IR 05000443/2011003

IR 05000443/2011003; on 04/01/2011-06/30/2011; Seabrook Station, Unit No. 1; Routine Integrated Report; Fire Protection; Operability Evaluations

ML112241543
Person / Time
Site:	Seabrook
Issue date:	08/12/2011
From:	Burritt A L Reactor Projects Branch 3
To:	Freeman P NextEra Energy Seabrook
Burritt A L RGN-I/DRP/PB3/610-337-5069
References
IR-11-003
Download: ML112241543 (49)

Text

UNITED STATES N UCLEAR REGULATORY COMMISSION REG]ON I 475 ALLENDALE ROAD KING OF PRUSSIA. PA 19406-1415 August 72, 2OIL Mr. Paul Freeman Site Vice President Seabrook Nuclear Power Plant NextEra Energy Seabrook, LLC c/o Mr. Michael O'Keefe P.O. Box 300 Seabrook, NH 03874

SUBJECT: SEABROOK STATION, UNIT NO. 1 - NRC INTEGRATED INSPECTION REPORT 05000443/201 1 003

Dear Mr. Freeman:

On June 30, 201 1, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at Seabrook Station, Unit No. 1. The enclosed report documents the inspection findings discussed on July 13,2011, with Mr. E. Metcalf and other members of your statf.These inspections examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

The report documents three NRC-identified findings of very low significance (Green) that were determined to involve a violation of NRC requirements.

However, because of the very low safety significance and because the issues were entered into your corrective action program, the NRCis treating the findings as non-cited violations (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.lf you contest any NCV in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region l; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident lnspector at the Seabrook Station. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region l, and the NRC Resident lnspector at the Seabrook Station. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.qov/readinq-rm/adams.html (the Public Electronic Reading Room).

Sincerely,I-t-/ it rfi4 t lrV Arthur L. Burritt, Chief Projects Branch 3 Division of Reactor Projects Docket No. 50-443 License No: NPF-86

Enclosure:

lnspection Report No. 050004431201 1003 wi

Attachment:

Supplemental Information cc w/encl: Distribution via ListServ

SUMMARY OF FINDINGS

lR 0500044312011003; 0410112011-0613012011;

Seabrook Station, Unit No. 1; Routine lntegrated Report; Fire Protection;

Operability Evaluations.

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

Three Green findings were identified.

The significance of most findings is indicated by their color (Green, White, Yellow, or Red) and determined using Inspection Manual Chapter (lMC) 0609, "significance Determination Process" (SDP). The cross cutting aspect of a finding is determined using the guidance in IMC 0310, "Components Within the Cross-Cutting Areas." Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.Gornerstone:

Mitigating Systems

Green.

The inspectors identified a non-cited violation (NCV) of Technical Specification tt.Sl O.Z.t.h, which requires that written procedures be established and implemented for the fire protection program. Contrary to TS 6.7.1.f , the inspectors identified combustible materials which were not controlled per fire protection procedure FP 2.2, Revision 12'Specifically, (i) combustible materials were stored within three feet of an energized sample panel in the primary auxiliary building room PB404, a PRA risk significant area;and, (ii) combustible materials in excess of the permissible amounts were stored in waste process building area W8505. The in$pectors identified materials stored in WB505 in excess of FP 2.2 limits on three occasions.

Collectively, the NRC observations indicate a weakness in the programmatic control of combustible materials despite the fact that in each case the combustible materials were promptly removed following identification by the inspector.

Seabrook entered this performance deficiency into their corrective action program.The performance deficiency was more than minor because, if left uncorrected, inadequate control of combustibles could affect the Mitigating Systems cornerstone objective to assure external factors (fires) do not impact the availability and reliability of syitems which mitigate events. The inspectors assessed the finding using Appendix F of the Significance Determination Process (SDP) Based on a degradation rating of low, which screens to Green in the fire protection SDP, the finding is of very low safety significance.

This finding has a cross-cutting aspect in Human Performance, Work Piactices tH.4(b)l because Seabrook personnel did not follow procedures for the control of transient combustibles. (Section 1R05)Green. The inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 6.7.1.a that requires written procedures be established and implemented, including administrative procedures that define authorities and responsibilities for safe operation with respect to operability determinations.

Contrary to TS 6.7.1.a, NextEra identified a degraded and nonconforming condition related to reduced modulus of elasticity for buiidings housing safety related equipment on May 27,2011 but did not complete an operability determination until EC250348 was issued on June 28,2011 (AR1664399).

The delayed entry of the issue into the corrective action process to assess operability was contrary to Section 4.3 of EN-AA-203-1001 that requires operability assessments be

completed in a time frame commensurate with the safety significance of the issue (within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />). Seabrook subsequently completed an evaluation of the concrete issues and determined that the buildings housing safety-related equipment remained operable.Seabrook entered this performance deficiency into their corrective action program.The performance deficiency was more than minor because a reasonable doubt of operability for the affected concrete structure$

existed until further engineering evaluations were completed to demonstrate the structures and systems that they housed would remain functional under design and licensing basis conditions.

The finding affected the Mitigating Systems cornerstone Objective to ensure the availability, reliability and capability of systems that respond to initiating events in order to prevent core damage. The issue was evaluated using IMC 0609, "Significance Determination Process" (SDP), and was determined to be of very low safety significance (Green)because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of a barrier function, and was not potentially risk significant for external events. The finding had a cross cutting aspect in the area of problem identification and resolution, P.1(a), because NextEra did not enter identified degraded concrete conditions for several site buildings into the corrective actions process in a timely manner, which would have ensured the shift manager completed timely operability evaluations for the affected structures. (Section 1R15.3)Green. The inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 6.7.1.a that requires that written procedures be established and implemented, including administrative procedures that define authorities and responsibilities for safe operation with respect to operability determinations.

Contrary to TS 6.7 .1.a, NextEra identified a degraded condition related to seryice water flow to the B emergency diesel generator (EDG) heat exchanger (HX) on June 28,2011 but did not fully evaluate the reduced flow under all plant conditions as required by NextEra procedure EN-AA-203-1001. Fouling of the heat exchanger tubes was subsequently identified and mitigated.

Seabrook also completed an evaluation of the B EDG service water flow issues and determined that the EDG remained operable.

Seabrook entered this performance deficiency into their corrective action program.The performance deficiency was more than rninor because a reasonable doubt of operability existed untilfurther engineering evaluations were completed to demonstrate adequate service water flow to the B EDG HX existed and the B EDG remained functional under design and licensing basis conditions.

The finding affected the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events in order to prevent core damage.The issue was evaluated using IMC 0609, "significance Determination Process" (SDP), and was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of a barrier function, and was not potentially risk significant for external events. The finding had a cross cutting aspect in the area of problem identification and resolution, P.1(c), because NextEra personnel did not thoroughly assess EDG operability to assure reduced HX SW flow was acceptable under all operating conditions, or assure appropriate corrective actions were timely completed.(Section 1R15.4)

5

REPORT DETAILS

Summarv of Plant Status The plant was shutdown at the start of the report period to conduct refueling outage OR14.NextEra completed reactor refueling and maintenance activities on the reactor and plant secondary systems. The plant was started up and the reactor was taken critical on May 23, 2011, and operation at full power resumed on May 26, 2011. The turbine was taken offline for maintenance on the secondary plant on June 4, 2011. Seabrook returned to full power on June 6,2011.1.

REACTOR SAFETY

Cornerstones:

Initiating Events, Mitigating Systems, and Barrier Integrity 1R01 Adverse Weather Preparation (71111.01 - 2 sample).1 Readiness for Seasonal Extreme Weather Conditions

a. Inspection Scope

The inspector completed one seasonal extreme weather conditions inspection sample.The inspectors assessed NextEra's readiness for the onset of hot weather. The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) descriptions for related design features and verified the adequacy of the station procedures for hot weather protection.

The inspectors reviewed NextEra's actions per procedure ON1490.09 for seasonal readiness, and procedure OS1200.03 for severe weather. The inspectors also performed walkdowns of susoeptible systems, specifically the emergency feedwater, electrical distribution and service water systems. The inspectors reviewed deficiencies related to extreme weather preparation and verified the issues were entered into the corrective action program. The documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

.2 Readiness

of Offsite and Alternate AC Power Svstems

a. Inspection Scope

The inspectors completed one summer readiness of offsite and alternate AC power systems inspection sample. The review focu$ed on NextEra procedure 051246.02,"Degraded Vital AC Power." The inspectors verified that plant features were maintained and procedures for operation were adequate to ensure the continued availability of AC power systems. The inspectors verified that communication protocols with the transmission system operator were adequate to ensure that appropriate information was exchanged when issues arose that could impact the offsite power system. The inspectors also observed NextEra's implementation of OS1246.02 during periods that Enclosure 6 challenged grid conditions between April 1, 2011 and June 30, 2011. The inspection included walkdowns of the onsite normal and emergency AC power systems and the inspectors reviewed deficiencies related to summer readiness of offsite and alternate AC power systems and verified these issues were entered into the corrective action program. The documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

1R04 Equipment

Aliqnment (71111.04Q - 5 samples;7111 1.04S - 1 sample).1 PartialWalkdown

a. Inspection Scope

The inspectors completed five partial system walkdown inspection samples for the plant systems listed below. The inspectors verified that valves, switches, and breakers were correctly aligned in accordance with Seabrook's procedures and that conditions that could affect system operability were appropriately addressed.

The inspectors reviewed applicable piping and instrumentation drawings and system operational lineup procedures.

The documents reviewed are listed in the Attachment.

o Primary component cooling water (PCCW) "A" Train with "8" service water (SW) and"8" PCCW out of service for work performed on April 11, 2011.. "A" train primary component cooling water (PCCW) during planned unavailability of the "B" train PCCW and SW systems on April 27,2011 through May 2, 2011'o Reactor and support system alignments on April 22, 2011, in preparation for plant startup from Mode 6.. Residual heat removal (RHR) system alignment for low temperature over pressure protection during shutdown cooling operations on April 1,2011 through April 4, 2011.. "B" train RHR on May 5,2011 through May 10,2011 during the removal and replacement of 1-RHR-P-8A motor and seal package.Findinqs No findings were identified.

Complete Svstem Walkdown Insoection Scope The inspectors completed one complete system walkdown inspection sample on the service water system, specifically, Train "A" during a Train "B" pipe replacement and SW ocean outage. The inspectors walked down the accessible portions of the system to verify the system's overall material condition; that valves were correctly positioned; that electrical power was available; that major system components were properly labeled;that hangers and supports were correctly installed and functional; and that ancillary equipment or debris did not interfere with system performance.

The inspectors reviewed b.,2 Enclosure 7 plant procedures, system drawings, the UFSAR, and the technical specifications (TS).The documents reviewed are listed in the Attachment.

b. Findinos No findings were identified.

1R05 Fire Protection

(71111.05Q - 5 samples).1 Quarterlv Review of Fire Areas Inspection Scope The inspectors completed five quarterly fire protection inspection samples. The inspectors examined the areas of the plant listed below to assess: the control of transient combustibles and ignition sources; the operational status and material condition of the fire detection, fire suppression, and manual fire fighting equipment; the material condition of the passive fire protection features; and the compensatory measures for out-of-service or degraded fire protection equipment.

The inspectors verified that the fire areas were maintained in accordance with applicable portions of Fire Protection Pre-Fire Strategies and Fire Hazard

Analysis.

The documents reviewed are listed in the Attachment.. Primary auxiliary building 53 FT (PAB-F-3A-Z).. Containment 26 FT (C-F-3-2).. Fuel storage building 7 FT (FSB-F-1-A).

o Site yard area with focus on containment outage access (PLT-F-1-0).. Containment 0 FT and +25 FT (C-F-Z-Z and C-F-3-Z).Findinqs Inadequate Control of Combustible Materials lntroduction:

The inspectors identified a non-cited violation (NCV) of Technical Specification (TS) 6.7.1.h, which requires that written procedures be established and implemented for the fire protection program. The inspectors identified combustible materials that were not controlled per NextEra procedure FP 2.2 in the primary auxiliary building and in the waste process building room W8505. The inspectors identified materials stored in W8505 in excess of FP 2.2 limits on 3 occasions from April 15,2011 to July 1,2011. Collectively, the NRC observations indicate inadequate programmatic control of transient combustible materials.

