
      Performance Indicator Submittal
LIC-13037, Documentation Supporting the Seabrook Station NRC 3rd Quarter 2013
      Performance Indicator Submittal
LIC-14004, Documentation Supporting the Seabrook Station NRC 4th Quarter 2013
      Performance Indicator Submittal
MSPI Derivation Reports
Section 4OA2: Problem Identification and Resolution
Non-Destructive Test Reports
40265240-01, UT Extent of Condition Thickness Examination, B Train
40265234-01, UT Extent of Condition Thickness Examination, A Train
Condition Reports
01897164      01637922
                                                                                  Attachment
 
                                                A-9
Maintenance Orders/Work Orders
40265234      40268662        40268965      40268967      94080896      94080893
Drawings
1-SW-B20794, Service Water System Nuclear Detail (Service Water Pump House)
1-SW-B20795, Service Water System Nuclear Detail (Turbine Bldg., Aux Bldg.)
SK-EC270504-2000, Installation Detail Service Water Piping Repairs
SK-EC156603-2001, Installation Detail of Weldolet Service Water Pipe Repair
SW 1802-09-EC 2080429, SW Piping Repair (Flush Patch) Line No 1-SW-1802-004
Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion
Procedures
MA-AA-100-1011, Equipment Troubleshooting, Revision 0
OS1235.03, SG Level Instrument Failure, Revision 14
Condition Reports
1948268        1952067
Maintenance Orders/Work Orders
40300038
Miscellaneous
Instrument Loop Diagram ILD-1-FW-L04220, Steam Generator RC-E-11B Feedwater Bypass
      Flow (Loop 2) 1-FW-L-4220, Revision 14
Operational Decision Making Bulletin, dated 3/18/2014
Section 4OA5: Other Activities
Procedures
ER-AA-102 Underground Piping and Tank Integrity Program, Revision 6
ER-AA-102-1000 Underground Piping and Tanks Integrity Examination Procedure, Revision 2
Seabrook Station Underground Piping and Tanks Inspection Program, Revision 2
SH 6.4 Dig Safe (01/06/12) Excavation of Site Locations Penetrating Plane of Ground,
    Revision 13
Miscellaneous
AR 00213052-01-00, Complete Initiative Action 1 Status Complete
AR 00213052-02-00, Complete Initiative Action 2 Risk Ranking Buried Piping
AR 00213052-03-00, Complete Initiative Action 3 Develop Inspection Plan by 06/30/11
AR 00213052-04-00, Complete Initiative Action 4 Implement Inspection Plan 06/30/12
AR00213052-05-00, Develop Asset Management Plan Status Complete 01/26/10
AR00213052-06-00, Inspect Buried Piping Containing radioactive material 09/16/10
AR00213052-07-00, Underground Piping and Tanks Procedure Oversight 12/22/11
AR00213052-08-00, Prioritize Underground Piping and Tanks 06/27/12
AR 01600464 (12/09/10) Aux Steam and Aux Steam Condensate Leak w/i guard pipe
NRC Temporary Instruction 2515/182, Issue 11/17/11 and 8/8/13; Review of the Implementation
      of the Industry Initiative to Control Degradation of Underground Piping and Tanks
NEI 09-14 Initial Issue, November 2009 Guideline for the Management of Underground Piping
      and Tank Integrity
NEI 09-14 Guideline for the Management of Underground Piping and Tank Integrity, Revision 1,
      December 2010
                                                                                    Attachment
 
                                              A-10
NEI 09-14 Guideline for the Management of Underground Piping and Tank Integrity Inspection
    and Analysis Methodologies, Revision 3
                                    LIST OF ACRONYMS
ADAMS                Agencywide Document Access and Management System
ALARA                as low as reasonably achievable
CAP                  corrective action program
CFR                  Code of Federal Regulations
CSSI                containment spray safety injection
EAL                  emergency action level
EDG                  emergency diesel generator
EFW                  emergency feedwater
EPD                  electronic personal dosimeter
ESFAS                engineered safety features actuation system
HRA                  high radiation area
IMC                  Inspection Manual Chapter
LHRA                locked high radiation area
MR                  Maintenance Rule
NCV                  non-cited violation
NEI                  Nuclear Energy Institute
NRC                  Nuclear Regulatory Commission
OOS                  out of service
PAB                  primary auxiliary building
RCS                  reactor coolant system
RG                  Regulatory Guide
RHR                  residual heat removal
RPM                  Radiation Protection Manager
RWP                  radiation work permit
SDP                  significance determination process
SEPS                supplemental emergency power system
SSC                  structure, system, or component
SW                  service water
TI                  temporary instruction
TS                  technical specification
UFSAR                Updated Final Safety Analysis Report
VAC                  volts alternating current
VHRA                very high radiation area
WO                  work order
                                                                                Attachment
}}

Revision as of 08:22, 21 November 2019

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