Seabrook removed the improperly stored material identified by the inspector and entered this performance deficiency into their corrective action program.Description:

Procedure FP 2.2, "Control of Combustible Materials", provides requirements for controlling combustible materials at Seabrook.

The inspectors identified the following conditions that did not meet the requirements of FP 2.2: b.1.Enclosure I

(a) During a walkdown of the Waste Processing Building area W8505 on April 15,2011, the inspectors identified ten rolls of new plastic bags. Section 4.7 of FP 2.2 allows permanent storage of combustible materials in room W8505 in quantities specified for normal operations on bag rack only (i.e., three rolls of bags). In addition to the bags on the rack, seven additional rolls of bags were beside the rack. Procedure FP 2.2, Section 4.4.3, provides a permit threshold for quantities that exceed 100 pounds of National Fire Protection Association (NFPA) flammability category 1 solid materials (Class A materials).

No transient combustible material permit was issued.The additional seven rolls of bags exceeded 100 pounds. The inspectors discussed this issue with Operations Management (OM). The materials were removed.On June 15,2011, inspectors identified a similar condition in room W8505 in that three additional rolls of bags were near the bag rack. The inspectors discussed the issue with Shift Manager (SM) and Fire Brigade Leader (FBL). Condition Report 1661217 was initiated and the non-permitted materials were removed. On July 1, 2011, the inspectors identified the same conditions in room W8505. Specifically, three additional rolls of bags were near the bag rack. Further, there was a half-full 55 gallon drum of used oil in the area. No transient combustible permit existed for the materials.

The inspectors discussed these observations with the on-duty FBL and SM. The non-permitted materials were removed. Condition Reports 1666354 and 1666363 were initiated to enter this issue in the corrective action program.(b) During a walkdown of the primary auxiliary building (PAB) room PB404 on June 15, 2011, the inspectors observed a cardboard box of paper towels and partially filled plastic bag stored next to (within three feet) 1-SS-CP-166-8, an energized sample analysis control panel. The materials were used by chemistry personnel to obtain steam generator blowdown samples. The sample panel area is an elevated platform that is approximately three feet wide. Section 4.2.5.c of FP 2.2 states that materials are not to be stored within three feet of energized electrical equipment (panels, etc.).Room PB4O4 houses the primary component cooling water pumps and heat exchangers, and is a designated PRA risk significant area, defined as an area that contributes the greatest majority of risk of core damage to fire initiated events. The inspectors discussed the observations with the on-duty Fire Brigade Leader and Shift Manager. The combustible materials were removed. Condition Report AR 1661010 was initiated to enter this issue in the corrective action program.Collectively, these NRC observations indicate programmatic weakness in the control of combustible materials.

The failure by worker$ to follow procedures and the ineffective NextEra actions to keep combustibles in WB505 below limits set by FP 2.2 raise a concern with the control of combustibles which, if left uncorrected, could lead to a more significance safety concern. Further, the inspectors identified that restrictions contained in the Final Safety Analysis Report, Appendix A, Responses to BTP APCSB 9.5-1 -materials near safety related tanks, were not reflected in FP 2.2. NextEra entered this issue into the corrective action program as AR 1667113.Analvsis:

The inspectors determined that the failure to properly implement procedure FP ZZtor the controlof transient combuStible materialwas a performance deficiency.

This finding was considered more than minor because, if left uncorrected, inadequate control of combustibles could affect the Mitigating Systems cornerstone objective to assure external factors (fires) do not impact the availability and reliability of systems which mitigate events.Enclosure 9 The inspectors performed a significance determination of this issue using IMC 0609,"significance Determination Process" (SDP), Appendix F, "Fire Determination Significance Determination Process." The issue met the Phase I qualitative screening criteria as discussed in Appendix F.Based on an evaluation using Step 1 of Appendix F, the inspectors determined the finding affected the category of Fire Prevention and Administrative Controls in that combustible material was not being properly Controlled; the finding had a "low" degradation rating; and, the finding was of very low safety significance (Green). The inspectors determined this event affected the cross-cutting area of H.4.(b), Human Performance, Work Practices, because of the failure of workers to follow station procedures.

Enforcement:

TS 6.7.1.h requires that written procedures shall be implemented for the Fire Protection Program (FPP). Fire Program procedure FP 2.2,"Controlof Combustible Materials," Revisionl2, limits the quantity of transient combustible material stored in W8505 (Section 4.7) and near electrical panels in P8404 (Section 4.2.5.c).

Contrary to the above, NextEra did not limit the quantity of transient combustible material stored in W8505 and near electrical panels in P8404 in accordance with the requirements of procedure FP 2.2. Specifically, NextEra stored ten rolls of plastic bags in WB505 and stored a cardboard box of paper towels and a partially filled plastic bag within three feet of an energized sample analysis control panel in PB404. Because the failure to control combustible materials was of very low safety significance and has been entered into NexEra's corrective action program (ARs 1661010, 1666354, 1666363), this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy (NCV 05000443/201 1 003-01, Inadequate Gontrol of Combustible Materials).

1R07 Heat Sink Performance

(71111.07 - 1 sample)Insoection Scope The inspectors completed one heat sink performance inspection sample. Specifically the inspectors reviewed the 2011 testing of the "B" component cooling water heat exchanger to verify that the heat exchanger could fulfill its design function.

The inspectors reviewed thermal performance monitoring (WO 01202862), trending data for heat exchanger temperatures and fouling factors, and ES1850.017, "SW Heat Exchanger Program".

The inspectors interviewed the system engineer to evaluate the process used to monitor the heat exchanger and commitments in Generic Letter 89-13,"Service Water System Problems Affecting Safety-Related Equipment." The inspectors performed system walk downs and reviewed condition reports to verify that issues associated with the heat exchanger were identified and corrected.

The documents reviewed are listed in the Attachment.

Findinos No findings were identified.

a.b.Enclosure 10 1R08 Inservice Inspection (71111.08 - 1 sample)a. Inspection Scope The purpose of this inspection was to review and assess the effectiveness of NextEra's In-service Inspection (lSl) program for monitoring degradation of the reactor coolant system (RCS) boundary, risk significant piping system boundaries, and the containment boundary.

The inspectors reviewed a sample of nondestructive examination (NDE)activities to verify compliance with American Society of Mechanical Engineers (ASME)Boiler and Pressure Vessel Code, Section Xl and applicable NRC Regulatory Requirements.

In addition, the inspectors reviewed samples of completed non-destructive examinations, inspection procedures and inspection test reports to verify compliance with the ASME Code, Section Xl. The inspectors reviewed the results of the reactor vessel nozzle dissimilar weld evaluation in the post-MSIP configuration (AR1644106).

Also, the inspectors reviewed repair and replacement activities which involved use of welding and NDE on pressure boundary risk significant systems.The inspectors observed the performance of NDE activities in process and reviewed documentation and examination reports for additional nondestructive examinations.

Non-destructive test processes inspected and reviewed included Visual (W), Magnetic Particle (MT), Penetrant (PT), Eddy Current (ECT), and Ultrasonic (UT) testing. The sample selection was based on the inspection procedure objectives, risk significance and sample availability.

The inspectors reviewed examination procedures, procedure and personnel qualifications and examination test results.The inspectors reviewed the procedures used to perform visual examinations for indications of boric acid leaks from pressure retaining components including the vessel upper head penetrations and their connections to the control rod drive mechanisms.

The inspectors reviewed samples of operability evaluations, engineering evaluations and corrective actions provided for active and inactive boric acid leaks and determined they were consistent with the requirements of the ASME Code and 10 CFR 50, Appendix B, Criterion XVl, Corrective Action.The inspectors performed a visual examination of the containment steel shell at the zero and minus 26 foot elevations within containment to evaluate the reported condition of the liner coating. The inspectors reviewed a sample of test reports, photographs and condition reports initiated as a result of the liner inspection performed by NextEra.Corrective action specified for conditions identified were evaluated by the inspectors to assess that the engineering organization was involved in providing evaluation and disposition.

The inspectors confirmed there was no notable damage or indication of leakage identified during the ASME Section Xl Section IWE evaluation.

Examinations I nsoected :. Magnetic particle test (MT) of weld F0104, field weld integrally attached pre-engineered pipe cap to 24 inch carbon steel SW pipe. Work order 1198488, ASME Xl, Code Class 3. MT examination procedure ES 1807'003 R 7 Ch 1.. Liquid penetrant (PT) test of weld F0104, examination of root pass of carbon steel attachment weld of cap to pipe using work document 1198488, ASME Xl, Code Class 3, PT examination procedure ES 1807.002R7 Ch 1.Enclosure 11 Ultrasonic thickness test (UT) of carbon steel SW piping to determine if wall thinning had occurred at various selected circumferential and axial locations of the pipe shown on drawing 1 198488. The pipe wall thickness was measured using UT procedure ES 1807.012 R 5.Visual test (W-2) of reactor pressure vessel bare metal upper head surfaces with attention to the area where control rod drives intersect the head using remote visual techniques using test procedure ES 10-01-23.The inspectors reviewed the steam generator (SG) degradation assessment (DA) to determine that NextEra had reviewed and incorporated the results of the previous outage degradation assessment, operational assessment (OA) and condition monitoring (CM) assessment.

The inspectors reviewed the eddy current test (ECT) procedure, sample plan and data acquired.

lt was noted that no steam generator tubes were identified which specified in-situ pressure testing or specified "plugging".

The inspectors reviewed documentation of reworl</repair activities which specified the development of ASME Section Xl repair plans with the use of welding processes to complete the repair. The work orders (WO) which detail these repair/replacement activities are:. WO 1198488 02 repair of thru wall leak of SW pipe line SW 1814-1-156 and modify support 1814-SG-02 by installation of a pre-engineered pipe cap, Drawing SK-EC145189-2000.

The lnspectors reviewed applicable welding and NDT procedures to determine compliance with the ASME Xl Code requirements.. WO 40055977 01 fabrication of new SW pipe spool and reducer for replacement of existing spoolwhich was degraded (wall thinning).

Pipe spoolwelds F0105, 0106 and 0107 were made using weld procedure ES0815.004.

The inspectors reviewed the fabrication and inspection procedures to verify compliance with the ASME Xl Code requirements.

The inspectors reviewed the replacement material, weld procedure specifications and qualifications, welder qualifications, weld filler metals, non-destructive tests acceptance criteria and post work testing for each activity.

The documents reviewed are listed in the Attachment.

b. Findinos No findings were identified.

1R11 Licensed Operator Requalification

Proqram (71111.11Q - 1 sample).1 Quarterlv Resident Inspector Review

a. Inspection Scope

The inspectors completed one quarterly licensed operator requalification program inspection sample. Specifically, the inspectors observed simulator just-in-time training of licensed operators on April 28,2011 for reactor and steam plant re-start activities.

The inspectors observed formal classroom and simulator activities.

The inspectors examined the operators capability to perform actions associated with high-risk activities, the Enclosure 12 Emergency Plan, previous lessons learned items, and the correct use and implementation of procedures.

The inspectors observed and reviewed the training evaluator's critique of operator performance and verified that deficiencies were adequately identified, discussed and entered into the corrective action program. The inspectors reviewed the simulator's physical fidelity in order to verify similadties between the Seabrook control room and the simulator.

The documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

1R12 Maintenance

Effectiveness (71111.12Q - 2 samples)a. Inspection Scope The inspectors completed two maintenance effectiveness inspection samples. The inspectors reviewed performance-based problems and completed performance and condition history reviews for the selected in-scope structures, systems or components (SSCs) listed below to assess the effectiveness of the maintenance program. Reviews focused on: proper Maintenance Rule (MR) scoping in accordance with 10 CFR 50.65;characterization of reliability issues; tracking system and component unavailability; 10 CFR 50.65 (aX1) and (a)(2) classifications; identifying and addressing common cause failures, trending key parameters, and the appropriateness of performance criteria for SSCs classified (aX2) as well as the adequacy of goals and corrective actions for SSCs classified (aX1). For the periodic assessment inspection sample, the inspectors reviewed the assessment frequency, the performance criteria, the use of operating experience and corrective actions. The inspectors reviewed system health reports, maintenance backlogs, and MR basis documents.

The documents reviewed are listed in the Attachment.. Residual heat removal (RH) system classified as MR (a)(2) with a focus on component performance impacting unavailability and reliability (AR 1647943).. SW system classified as MR (a)(2) with a focus on pipe wall thinning identification and repair (ARs 161 2061, 1637922, 1639537).b. Findinqs No findings were identified.

1 R13 Maintenance Risk Assessments and Emerqent Work Control (71111

.13 - 6 samples)a. Inspection

Scope The inspectors completed six maintenance risk assessment and emergent work control inspection samples. The inspectors reviewed the scheduling and control of planned and emergent work activities in order to evaluate the effect on plant risk. The inspectors conducted interviews with operators, risk analysts, maintenance technicians, and engineers to assess their knowledge of the risk associated with the work, and to ensure Enclosure 13 that other equipment was properly protected.

The inspectors reviewed the availability of opposite train guarded and protected equipment.

The compensatory measures were evaluated against Seabrook procedures, Maintenance Manual 4.14,"TroLtbleshooting," Revision 0 and Work Management Manual 10.1, "On-Line Maintenance," Revision 3.Specific risk assessments were performed using Seabrook's "Safety Monitor", as applicable.

The inspectors reviewed the maintenance items listed below. The documents reviewed are listed in the Attachment.. Risk mitigation actions for orange risk condition associated with reactor head removal on April 5,2011 through April 6, 2011 (WO 1205089).. Planned work associated with SW "A" Train during "B" train pipe replacement and maintenance WO 00626035 for work performed April 14,2011'. Risk mitigation actions due to unplanned entry into orange risk condition due to degraded grid during planned work to remove steam generator nozzle dams on April 22, 201 1 (WO 1203031 ).o Emergent work associated with the heavy lift for the removal'and replacement of"A" train RHR pump (1-RH-P-8A)on May 6,2011 and May 9, 2011 (WO 40083875).

o Emergent work associated with the "A" train SW piping replacement and maintenance for leak downstream of heat exchanger isolation valve SW-V-16 on April 20, 2011 through April 22,2011 (WO 40078357).

o Emergent work associated with the temporary repair of safety injection system check valve SI-V82 which had a body to bonnet leak that was leak sealed with the plant at normal operating temperature and pressure on May 18,2011 through May 22,2011 (WO 40086371).

b. Findinqs No findings were identified.

1R15 Operabilitv

Evaluations (71111.15

• 5 samples)a. Inspection Scope The inspectors completed five operability evaluation inspection samples. The inspectors reviewed operability evaluations and condition reports to verify that identified conditions did not adversely affect safety system operability or overall plant safety. The evaluations were reviewed using criteria specified in NRC Regulatory lssue Summary 2005-20,"Revision to Guidance formerly contained in NRC Generic Letter 91-18, Information to Licensees Regarding two NRC lnspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability

and Inspection Manual Part 9900,"Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety." In addition, where a component was determined to be inoperable, the inspectors verified that TS limiting condition for operation implications were properly addressed.

The documents reviewed are listed in the Attachment.

The inspectors also performed field walk downs and interviewed personnel involved in identifying, evaluating or correcting the identified conditions.

The following items were reviewed:. AR1 662418, operability of the pressurizer code safety relief valve (1-RC-V-1 17) due to seat leakage, June 20, 2011.Enclosure b.14. AR1641413, evaluation of containment shellwith April 20, 2011.. AR1644074, operability of containment enclosure craze cracking in concrete, building with reduced modulus of elasticity, April 21, 2011.. AR1664399, operability of concrete structures with reduced modulus of elasticity, June 27,2011.. AR1664708, operability of B diesel generator with cooling water flow oscillations, June 28,2011 Findinqs NextEra wrote ARs1644074 and 1664399 to document the preliminary laboratory results for concrete core samples taken for the containment enclosure building (CEB) and four other seismic Category I buildings.

Twenty core samples were taken as part of an extent of condition investigation for AR 581434 in which NextEra determined that sections of the below grade concrete walls could be affected by alkali-silica reaction (ASR). Prior NRC review of this area was documented in Inspection Reports 2010-04, 201 0-05, 2011-02 and 2011-07 .Inadequate 50.59 Screeninq for Desiqn Chanqe EC 272057 - AR1664074 NextEra issued EC272057, "Concrete Modulus of Elasticity Evaluation," on April24,2Q11 to address the results of testing that showed a reduction in the concrete modulus of elasticity in the CEB (AR 1644074).

EC272057 also address the reduced modulus in the Control Building/Electric Tunnel CB/ET (AR581434).

The lowest measured modulus was 2.16E+03 ksifor the CEB and 2.1E+03 ksi for the CB/ET, both less than the design value of 3.62E+03 ksi. EC272057 was supported by calculations C-S-1-10150 and C-S-1-10156 which reflected the degraded conditions in the design calculations CD-20-CALC and CE-4-CALC for the control and containment enclosure buildings, respectively.

NextEra concluded the structures remained operable and used EC272057 to disposition the degraded condition as "use-as-is," by incorporating the degraded condition into the design basis. In a safety evaluation screen per 10 CFR 50.59 for EC272Q57, NextEra concluded the change to the facility did not require a complete evaluation per 50.59(cX2)because adequate design margin existed and there was no adverse affect on an UFSAR described design function.The inspectors determined the 50.59 Screen for EC272057 did not correctly address Screen Question 5.a: "Does the proposed activity involve a change to an SSC that adversely affects an IJFSAR design function?

Using the guidance of the Seabrook 1OCFR5059 Resource Manual and NEI 96-07, Revision 1, the inspectors determined that a 50.59 evaluation is specified for changes that adversely affect design function.

ln this situation, the ASR impacted concrete with reduced modulus of elasticity which reduces the flexural capacity of the walls would be an adverse effect. Therefore, NextEra should have evaluated the change to the facility per 10 CFR 50.59(cX2).

The item is unresolved pending action by NextEra to complete a full 50.59 evaluation for EC272057 and subsequent NRC review of that evaluation to determine whether the.1 Enclosure 15 performance deficiency is more than minor. (URl 05000443/2011003-02, lnadequate 50.59 Screening for Design Ghange EC 272057).Effects of Reduced Modulus on Concrete Structures - AR1644074 and 1664399 NextEra's analysis of the CEB samples found that the concrete has acceptable compressive strength and reduced but acceptable modulus of elasticity.

To evaluate the effects of the reduced modulus, NextEra assessed the increase in strain for CEB building elements and found that the strain at the most limiting element remained less than the American Concrete Institute ACI-318 design stress limit and thus was acceptable.

NextEra evaluated the impact on flexural capacity by reviewing the change in bending moment of structural elements.

The reduced modulus causes the concrete to have increased flexure which has the effect of shifting the balance point in how load is transferred between the concrete and the imbedded steel (rebar). The reduced modulus causes a shift toward the reinforced steel in tension. The resultant change in bending moment was evaluated to show that the reduction in capacity was minimal and the stresses on the steel and concrete remain below the design stress limits with margin.NextEra's evaluation of the condition concluded that a change in the dynamic seismic response of the structure would be minor, and the CEB remains capable of performing its design function.The prompt operability determination for the CEB (ARs 1644074 and 1664399)evaluated how the reduced modulus would affect the structure by analysis of locally impacted sections.

The evaluation did not address the effects of reduced modulus on the changes to the natural frequencies of the structure and the global response of the structurelo seismic loads. The inspectors requested further information on the effects of the reduced modulus on stresses and strain in the concrete and rebar system for which NextEra will complete additional analyses.The prompt operability determination (POD) for the Control Building (AR581434)as well as for other ASR impacted structures (AR1664399)evaluated the effects of reduced modulus on portions of the below grade structures and the components housed within them. The evaluations lacked details to explain the effects of the reduced modulus on structuralflexure as related to components attached to the structures, such as pipe supports and cable trays. Similarly, the evaluations lacked details with regard to the structure's response to seismic events as related to structure rigidity and changes in the natural frequency, and the bases to use the ground response spectra. Further, the evaluations lacked details to explain how the function of support anchor bolts would not be adversely impacted by reduced concrete compressive strength in the CBIET.This item is unresolved pending further NRC review of the above issues, action by NextEra to complete additional analysis of the CEB conditions and subsequent NRC Region I review of that analysis, and the completion of reviews by the NRC Office of Nuclear Reactor Regulation specified in the associated task interface agreement (ADAMS No. ML111610530).

The result of these reviews will determine whether there is a performance deficiency associated with this item. (URl 05000443/2011003-03, Operability Evaluation for Degraded Goncrete in ASR Affected Plant Structures).

16 Untimelv Operabilitv Determination - AR 1664399 lntroduction.

The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 6.7.1.a that requires written procedures be established and implemented, including administrative procedures that define authorities and responsibilities for safe operation.

NextEra identified a degraded and nonconforming condition related to reduced modulus of elasticity for buildings housing safety related equipment on May 27,2011, but did not complete an operability assessment until June 27, 2011, when AR1664399 and EC250348 were issued. The delayed entry of the issue into the corrective action process to assess operability was contrary to Section 4.3 of EN-AA-203-1001 that requires an operability determinations (OD) be completed in a time frame commensurate with the safety significance of the issue (in most cases within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).Description.

Procedure EN-AA-203-1001, "OperabiIity Determinations/Functional Assessments," provides requirements for evaluation of degraded conditions and nonconforming conditions and requires:

the Shift Manager (a licensed Senior Reactor Operator)make an OD for each condition that involves equipment issues related to the ability of an SSC to perform its TS function (Section 3.2.1); degraded or nonconforming conditions be entered in to the corrective action program (Section a.2.1); an immediate OD be performed following the discovery of a degraded or nonconforming condition (Section 4.1.7); and, the immediate operability determination shall be completed in a manner commensurate with the safety significance (in most cases during the shift when a concern was generated or 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) and consider all plant conditions (Section 4.3.1).On April 21,2011, NextEra issued AR 1644074 upon receipt of initial test results showing the modulus of elasticity for concrete core samples taken from the containment enclosure building (CEB) was below the American Concrete lnstitute ACI-318 design value 3.62 E+03 ksi. A measured modulus as low as 2.16E+03 ksi (60% of the design value) was a degraded and nonconforming condition with respect to the properties of concrete in Category I structures as described in Section 3.8 of the UFSAR. A reduced modulus impacts the flexural capacity of the impacted walls and thus the function of the building.

The Shift Manager, with input from Engineering, documented the basis for an immediate operability determination for the CEB in AR 1644074. NextEra issued EC250348 and Calculation C-S-1-10156 on April 25, 2011 , which evaluated the integrity of the CEB with consideration of the reduced modulus to disposition the CEB as operable.

The evaluations supporting EC250348 relied upon Calculation C-S-110150, completed for the Control Building on September 23,2010, to show that the reduced modulus had minimal impact on flexure and bending moment capacity of the building walls that are heavily reinforced with steel. NextEra also initiated core sampling in several buildings including - the equipment vault, the emergency feedwater the emergency diesel generator buildings as part of the extent of condition review for the issues identified in the control building.The initial report for the results of the additional testing performed identified reduced modulus of elasticity in all of the buildings in the expanded scope (equipment vault, emergency feedwater and emergency diesel generator buildings).

The information was provided to the responsible engineer and made available in a draft report to NextEra on May 27, 2011, but subject to further review and comment with the vendor for final acceptance.

On June 27, 2011, NextEra issued AR1664399 with an immediate operability determination to address the same reduced modulus condition described in Enclosure

.4 17 AR1644074.

A Prompt Operability Determination (POD) was issued on June 28,2011, to disposition the degraded condition for all impacted buildings which determined the structures were fully operable with margins. The inspectors identified that, although the data for the 3 buitdings was preliminary (final reports were not issued by the vendor until July 1 and 27), NextEra should have initiated a condition report on May 27 ,2011 to establish an immediate operability determination for the buildings since the reduced concrete modulus was a degraded and nonconforming condition as described in UFSAR Section 3.8. The failure to initiate a timely operability determination on May 27 was contrary to Sections 4.2.1 and 4.3.1 of EN-AA-203-1001.

The failure to promptly enter a degraded condition into the corrective actions process to allow the Shift Manager to make timely operability evaluations was a performance deficiency.

Analvsis.

The inspectors determined that the failure to properly implement procedure EN-AA-203-1001 for the degraded and nonconforming condition discussed above was a performance deficiency.

This performance deficiency was considered more than minor based on a comparison with Examples 3.j and 3.k of Appendix E of IMC 0612.Specifically, the performance deficiency was more than minor because a reasonable doubt of operability for the affected concrete structures existed until further engineering evaluations were completed to demonstrate the structures and systems that they housed would remain functional under design and licensing basis conditions.

The finding affected the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events in order to prevent core damage. The issue was evaluated using IMC 0609, "Significance Determination Process" (SDP), and was determined to be of very low safety significance (Green)because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of a barrier function, and was not potentially risk significant for external events. The finding had a cross cutting aspect in the area of problem identification and resolution, P.1(a), because NextEra did not enter identified degraded concrete conditions for several site buildings into the corrective actions process in a timely manner that would have ensured the shift manager completed timely operability evaluations for the affected structures.

Enforcement.

Technical Specification 6.7.1.a, Procedures and Programs, requires that procedures be established and implemented covering administrative procedures that define authorities and responsibilities for safe operation.

Procedure EN-AA-203-1001 defines responsibilities and requirements for r:naking immediate ODs to establish the acceptability of continued plant operation when SSCs are found to be degraded or nonconforming.

Contrary to the above, NextEra did not make an immediate operability determinations to establish the acceptability of continued plant operation when the reduced concrete modulus for several plant structures was identified on May 27,2011.Because this failure to make timely operability determinations is of very low safety significance and was entered into NextEra's Corrective Action Program (CR1673102), this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000443/20 1 1 003-04, U ntimely Operabil ity Determi nation for Degraded Goncrete Structures Housing Safety-Related Equipment)

Inadequate Operabilitv Determination - AR 1664708 lntroduction.

The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 6.7.1.a that requires that written procedures be established and implemented, including administrative procedures that define authorities and Enclosure 18 responsibilities for safe operation.

NextEra identified a degraded condition related to reduced service water (SW) flow to the B emergency diesel generator (EDG) heat exchanger (HX) on June 28, 2Q11, but did not fully evaluate the reduced flow under all plant conditions as required by NextEra procedure EN-AA-203-1001.

Description.

Procedure EN-AA-203-1 00'1, "Operabi l ity Determ inations/Functional Assessments," provides requirements for evaluation of degraded conditions and nonconforming conditions and requires:

the Shift Manager (a licensed Senior Reactor Operator)make an OD for each condition that involves equipment issues related to the abif ity of an SSC to perform its TS function (Section 3.2.1); degraded or nonconforming conditions be entered in to the corrective action program (Section 4.2.1); an immediate OD be performed following the discovery of a degraded or nonconforming condition (Section a.1.7); and, the immediate operability determination shall be completed in a manner commensurate with the safety significance (in most cases during the shift when a concern was generated or 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) and consider all plant conditions (Section 4.3.1).On June 28,2011, NextEra issued AR1664708 when operators observed reduced SW flow through the B EDG heat exchanger during weekly testing of valve SW-V18. The flow initially varied from 500 to 1000 gpm, but increased to 1400 gpm with additional valve strokes. The operators questioned the adequacy of flow indication from SW-FE-6191 , but the operability evaluation documented in AR1664708 accepted the flow with"no operability issues noted" and with plans to monitor the condition.

NextEra observed"normal" flow during a subsequent valve stroke.The operators again observed reduced and erratic SW flows (800-1500 gpm) during the next valve test on July 9, 2011. The B EDG was declared inoperable, AR1667857 was written and a POD per Section 4.3.1 .C of EN-AA-203-1001 was requested at that time.A more detailed investigation and evaluation determined that flow element SW-FE-6191 was partially plugged contributing to variability in flow indication (but not the reduction), and that marine groMh inside the SW pipe (line 1806-1-1 53-16) upstream of diesel heat exchanger DG-E428 was causing intermittent fouling of the heat exchanger tubes and reduced flow. Corrective actions were initiated to address potential heat exchanger fouling in both EDG SW cooling loops. The engineering evaluation associated with the POD confirmed that the "8" EDG and cooling subsystem was and had been fully operable under design basis conditions, including cooling water temperatures at the environmental extremes for operation on the ocean or the cooling tower.The inspectors noted that diesel operating procedure OS 1026.09 requires a minimum of 900 gpm when SW cooling is provided on the ocean, and 1800 gpm when SW cooling is provided on the cooling tower. The inspectors noted further that normal SW flow through the heat exchanger varies from 1500 to 1900 gpm depending on ocean level at the intake. The inspectors determined that the June 28 operability evaluation documented in AR 1664708lacked sufficient basis to explain the reduced flow at 500-1400 gpm, and failed to fully evaluate the EDG cooling function.

The June 28 evaluation did not fully evaluate flow under all plant conditions as specified by Section 4.3.1.A of EN-AA-203-1001, namely, whether flow was adequate relative to the 1800 gpm needed for operation on the cooling tower. NextEra should have fully investigated the flow anomaly on June 28 and evaluated EDG operability under all design basis conditions per Section 4.3.1.C of EN-AA-203-1001 , since reduced SW flow in the range of 500 to 1000 gpm was a degraded condition that impacted engine cooling. When EDG cooling was further investigated on July 9, the presumption on June 28 that the flow anomaly Enclosure 19 was caused by erratic indication was proven wrong. The initial incomplete operability evaluation resulted in the delayed identification, assessment and mitigation of a degraded condition impacting the functionality of the EDG. The failure to promptly and thoroughly evaluate degraded conditions for operability was a performance deficiency.

Analvsis:

The inspectors determined that not properly implementing procedure EN-AA-203-1001 for the degraded condition discussed above was a performance deficiency.

This performance deficiency was considered more than minor based on a comparison with Examples 3.j and 3.k of Appendix E of IMC 0612. Specifically, the performance deficiency was more than minor because a reasonable doubt of operability existed until further engineering evaluations were completed to demonstrate adequate service water flow to the B EDG HX existed and that the B EDG remained functional under design and licensing basis conditions.

The finding affected the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events in order to prevent core damage. The issue was evaluated using IMC 0609, "significance Determination Process" (SDP), and was determined to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not result in an actual loss of safety function, was not a loss of a barrier function, and was not potentially risk significant for external events. The finding had a cross cutting aspect in the area of problem identification and resolution, P.1(c), because NextEra personnel did not adequately evaluate operability to ensure that EDG cooling flow was acceptable under all operating conditions and assure appropriate corrective actions were timely comPleted.

Enforcement.

Technical Specification 6.7.1.a, Procedures and Programs, requires that procedures be established and implemented covering administrative procedures that define authorities and responsibilities for safe operation.

Procedure EN-AA-203-1001 defines responsibilities and requirements for making immediate ODs to establish the acceptability of continued plant operation when SSCs are found to be degraded.Contrary to the above, NextEra did not fully evaluate degraded service water flow on June 28, 2011, resulting in the delayed identification, assessment and correction of a condition that impacted B EDG cooling. Because the finding is of very low safety significance and was entered into NextEra's corrective action program (CR1673102), this violation is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Pol icy. (NCV 05000443/201 1 003-05, I nadequate Operabil ity Determination for Reduced EDG HX Gooling Water Flow)1R18 Plant Modifications (71111.18 - 3 samples).1 Permanent Modification - EC 145280: Proiect 52 SY Upqrade

a. Inspection Scope

The inspectors completed one permanent modification inspection sample. The inspectors reviewed modification package EC145280 that completed changes in the 345 kV switchyard to enhance reliability.

The modifications included the installation of new breakers and bus sections to connect the Seabrook generator and unit auxiliary transformer to the grid. The review was completed to verify that the design bases and performance capability of the system was not degraded.

The inspectors verified the new configuration was accurately reflected in the design documentation, and that the post-modification testing was adequate to ensure the SSCs would function properly.

The Enclosure b..2 a.20 inspectors interviewed plant staff, and reviewed issues entered into the corrective action program to verify that NextEra was effective at identifying and resolving problems associated with temporary modifications.

The documents reviewed are listed in the Attachment.

Findinqs No findings were identified.

Temporarv Modification - EC 272290: lnstall Varistor in Panel for CBA-CP-177 Inspection Scope The inspectors completed one temporary modification inspection sample. The inspectors reviewed modification package EC 272290 associated with operation of the A EDG. The modification installed a varistor in the unit sub panel for CBA-CP-177.

The varistor was installed across relay coils 52X and 52Y to minimize electrical transients when the anti-pump and breaker closing relays operated.

The purpose of the varister is to suppress induced voltages in the emergency power sequencer logic circuits that can negatively affect proper operation of the A emergency diesel generator (CR1645405).

The review was completed to confirm that the design bases and performance capability of the system were not degraded.

The inspectors verified the new configuration was accurately reflected in the design documentation (reference Drawing 310926 Sheet AC4b), and that the poslmodification testing was adequate to ensure that affected SSCs would function properly.

The inspectors also interviewed plant staff, and reviewed issues entered into the corrective action program to verify that NextEra was effective at identifying and resolving problems associated with temporary modifications.

The documents reviewed are listed in the Attachment.

Findinos No findings were identified.

Temporarv Modification - EC 272512: Leak Sealinq of Sl-V82 Inspection Scope The inspectors completed one temporary modification inspection sample. The inspectors reviewed modification package EC272512 that installed a mechanical clamp and seal on safety injection system check valve Sl-V82. The leak seal was installed to minimize gasket leak at the body to bonnet flange identified during plant startup to begin operating cycle 15. The review was completed to confirm that the design bases and performance capability of the system were not degraded.

The inspectors verified the new configuration was accurately reflected in the plant documentation and that the clamp installation and sealing process would not adversely affect the check valve design functions.

The inspectors also interviewed plant staff, and reviewed issues entered into the corrective action program to verify that NextEra was effective at identifying and resolving problems associated with temporary modifications.

The documents reviewed are listed in the Attachment.

b..3 Enclosure 21 b. Findinqs No findings were identified.

1R19 Post-Maintenance

Testinq (71111.19 - 7 samples)a. Inspection Scooe The inspectors completed seven post-maintenance testing (PMT) inspection samples.The inspectors observed portions of PMT activities in the field to verify the tests were performed in accordance with the approved procedures.

The inspectors assessed the test adequacy by comparing the test methodology to the scope of the maintenance work performed.

The inspectors evaluated the test acceptance criteria to verify that the test procedure ensured that the affected systems and components satisfied applicable design, licensing bases and TS requirements.

The inspectors also reviewed recorded test data to confirm all acceptance criteria were satisfied during testing. The documents reviewed are listed in the Attachment.

The activities reviewed are listed below:. Retest of main steam to emergency feedwater pump turbine steam supply valve 1-MS-V-393 on May 18,2A11, following overhaul per WO 40065448.. Retest of chemical and volume control system charging flow control valve 1-CS-FCV-121 on May 18, 2011, following overhaul per WO 1 199620'o Retest of "8" train charging pump 1-CS-P-28 on April 26,2011, following motor replacement per WO 627385.o Retest of reactor coolant loop 1 residual heat removal pump suction isolation valve 1-RC-V-22 on April 11,2Q11, following overhaul per WO 1196480'o Retest of "A" train RHR pump 1-RH-P-8A following replacement on May 16,2011, per WO 40085334.o Retest of "A" emergency diesel generator on May 9, 2011, following failure of the emergency power sequencer during testing per WO 40082703 (CR1645405).. Retest of Sl check valve Sl-V-82 in accordance with OX1 401.04 on May 22, 2011, following leak seal repair per WO40086371.

b. Findinqs No findings were identified.

1R20 Refuelinq

and Outaqe Activities (71111.20 - 1 sample).1 Refuelinq Outaqe OR14

a. Inspection Scope

The inspectors completed one refueling and outage activities inspection sample. The inspectors reviewed the operational, maintenance, and testing activities for the fourteenth refueling outage (OR14) starting on April 1, 2011. The documents reviewed are listed in the Attachment.

22 Review of Outaqe Plan The inspectors reviewed the outage plans to evaluate NextEra's ability to assess and manage the outage risk. The inspectors reviewed the outage risk assessment provided in Engineering Evaluation EE-11-02, "OR14 Outage Schedule Initial Shutdown Risk Review." Monitorinq of Shutdown Activities The inspectors reviewed activities to shut the plant down in accordance with plant procedures.

The inspectors observed completion of various activities specified to place the plant in a cold shutdown condition to assess operator performance, communications, command and control and procedure adherence.

The inspectors reviewed operator adherence to TS specified cooldown limits. The inspectors performed inspection tours of plant areas not normally accessible during plant power operations to verify the integrity of structures, piping and supports, and to confirm that systems appeared functional.

Refuelinq Activities and Reactivitv Control The inspectors verified that refueling activities were performed in accordance with procedures OS1000.09 and RS0721 . The inspectors independently verified on a sampling basis that requirements for core alteration were met. The inspectors observed NextEra actions during core alterations to assure core reactivity was controlled.

The inspectors observed activities from the control room, the reactor cavity and the spent fuel pool at various times. The inspectors verified that fuel movement was tracked in accordance with the fuel movement schedule.

The inspectors verified NextEra action to meet the requirements of TS 3.9 for refueling operations, including the requirements for boron concentration and core monitoring using the source range monitors.

The inspectors observed communications and coordination of activities between the control room and the refueling stations while fuel handling activities were in progress.

The inspectors verified reactivity was controlled in accordance with the requirements of Technical Specification 3.9.Control of Outaoe Risk and Activities The inspectors reviewed daily shutdown risk assessments during OR14 to verify that NextEra addressed the outage impact on defense-in-depth for the critical safety functions:

electrical power availability, inventory control, decay heat removal, reactivity control, and containment.

The inspectors reviewed how NextEra provided defense-in-depth for each safety function and implemented the planned contingencies in order to minimize overall risk where redundancy was limited or not available.

The inspectors periodically reviewed risk updates accounting for schedule changes and unplanned activities.

The inspectors reviewed management controls to manage fatigue by reviewing waiver requests and assessments.

Controlof Heaw Loads The inspectors reviewed NextEra's activities to control the lift of heavy loads in accordance with plant procedures and the commitments to NUREG 0612. The inspectors observed the lift preparations and lift activities to verify adherence to established procedures and controls.

The inspectors used an operating experience smart sample as a reference for this review.Enclosure 23 Clearance Activities and Confiquration Control The inspectors reviewed a sample of risk significant clearance activities and verified tags were properly hung and/or removed, equipment was appropriately configured per the clearance requirement, and that the clearance did not impact equipment credited to meet the shutdown critical safety functions.

Inventorv Control The inspectors reviewed NextEra actions to establish, monitor and maintain the proper water inventory in the reactor during the outage, and in the reactor and spent fuel pool after flooding the reactor cavity for refueling activities.

The inspectors reviewed the plant system flow paths and configurations established for reactor makeup and reactivity control, and verified the configurations were consistent with the outage plan.Reduced Inventorv and Mid-Loop Conditions The inspectors reviewed NextEra's procedures to implement commitments from Generic Letter 88-17 and confirmed that controls for mid-loop operations were in place. The inspectors verified reactor coolant system instrumentation was installed and configured to provide accurate indication.

The inspectors reviewed outage activities that were performed during periods when there was a short time-to-boil to assure adequate controls were in place. Periodically, during the decreased inventory conditions, the inspectors verified that the configurations of the plant systems were in accordance with the commitments.

During reduced inventory operations, the inspectors observed NextEra's control of distractions to assure the operator could maintain the specified reactor vessel level.Foreiqn Material Exclusion The inspectors reviewed the implementation of Seabrook procedures for foreign material exclusion control (FME) for the open reactor vessel, reactor cavity and spent fuel pool.The inspectors reviewed NextEra actions to verify that FME issues were documented and resolved.Electrical Power Thl inspectors verified that the status of electrical systems met TS requirements and the outage risk control plan. The inspectors verified that compensatory measures were implemented when electrical power supplies were impacted by outage work activities and that credited backup power supplies were available.

RHR Svstem Monitorinq The inspectors observed spent fuel pool (SFP) and reactor decay heat removal system status and operating parameters to verify that the cooling systems operated properly.The review included periodic review of SFP and reactor cavity level, temperature, and RHR flow. The inspectors reviewed system status to verify the proper system alignment was established for vessel and cavity level measurement.

Containment Control The inspectors reviewed NextEra activities during the outage to control primary containment closure and integrity, and to prepare the containment for closure prior to plant restart. The inspectors performed tours of all levels in the containment throughout the outage and prior to plant startup per procedure OS101 5.18 to review NextEra's cleanup and demobilization controls in areas where work was completed to assure that tools. materials and debris were removed. This review focused on the control of Enclosure 24 transient combustibles and the removal of debris that could impact the performance of safety systems.Monitorinq Plant Heat up. Approach to Critical and Startup The inspectors observed operator performance during the plant startup activities performed between April 28, 2011 and May 26, 2011. The inspection consisted of control room observations, plant tours and a review of the operator logs, plant computer information, and station procedures.

The inspectors observed pre-job briefs for key evolutions.

The inspectors reviewed the preparations for changes in operating modes.The reactor was taken critical on May 23,2011 at 03:08 a.m., and completed power ascension to 100% FP on May 26, 2011. The inspectors verified, on a sampling basis, that TS, license conditions, and other requirements for mode changes were met. The inspectors verified RCS integrity throughout the restart process by periodically reviewing RCS leakage calculations and by review of systems that monitor conditions inside the containment.

Problem ldentification and Resolution The inspectors reviewed NextEra actions to identify outage related issues and enter them into the corrective action program. This inspection included a review of the corrective actions for Condition Report 1640003. The inspectors reviewed a sample of the corrective actions to verify they were appropriate to resolve the identified issues.b. Findinos No findings were identified.

1R22 Surveillance

Testinq (71111.22 - 7 samples)Inspection Scope The inspectors completed seven surveillance testing inspection samples. The inspectors observed portions of surveillance testing activities for safety-related systems to verify that the system and components were capable of performing their intended safety function, to verify operational readiness, and to ensure compliance with specified TS and surveillance procedures.

The inspectors attended selected pre-evolution briefings, performed system and control room walk downs, observed operators and technicians perform test evolutions, reviewed system parameters, and interviewed the system engineers and field operators.

The test data recorded was compared to procedural and TS requirements, and to prior tests to identify any adverse trends. The documents reviewed are listed in the Attachment.

The following surveillance activities were reviewed:. EX 1804.033, Containment Spray System 10 Year Air Flow Test, April 8,2011 (wo1 209232 11209233).. OX1426.34, Diesel Generator 1A 18 Month Operability Surveillance, April 26,2011, May 2, 2011 and May 1 1, 2011 (WO 40077892).. OX1413.08, Residual Heat Removal Pump 8A Comprehensive Test (lST), April 18,2011 (WO 01203773).. RS1748, Subcritical Physics Testing Using SRWM, May 17 ,2011'. OX1 426.32, Diesel Generator 1B 18 Month Operability Surveillance, April 24,2011 Enclosure 25 through April 25, 2011 (WO 40076902).. EX1803.003, Local Leakage Rate Testing of FP-V-588 and FP-V-592 (LLRT), April 1, 2011 (WO01209198).. EX1803.003, Local Leakage Rate Testing of Penetration X358, Pressurizer Sample Line (LLRT), April 5, 2011 (WO01209191

).The inspectors reviewed deficiencies related to surveillance testing and verified that the issues were entered into the corrective action program. The documents reviewed are listed in the Attachment.

b. Findinqs No findings were identified.

2. RADIATION

SAFEW Cornerstone:

Occupational Radiation Safety 2RS01 Radioloqical Hazard Assessment and Exoosure Controls (71124.01)

a. Inspection Scope

During the period May 9, 2011 through May 12,2011, the inspector performed the following activities to verify that NextEra was evaluating, monitoring, and controlling radiological hazards for work performed during the OR-14 refueling outage in locked high radiation areas (LHRA) and other radiological controlled areas. lmplementation of these controls was reviewed against the criteria contained in 10 CFR Part20, Technical Specifications, and NextEra's procedures.

The documents reviewed are listed in the Attachment.

Radioloqical Hazards Control and Work Coveraqe The inspector identified work performed in radiological controlled areas and evaluated NextEra's assessment of the radiological hazards. The inspector evaluated the survey maps, exposure control evaluations, electronic dosimeter dose/dose rate alarm set points, and radiation work permits (RWP), associated with these areas, to determine if the exposure controls were acceptable.

Specific work activities evaluated included transferring the 8A residual heat removal (RHR) pump into the decay heat vault (RWP 65) and hydrolazing the spent fuel pool (SFP) leak-off lines (RWP 61). For these tasks, the inspector attended the pre-job briefings, reviewed relevant documents, and discussed the job assignments with the workers. Radiation protection technicians were questioned regarding their knowledge of plant radiological conditions for these jobs, and the associated controls.The inspector reviewed the air sample records for samples taken prior to installing steam generator (SG) nozzle dams, to determine if the samples collected were representative of the breathing air zone and analyzed/recorded in accordance with established procedures.

During plant tours, the inspector verified that continuous air monitors were strategically located to assure that potential airborne contamination could be identified in a timely manner and that the monitors were located in low background areas.Enclosure 26 The inspector toured accessible radiological controlled areas located in the primary auxiliary building, fuel handling building, decay heat vaults, and waste processing building.

With the assistance of a radiation protection technician, independent radiation surveys were performed of selected areas to confirm the accuracy of survey data, and the adequacy of postings.Additionally the inspector reviewed the RWPs developed for other work performed during OR-14 including installation of temporary shielding and scaffolding.

ln particular, the inspector reviewed the electronic dosimeter dose/dose rate alarm set points, stated on the RWP, to determine if the set points were consistent with the survey indications and plant policy.lnstructions to Workers By attending pre-job briefings, the inspector determined that workers, performing radiological significant tasks, were properly informed of electronic dosimeter alarm set points, low dose waiting areas, stay times, and work site radiological conditions.

By observing work-in-progress, the inspector determined that stay times were appropriately monitored by supervision to assure no procedural limit was exceeded.

Jobs observed included transferring the 8A RHR pump into the decay heat vault and hydrolazing SFP leak off lines.During plant tours, the inspector determined that locked high radiation areas (LHRA) and a very high radiation area (VHRA) had the appropriate warning signs and were properly secured.The inspector inventoried the keys to LHRAs to determine if the keys were appropriately controlled, as specified by procedure.

The inspector discussed with radiation protection supervision the procedural controls for accessing LHRAs and VHRAs and determined that no changes have been made to reduce the effectiveness and level of worker protection.

Contamination and Radioactive Material Control During plant tours the inspector confirmed that contaminated materials were properly bagged, surveyed/labeled, and segregated from work areas. The inspector observed workers using contamination monitors to determine if various tools/equipment were potentially contaminated and met criteria for releasing the materials from the RCA.Radioloqical Hazards Control and Work Coveraqe By observing preparations for installing the 8A RHR pump and for hydrolazing the SFP leakoff lines, the inspector determined that workers wore the appropriate protective equipment, had dosimetry properly located on their bodies, and were under the positive control of radiation protection personnel.

Supervisory personnel specified the roles and responsibilities of each worker and reviewed the potentialjob hazards to assure that exposure was minimized and that industrial safety measures were implemented.

Radiation Worker Performance During job performance observations, the inspector determined that workers complied with RWP requirements and were aware of radiological conditions at the work site.Additionally, the inspector determined that radiation protection technicians were aware of RWP controls/limits applied to various tasks and provided positive control of workers to reduce the potential of unplanned exposure and personnel contaminations.

27 Problem ldentification and Resolution A review of Nuclear Oversight field observations (OR-14 Daily Quality Summaries)reports, dose/dose rate alarm reports, personnel contamination event reports and associated condition reports, was performed to determine if identified problems and negative performance trends were entered into the corrective action program and evaluated for resolution and to determine if an observable pattern traceable to a similar cause was evident.Relevant condition reports (CR), associated with radiation protection control access and radiological hazard assessment, initiated between January and May 2Q11, were reviewed and discussed with NextEra staff to determine if the follow up activities were being performed in an effective and timely manner, commensurate with their safety significance.

b. Findinqs No findings were identified.

2RS02 Occupational ALARA Plannino and Controls (71124.02)lnspection Scope During the period May 9, 2011, through May 12,2011, the inspector performed the following activities to verify that NextEra was properly implementing operational, engineering, and administrative controls to maintain personnel exposure as low as is reasonably achievable (ALARA) for tasks performed during the OR-14. lmplementation of this program was reviewed against the criteria contained in the 10 CFR ParL 20, applicable industry standards, and NextEra's procedures.

The documents reviewed are listed in the Attachment.

Radioloqical Work Planninq The inspector reviewed pertinent information regarding site cumulative exposure history, current exposure trends, and the ongoing exposure challenges for the outage. The inspector reviewed various OR-14 ALARA plans.The inspector reviewed the exposure status for tasks performed during the outage and compared actual exposure with forecasted estimates contained in various project ALARA plans (AP). In particular, the inspector evaluated the effectiveness of ALARA controls for alljobs that were estimated to exceed 5 person-rem.

These jobs included reactor vessel disassembly/reassembly (AP 11-01), steam generator (S/G) eddy current testing (ECT) (AP 1 1-02), and reactor vessel nozzle walk downs (AP 11-13).The inspector reviewed the ALARA plans and associated Work-ln-Progress (W-l-P)ALARA reviews for those jobs whose actual dose approached the forecasted estimate.Included in this review were the W-l-P's for cavity decontamination, reactor coolant pump seal replacement/motor maintenance, and scaffolding installation.

The inspector evaluated the departmental interfaces between radiation protection, operations, maintenance crafts, and engineering to identify missing ALARA program elements and interface problems.

The evaluation was accomplished by interviewing site Enclosure 28 staff, reviewing outage W-l-P reviews, and reviewing recent station radiation safety committee (RSC) meeting minutes. Included in this review were the actions taken by the RSC to lower outage pro.yect dose goals, as a result of lowering the plant's source term by an effective primary system cleanup.Verification of Dose Estimates The inspector reviewed the assumptions and basis for the OR-14 ALARA forecasted exposure.

The inspector also reviewed the revisions made to various outage project dose estimates due to a reduced source term (i.e., lower dose rates); including reactor disassembly/reassembly activities, reactor coolant pump (RCP) maintenance, and steam generator maintenance.

The inspector evaluated the implementation of the NextEra's procedures associated with monitoring and re-evaluating dose estimates and allocations when the forecasted cumulative exposure for tasks exceeded the actual exposure.

Included in the review were W-l-P reports, that evaluated the effectiveness of ALARA measures, including source term controls, and actions by the RSC to subsequently lower dose goals from the original estimates.

Additionally, the inspector reviewed the exposures for the ten

(10) workers receiving the highest doses for 2Q11 to confirm that no individual exceeded the regulatory limits or performance indicator thresholds.

Source Term Reduction and Control The inspector reviewed the status and historical trends for the source term. Through review of survey maps and interviews with the Radiation Protection Manager, the inspector evaluated recent source term measurements and control strategies.

Specific strategies being employed included use of macro-porous clean up resin, use of submersible ion exchange filters in the reactor cavity, and installation of permanenUtemporary shielding.

The inspector reviewed the effectiveness of temporary shielding by reviewing pre/post installation radiation surveys for selected components having elevated dose rates.Shielding packages reviewed included those placed on the RHR piping, pressurizer spray piping, steam generator penetrations, and RCP piping.Job Site Inspections During plant tours, the inspector assessed the implementation of ALARA controls specified in APs and RWPs, performed during OR-'14. These activities include work on the 8A RHR pump (AP 11-019) and hydrolazing SFP leak off lines. Workers were questioned regarding their knowledge of job site radiological conditions and ALARA measures applied to their tasks.Problem ldentification and Resolution The inspector reviewed elements of NextEra's corrective action program related to implementing the ALARA program to determine if problems were being entered into the program for timely resolution, the comprehensiveness of the cause evaluation, and the effectiveness of the corrective actions. Specifically, recent condition reports related to programmatic dose challenges, personnel contaminations, dose/dose rate alarms, and the effectiveness in predicting and controlling worker exposure were reviewed.Enclosure 29 b. Findinqs No findings were identified.

2RS03 ln-Plant Airborne Radioactivitv Control and Mitioation (7 1 124.03)b.Inspection Scope During the period May 9, 2011 through May 12,2011, the inspector performed the following activities to verify that in-plant airborne concentrations of radioactive materials are being controlled and monitored, and to verify that respiratory protection devices are properly selected and used by qualified personnel.

lmplementation of these programs was evaluated against the criteria contained in 10 CFR Parl20, applicable industry standards, and NextEra's procedures.

The documents reviewed are listed in the Attachment.

Enqineerinq Controls The inspector evaluated the use of portable HEPA ventilation systems installed in various plant areas during the OR-14 outage. The inspector determined that the ventilation systems were located at work locations; e.9., steam generators, and the 8A RHR pump cubicle where airborne contamination could potentially occur. The inspector reviewed testing records for portable HEPA ventilation systems to determine that procedural performance criteria were met.Respiratorv Protection The inspector reviewed the use of respiratory protection devices worn by workers. The inspector reviewed initial radiation survey and air sampling records for S/G nozzle dam installations in the A through D hot and cold legs, associated RWPs, and APs to determine if the use of respiratory protection devices was commensurate with the potential externaldose that may be received when wearing these devices. Additionally, the inspector evaluated the use of respiratory protection; i.e., Delta Suits, for other outage tasks, including cavity decontamination.

Problem ldentification and Resolution The inspector reviewed elements of NextEra's corrective action program related to implementing the airborne monitoring program to determine if problems were being entered into the program for timely resolution, the comprehensiveness of the cause evaluation, and the effectiveness of the corrective actions. Specifically, condition reports related to monitoring challenges, personnel contaminations, dose assessments, and the reliability of monitoring equipment were reviewed.Findinqs No findings were identified.

30 2RS04 Occupational Dose Assessment (71124.04)

Inspection Scope During the period May 9, 2011 through May 12,2011, the inspector performed the following activities to verify the accuracy and operability of personal monitoring equipment and the effectiveness in determining a worker's total effective dose equivalent.

lmplementation of these programs was evaluated against the criteria contained in 10 CFR Part.2O, applicable industry standards, and NextEra's procedures.

The documents reviewed are listed in the Attachment.

External Dosimetrv The inspector verified that NextEra's dosimetry processor was accredited by the National Voluntary Laboratory Accreditation Program (NVLAP). The inspector verified that the approved dosimeter irradiation categories were consistent with the types and energies of the site's source term. The inspector reviewed NextEra's semi-annual quality control evaluation; i.e., TLD blind spiking, of the dosimetry processor.

The inspector confirmed that NextEra has developed "correction factors" to address the response differences of electronic dosimeters as compared to thermoluminescent dosimeters.

Internal Dosimetrv The inspector evaluated the equipment and methods used to assess worker dose resulting from the uptake of radioactive materials.

Included in this review were bioassay procedures, whole body counting equipment (FastScan, Chair counter, portal contamination monitors)calibration checks and operating procedures, and the analytical results for 10 CFR Part 61 samples.The inspector determined that the procedural methods include techniques to distinguish internally deposited radioisotopes from external contamination, methods to assess dose from hard-to-measure radioisotopes, and methods to distinguish ingestion pathways from inhalation pathways.The inspector reviewed the results from two whole body counts to assess the adequacy of the counting time, background radiation contribution, and the nuclide library used for assessing deposition.

No individual exposure exceeded a committed effective dose equivalent (CEDE) of 10 mrem.Special Dosimetric Situations Declared Preqnant Workers The inspector reviewed the procedural controls, and associated records, for managing declared pregnant workers (DPW) and determined that no DPWs were employed during the outage. The inspector reviewed the procedural controls to assure compliance with 10 CFR Part20.Multi-Dosimetrv Methods The inspector reviewed NextEra's procedures for monitoring external dose where significant dose gradients exist at the work site. For OR-14, external effective dose Enclosure b.31 equivalent (EDEX) methods were used to evaluate personnel exposure for installing/removing steam generator nozzle dams. The inspector reviewed the dosimetric results for these jobs. The inspector confirmed that in addition to the TLDs worn, workers also wore electronic dosimeters, equipped with telemetry, to assure that dose fields were promptly monitored by radiation protection technicians.

Problem ldentification and Resolution The inspector reviewed elements of NextEra's corrective action program related to implementing the dosimetry program to determine if problems were being entered into the program for timely resolution, the comprehensiveness of the cause evaluation, and the effectiveness of the corrective actions. Specifically, condition reports related to dose assessments, personnel contaminations, and dose/dose rate alarms were reviewed.Findinos No findings were identified.

OTHER ACTIVITIES

ldentification and Resolution of Problems (71152 - 2 sample)Review of ltems Entered into the Corrective Action Prooram Inspection Scope As specified by Inspection ProcedureTll52, "ldentification and Resolution of Problems," and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the Seabrook corrective action program (CAP). This review was accomplished by accessing NextEra's computerized database.

The documents reviewed are listed in the Attachment.

Findinqs No findings were identified.

Semi-Annual Review to ldentifv Trends lnspection Scope As specified by Inspection Procedure 71152, "Problem ldentification and Resolution," the inspectors performed a semi-annual review of site issues to identify trends that might indicate the existence of more significant safety issues. The inspection included a review of repetitive or closely-related issues documented by NextEra outside of the corrective action program, such as assessment reports, trend reports, performance indicators, major equipment problem lists, system health reports, and maintenance or corrective action program backlogs.

The inspectors reviewed the Seabrook corrective action program database for the first and second quarters of 2011, to assess CRs written in various subject areas (equipment problems, human performance issues, etc.), as well as individual issues identified during the NRCs daily CR review (Section 4OA2.1). The inspectors reviewed the 2011 First Quarter trend reports by the 4.4c.A2.1 a.b..2 a.Enclosure b.32 operations, security and nuclear projects departments, together with the Fourth Quarter 2010 trend report to verify that NextEra was appropriately evaluating and trending adverse conditions in accordance with procedure Pl-AA-207, "Trend Coding and

Analysis.

" Assessment and Observations No findings were identified.

The inspectors did not identify any trends that NextEra had not identified.

The inspectors reviewed a sample of issues and events that occurred over the past two quarters that were documented in the corrective action program. The inspectors verified that NextEra appropriately considered identified issues as emerging trends, and in some cases, verified the adequacy of the actions completed or planned to address the identified trends.NextEra noted the need for continued focus on human performance.

NextEra completed a common cause evaluation for an adverse trend in human performance in Operations (CR594198)with improvements noted in the first quarter of 2Q11. During periodic meetings with station management, the inspectors discussed NRC observations related to human performance.

One example included the inadvertent loss of 345KV Line 394 (CR 1640003) that was caused by a combination of inadequate work package instructions and inadequate worker knowledge of tagout conditions.

Another example included the inadequate performance of a reactor coolant system (RCS) leakage surveillance per Technical Specification 4.6.2.1.e (CR1663219), in which valve RC-V147, whose position is indicated on the main control board, remained closed for thirty

(30) days. While the procedures used for the RCS leakage surveillance could be enhanced, the cause of the issue was the failure to use fundamental operator skills during the performance of routine duties. NextEra corrective actions include a renewed emphasis on operator fundamentals in the operator training program. NextEra continues to address human performance site wide through procedure enhancements, management observations and a focus on procedure compliance in continuing training sessions.NextEra continued to focus on equipment performance and reliability.

Performance problems with secondary plant equipment continue to challenge operators and have resulted in the need to reduce plant power or take the turbine offline three times in three quarters (CRs 591828,1616988,1657622), as reflected in an adverse trend in the NRC Performance indicator for Unplanned Power changes. During periodic meetings with station management, the inspectors discussed emergent equipment issues that impacted safety system availability

[e.9., service water system corrosion (CR1633034), EDG sequencer failure (CR1645405), A RHR pump seal leakage (CR1647943)lor impacted the primary system boundary [e.9., Sl check valve leakage (CR1652573)and safety valve RC-V117 leakage (CR1662418).

NextEra continues to use the preventive maintenance optimization process and the plant health committee reviews of system health reports to focus on equipment issues. Self-assessments have been effective to identify the need for additional actions to address service water system piping degradation (CR 1 637 922).Enclosure

.1 40A5 4046 33 Other Activities (Closed) NRC Temporarv

Instruction 2515/183, "Follow up to the Fukushima Daiichi Nuclear Station Fuel Damage Event" The inspectors assessed the activities and actions taken by NextEra to assess its readiness to respond to an event similar to the Fukushima Daiichi nuclear plant fuel damage event. This included

(i) an assessment of NextEra's capability to mitigate conditions that may result from beyond design basis events, with a particular emphasis on strategies related to the spent fuel pool, as required by NRC Security Order Section 8.5.b issued February 25, 2002, as committed to in severe accident management guidelines, and as specified by 10 CFR 50.54(hh);

(ii) an assessment of NextEra's capability to mitigate station blackout (SBO) conditions, as required by 10 CFR 50.63 and station design bases; (iii) an assessment of NextEra's capability to mitigate internal and external flooding events, as specified by station design bases; and

(iv) an assessment of the thoroughness of the walkdowns and inspections of important equipment needed to mitigate fire and flood events, which were performed by NextEra to identify any potential loss of function of this equipment during seismic events possible for the site. lnspection Report 05000443/201 1009 (ML1 1 1300174) documented detailed results of this inspection activity.(Closed) NRC Temporarv lnstruction 2515/184. "Availabilitv and Readiness Inspection of Severe Accident Manaqement Guidelines (SAMGS)" On May 20,2011, the inspectors completed a review of NextEra's severe accident management guidelines (SAMG), implemented as a voluntary industry initiative in the 1990's, to determine

(i) whether the SAMGs were available and updated,

(ii) whether NextEra had procedures and processes in place to control and update its SAMGS, (iii)the nature and extent of NextEra's training of personnel on the use of SAMGS, and (iv)licensee personnel's familiarity with SAMG implementation.

The results of this review were provided to the NRC task force chartered by the Executive Director for Operations to conduct a near-term evaluation of the need for agency actions following the Fukushima Daiichifuel damage event in Japan. Plant-specific results for Seabrook Station were provided in an Attachment to a memorandum to the Chief, Reactor lnspection Branch, Division of Inspection and Regional Support, dated May 27,2011 (M1111470361).

Meetinqs.

Includinq Exit On July 13,2011, the resident inspectors presented the results of the second quarter routine integrated inspections to Mr. E. Metcalf and Seabrook Station staff. The inspectors also confirmed with NextEra that no proprietary information was reviewed by inspectors during the course of the inspection,.2 ATTACHMENT:

=SUPPLEMENTAL

INFORMATION=

KEY POINTS OF CONTACT

NextEra Personnel J. Ball. Maintenance

Rule Coordinator

K. Boehl, Health Physics Analyst

B. Brown, Supervisor, Civil Engineering

V. Brown, Senior Licensing

Analyst

K. Browne, Operations

Manager

M. Collins, Manager, Design Engineering

W. Cox, Radiological

Engineer

R. Gutherie, Plant System Engineer

F. Haniffy, Senior Radiation

Protection

Analyst

L. Hansen, Plant Engineering

N. Levesque, Plant Engineering

E. Metcalf, Plant General Manager

W. Meyer, Radiation

Protection

Manager

M. O'Keefe, Licensing

Manager

M. Nadeau, System Engineer, Control Building Air Handling

D. Perkins, Supervisor, Radiation

Protection

Technical

Services

M. Scannell, Radiation

Protection

Technical

Specialist

R. Sterritt, ALARA Coordinator

T. Vassallo, Principal

Engineer - Nuclear

J. Walsh, Nuclear Steam Supply System, Supervisor

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened:

05000443/201

1003-02

05000443/201

003-03 Opened and Closed:

05000443/201

1003-01

05000443/2011003-04

05000443/201

003-05 Closed:

05000443/25151183

05000443/25151184

Discussed:

None URI URI NCV NCV NCV lnadequate

50.59 Screening

for Design Change EC 272057 Operability

Evaluation

for Degraded Concrete in ASR Affected Plant Structures

Inadequate

Control of Combustible

Materials U nti mely Operability

Determ i nation for Deg raded Concrete Structures

Housing Safety-Related

Equipment I nadeq uate Operabil ity Determ i nation for Red uced EDG HX Cooling Water Flow Followup to the Fukushima

Daiichi Nuclear Station Fuel Damage Event (Section 4OA5.1)Availability

and Readiness

Inspection

of Severe Accident Management

Guidelines (Section 4OA5.2)TI TI Attachment

LIST OF DOCUMENTS

REVIEWED

Section 1R01: Adverse Weather Protection

OP-AA-1 02-1002, Seasonal Readiness, Revision 0

OAl.42, Operations Department - Severe Weather Plan lmplementation

OS1200.03, Severe Weather Conditions, Revision 18 NM11800, Hazardous Condition Response and Recovery Plan ON1490.09, Summer Readiness surveillance, Revision 5 ON0443.59, Yard Hydrant SemiAnnual Inspection, Revision 5 2011 Summer Readiness Site Certification

SBK 1 1-018, Nuclear Oversight Report - Summer Readiness Condition Report;

AR1655329,

1653764,

1607562 Work Order:

40083993,

40038324,

1384685,

40038436

ODI 61, Redeclaration

/ Joint Owners & NDDO Notification Guideline, Revision 47

ODI 90, 345kV Etectrical Disturbance Communication, Analysis, & Reporting Guideline, Revision 6 051246.02, Degraded VitalAC Power, Revision 10 Seasonal Readiness Review - System Engineering

ER1.1, Classification of Emergencies, Revision 49 Operations Department Turnover Report Daily Status Report Station Operating Logs - various

Section 1R04: Equipment

Aliqnment

OS 1412.09, Rev. 7, PCCW Monthly Flow Check

OX 1412.05, Rev. 8, Monthly PCCW Loop A Valve Verification

D rawi n g s 1 -CC-820205, 1 -CC- B 20206, 1 -CC-820207

UFSAR Section 9.2.2 Cooling for Reactor Auxiliaries

Work Orders

40040600,

40073132 OX1416.01, Monthly Service Water Valve Verification

OX1416.06, Service Water Discharge Valves Quarterly Test and 18 Month Position Verification System Health Report - Service Water System Operations Logs - various PID: 1 -SW-820795, 1 -SW-820794, 1 -NHY-20247

UFSAR Section 9.2,7.3 Technical Specifications

3.7.4 Service Water System/Ultimate Heat Sink Detailed System Text - Service Water System Plant Engineering Action Plan Register Operations Logs - various OS1013.03, Residual Heat Removal System Train A Startup and Operation, Revision2l

OS1013.04, Residual Heat Removal System Train B Startup and Operation, Revision22

OS1001 .1 1, Reactor Coolant System Shutdown Level, Revision 5 OS1016.03, Service Water Train A Operation, Revision 11 OS1016.04, Service Water Train B Operation, Revision 13 OS1016.05, Service Water Cooling Tower Operation, Revision 19 Section 1R05: Fire Protection Fire Protection Pre Fire Strategies Fire lmpairment List Technical Requirement Fire Rated Assemblies Technical Requirement Fire Detection Instrumentation

UFSAR Section 9.5.1 Fire Protection Systems UFSAR Section 13.2.2.9 Fire Protection Personnel OS1200.004, Fire Hazards Analysis for Affected Area I Zone - Appendix A OS1200.00, Response to Fire or Fire Alarm Actuation, Revision 15 NUREG 1805 Chapter 8 FP 2.2, Control of Combustible Materials, Revision 13 (draft)Response to NRC Fire Protection issue 1-SS-CP-1668

Fire Zone W-F-1A, 1B-Z & W-F-5-0 Station Operating Logs - various

Section 1R08: Inservice

Inspection

ES1807.002

Rev 9, Liquid Penetrant Examination - Solvent Removable ES1807.003

Rev 8, Magnetic Particle Examination

ES1807.001

Rev 7 CH 2, Visual Examination

Procedure

for Welding ES03-01 -27 Rev 2, PDI Generic Procedure for Manual Ultrasonic through Wall and Length Sizing of Ultrasonic lndications in Reactor Pressure VesselWelds (PDt-UT-7)

ES10-01-32

Rev 00, Remote lnservice Examination of Reactor Vessel Nozzle to Safe End, Nozzle to Pipe, and Safe End to Pipe Welds Using the Nozzle Scanner (PDl-lSl-254-SE-NB, Rev 1 )ES1807.025

Rev 5, lnservice Inspection (lSl) Visual Examination

Procedure

(W-2)ES1807.012

Rev 6, Ultrasonic Thickness Measurements

MA 10.3 Rev 5, Boric Acid Corrosion Control Program P1-AA-102

Rev 3, Non-Safety Operating Experience Program P1-AA-102-1001

Rev 4, Operating Experience Program Screening and Responding (tncoming)

AR 00569156, 81 Boric Acid Leak in Outlet lsolation Valve Packing Area

AR 01636221, Medium to Heavy Boric Acid Leakage from RHR Pump Suction Packing

AR 00210637, Boric Acid Leak at Packing on Valve FCV121

AR 0021 9427, Boric Acid Leak from Transmitter Fitting

RC-FT-415

AR 00213435, Boric Acid Leak at Packing Charging Header Vent Valve 1-CS-V-836

AR 01640609, Reactor Vessel Hot Leg Post MSIP Exam (158 degree nozzle)Examination Reports

1198488, Liquid Penetrant of

SW-1814 Joint F0104, dwg SKEC145189-2000

1198488, Magnetic Particle Exam of

SW-1814 Joint F0104, dwg SK-EC145189-2000

1 98488, Ultrasonic Examination of SW1 81 4-1 -156-24 Thickness Report 1-SW-1814-001

,!/i'-2 Visual Examination Form, Service Water System

40055977-01, Magnetic Particle Exam Data Sheet (SW) Weld F0105, 106 and 107

01209165, Visual Examinat5ion (W-2) of Pressurizer Heater Sleeves

1208874, Remote Visual (W-2) Examination of RPV Bare Metal Upper Head

SP-SWOL-DS01, Ultrasonic Exam of Pressurizer Spray Nozzle (Phased Array)S-SWOL-DS01, Ultrasonic Exam of Pressurizer Surge Nozzle (Phased Array)Work Orders

WO 01

199620 01, 1-CS-F?V-121

Overhaul Valve Replace Valve Trim

WO 01202400 01,

CS-V-836-B3 (Wet) Boric Acid Leak at Packing

WO 01198488 02, Weld Repair of Salt Service Water Line lnstall Repair Cap

WO 40055977 01, Fabrication of Salt Service Welded Pipe Replacement Spool Piece Work Requests

WR 94002854, Boric Acid Leak at Charging Flow Control Valve

FCV 121 Attachment

WR 940d3420, Charging Header Vent Valve Boric Acid Leak

CS-V-836 WR 94002533.

Fabricate and Install Reducer in Line 1814-01Salt Service Water Weldinq Procedures (WPS) and Procedure Qualification Records (PQR)WPS ES0815.004, Manual Gas tungsten (GTAW) and Shielded Metal Arc (SMAW) Welding of Carbon Steel to Carbon Steel (Pl to Pl )WPS ES0815.004, Manual SMAW of carbon steel to carbon steel PQR SBKI-8'15.004-1

Weld Procedure Qualification Record GTAW/SMAW

of P1 to P1 with Post Weld Heat Treatment (PWHT)PQR SBKI -815.004-2

WPS for P1 to P1 without PWHT

UC 371 & 391, Welder Performance Qualification Record Review to use ES0815.004-1

Drawinos

SK-EC270505-2000, Installation Detail Service Water Piping Repair (SW 1814)SK-EC270505-2001, Fabrication Detail Service Water Piping Repair (SW 1814)Miscellaneous

AR 220564, Self Assessment - Boric Acid Corrosion Control Program 2010 3rd Qtr, Program Health Report - Boric Acid Corrosion Control Program 2010 4th Qtr, Program Health Report - Boric Acid Corrosion Control Program

CR 05-1 1634, Engineering Evaluation for 1-CS-FCV-121

CRO1636130, UT results of SW Piping Indicates Wall Thinning CR (AR 00213435), Boric Acid Corrosion Evaluation (EDl 30560) Valve 1-CS-V-836

CR (AR 210637), Boric Acid Corrosion Control ASME Bolting Evaluation

4-10-2011 MSE#:05-040, Maintenance Support Evalfor Valves

CS-FCV-121

and 1-CS-HCV-182

EC145189, ASME Xl Repair/Replacement Plan Traveler Component

SW-1814

EC 271779 R0, Temp Installation for Repair of Section of SW-1814-001

EDI 30560, Boric Acid Corrosion Evaluation of Valve 1-CS-V-836

Vent Valve SllR, Inservice Inspection Program Plan for 3'o Ten Year lnterval

Section 1R11: Licensed Operator Requalification

Proqram OS1000.02, Plant Startup from Hot Standby to Minimum Load, Revision 20 OS1000.05, Power Increase, Revision 16 OS1000.07, Approach to Critical, Revision 10 OS1007.01, Automatic and Manual Rod Control, Revision 10 OS1056.03, Containment Penetrations, Revision 6 OS1213.01, Loss of RHR While in Reduced Inventory, Revision ON1O31 .02, Starting and Phasing the Turbine Generator, Revision 26 ON1031.13, Post Maintenance Turbine Startup, Revision 12 RS1735, Reactivity Calculations, Revision 4

ODt.101, Guarded Equipment Recommendations for Refueling Outages, Revision 5

ODl.82, Mode Change Notice, Revision 15 Section 1 Rl2: Maintenance Effectiveness System Health Report - RHR system Maintenance Rule Performance and Scope Report UFSAR Section Condition Reports

1612061,

1632409,

1633034,

1636533 PODs for CR1 61 2061 l

16324091

1633034 Drawing 1-SW-820795

OR14 Service Water Inspections

/ Results Attachment

OR14 Service Water Piping Assessment

AR1939781 - SW Train B Pump House Inspection

Work Orders

40080265,

0062557 1 02, 4007 8949 EC272058, SW Pipe Repairs Design Engineering Review for Service Water Pipe

SW-1814 83 Day UT Results

AR1637922 - DQS of Service Water Corrective Actions System Health Reports - Service Water System Plant Engineering Action Register Condition Reports 2010-201 1 Work Requests 2O1O-201 1 Station Operating Logs - various

Section 1R13: Maintenance

Risk and Emerqent Work OR14 Outage Schedule Initial Shutdown Risk Review Rev. 0 OR14 SW Extent of Condition Inspection Matrix

411312011

WM-AA-1000

Work Activity Risk Management Rev. 6

OS 1016.1 1 Contingency Ocean Pump Restoration for SW Work Activities with Ocean Service Water Pumps not in Service. Revision 01 UFSAR 9.2.5 Ultimate Heat Sink Drawings 1 -SW-820795, 1 -SW-B -8.20794 M-Rule a(4), Risk Assessment Reports Station Operating Logs - various AR 1

640932,

1610327,

1639921, 1 631 769,

1631776,

1640932,

1640932 wo 1

199040,

1205038,

1203446,

249348,

00626035 Lift Plan and Rigging Evaluation - A RHR Pump Roof Plug and RHR motor TS - Various RHR leak rate summary Plant Engineering Register EX1801.002, Leakage Reduction Program Surveillance, Revision 9 MS0523.24, Ingersoll-Rand Residual Heat Removal Pump Maintenance, Revision 7 OS1213.02, Loss of RHR while Operating at Reduced lnventory or Midloop Conditions, Revision 12 OS1215.05, Loss of Refueling Cavity Water, Revision 15 OS1213.01, Loss of RHR During Shutdown Cooling, Revision 15 OS1056.03, Containment Penetrations, Revision 6

ODl.103, Conduct of Infrequently Performed Tests or Evolutions, Revision 0

OD1.101, Guarded Equipment Recommendations for Refueling Outages, Revision 5 Work Orders

40086371 Tasks 1, 2, 3 and 4,

WO 1382815 Adverse Condition Monitoring Plan for Sl-V82 dated

618111 Sl-V-82 Operational Decision Making MS0526.09, On Stream Leak Repairs, Revision 4 Insulation Removal Evaluation for Sl-V-82 lN93-90, Unisolatable Reactor Coolant System Leak Following Repeated Applications of Leak Sealant Section 1 Rl5: Operabilitv Evaluations

ODM, Operational Decision Making for

RC-V-117 Leak (AR 1633034)Station Operating Logs - various Adverse Condition Monitoring Plan Plant Engineering Register -IMC ggo0, Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety Attachment Seabrook 1 0CFR5059 Resource Manual

NEI 96-07, Guidelines for 10

CFR 50.59 lmplementation, Revision 1

EN-AA-203-1001 , Operability Determinations

/ Functionality Assessments, Revision 5 Prompt Operability Determinations for

AR 581434,

1664399 Condition Reports

581434,

1641413,

1644074,

1644399,

1629282,

1664708 Calculations

C-S-1-10156, C-S-1-101

50, C-S-1-10155

Design Change EC250348 ,

272057,

CFR 50.59 Screen for EC272057 OD/FA-11-0005, Reduced Concrete Modulus in Below Grade Walls in CEB, EHE EV, EFWPH, B DG FOST Room Section 1R18: Plant Modifications Permanent ModificationEC12T3S, Seabrook Substation Reliability Upgrade Project Permanent Modification

EC145280, Seabrook Substation Reliability Upgrade Project Phase ll Foreign Print

100606 5059 Screen for EC1 45280 EC145280 Procedure and Training Needs Temporary Modification

EC272512,Team Inc Repair for Sl-V-82 Temporary Modification

EC 272290, Install Varistor in Panel for CBA-CP-177

Switchyard

Work Orders

01

384069-12, 01384069-1

EN 10-01 -20,345KV

Bus#6 System 1 Testing

EN 10-01-22,345KV

Switchyard Circuit Load Test LN0561.45, 345KV Bus Primary and Line Cable Differential Relay Testing & Calibration

Condition Report

1640003 Apparent Cause Report - Breaker 294 Open During Switchyard Modifications

SORC Meeting #11-015 Condition Report

1652573, 1

652598

IMC 9900, On-Line leak Sealing Guidelines for ASME Code Class I and 2 Components

EPRI Technical Report

NP-6523-D, On-Line Leak Sealing: A guide for Nuclear Power Plant Maintenance Personnel Boric Acid Corrosion Control ASME

SA 453 Grade 660 Bolting Evaluation for Sl-V-82,511912011

Form

NPV-1 Manufacturer's Data Report

BS-72305-AR6-AR1

for 6" 1655 Swing check Valve, 9t26t78 Westinghouse Certification of Vendor Test Results Valve Studs & Nuts,

9128177 Work Order

40086371 Work Request 91W004461 Team lndustrial Services Work Order 237-05131 Drawing 1 -Sl-D20446, 1 -S l-25 1 - 1 3, D-048 08-837 4D 48 Foreign Print 52914,

100630, Calculation

C-S-1 -45864, Piping Qualification for Leak Repair of Sl-V-82 Calculation

C-S-1-10158, Sl-V-82 Injection Pressure Bolting Evaluation, Revision 0,5119111 Calculation

C-S-1-10158, Sl-V-82 Injection Pressure Bolting Evaluation, Revision 1,5120111

Section 1Rl9: Post Maintenance

Testinq OX1413.01, A Train RHR Quarterly Flow and Valve Stroke Test and 18 Month Valve Stoke Observation, Revision 16 OX1456.86, Operability Testing of IST Pumps, Revision 4 wo

40083875,

1205107,

1205112,

1205043,

40084983,

1203622,

40068999 ,

620087 , 1 1 94007, 12037 97, 12037 98 Attachment

A-B AR 1 65591 0, 1

656350,

1660228, 1

660236,

164187 5,

1642125, 1 631 81 1, 1647 943,

1647949,

1647983,

1646546,

1645417,

1633233,

1649428, EC24938, 27 2291, 27 2303,

0002466 08MSE055 PtD D20662 Technical Specification - various Plant Engineering Action Plan Register -Station Operating Logs - various

WO 40082703 Task 6.40082746

Foreign Prints 31 417, 31 425, 31 61 0 Foreign Print 31919, Emergency Power Sequencing System EPS Logic Drawing 2948-1020, Sheets 1, 3, 4, 6, 10 OX1426.34, Diesel Generator

1A 18 Month Operability Surveillance

OX1401 .04, Reactor Coolant system Pressure lsolation Valve Leakage Rate Tests, Revision 0

DCR 98-039, CBA Replacement Project Condition Report

1645405 - DG A EPS Did Not Fully Sequence Failure Investigation Process for DG A EPS (AR1645405)

Seabrook Train A Emergency Power Sequencer Troubleshooting

Section 1R20: Refuelinq

and Outaqe Activities

Action Request

1640003 Clearances

MTO, 1 -CC-V-1 1 12, 1-CC-V-1092

Control Room Narrative Logs Condition Reports and Repair,

515111 Engineering Evaluation

EE-1 1-02, OR14 Outage Schedule Initial Shutdown Risk Foreign Print98727 - Reactor Vessel Outlet Nozzle DM Weld Flaw Evaluation in Main Control board and MPCS Plant Parameter Displays and Trends MSIP Configuration (AR1 644106), 4121 11 1 Mode Change Report Mode 6 to Mode 5 Mode Change Report Mode 5 to Mode 4 Mode Change Report for Modes 3,2, 1 Open Condition Reports and Actions with Mode Restrictions Operations Component Deviation Log - various dates Outage and Operations Department Turnover Sheets MS0504.15, Reactor Vessel Upper Internals Assembly Installation, Revision 12 MS0504.16, Upper Internals Installation, Revision 11

OD1.101, Guarded Equipment Recommendations for Refueling Outages, Revision 5

ODl.82, Mode Change Notice, Revision 15

OM-AA-O4, OR14 Scope Change Meeting Report

OM-AA-04, Plant Readiness for Operations, Revision 2 ON1031.02, Starting and Phasing the Turbine Generator, Revision 26 ON1031.13, Post Maintenance Turbine Startup, Revision 12

OP-AA-103-1000, Reactivity Management, Revision 0 OR14 Mode Hold / Milestone Report,

412512011 OS1000.02, Plant Startup from Hot Standby to Minimum Load, Revision 20 OS1000.03, Plant Shutdown from Minimum Load to Hot Standby, Revision 18 OS1000.04, Plant Cooldown from Hot Standby to Cold Shutdown, Revision 30 OS1000.05, Power Increase, Revision 16 OS1000.06, Power Decrease, Revision 15 OS1000.07, Approach to Critical, Revision 10 OS1000.09, Refueling Operation, Revision 14 Evaluation.

3118111 the Post Attachment

OS1000.12, Operation with RCS at Reduced Inventory/Midloop Conditions, Revision 9 OS1000.14, Reactor Coolant system Evacuation and Fill, Revision 10 OS1007.01, Automatic and Manual Rod Control, Revision 10 OS1001 .1 1 , Reactor Coolant System Shutdown Level, Revision 5 OS1013.03, Residual Heat Removal System Train A Startup and Operation, Revision 21 OS101 4.02, Operation of Spent Fuel Cooling and Purification System, Revision '15 OS1015.05, FuelTransfer System and Upender Operation, Revision 7 OS1015.07, Spent Fuel Bridge Assembly Operation, Revision 16 OS1015.1 8, Setting Containment Integrity for Mode lV Entry, Revision 6 OS1056.03, Containment Penetrations, Revision 6 OS1213.01, Loss of RHR While in Reduced Inventory, Revision RD0717, Automated

EXCEL Core Offload Tracking, Revision 0 RS0721, Refueling Administrative Control, Revision 9 RS1735, Reactivity Calculations, Revision 4 Technical Specification

3.9 Technical Specification and Commitment Logs

WO 01203652, Containment and Containment Spray Recirculation Sump Surveillance, Slllll Work Order

1209198, Station Operating Logs - various Section 1 R22: Surveillance Testinq

EE 11-003, Containment Spray System Spray Nozzle Test Surveillance Frequency Modification wo

40082703,

40077896, 40078 102,

40077894,

40077892,

40049329,

40049337 , 01 21 051 23,

01209191, 01 2091 90, 01 2091 99, 01 2091 98,

01203722 Calc C-S-1-50006, Specification

9763.006-238-5

Primary Component Cooling Water Pumps

AR 1645405,

DBD-ESF-1, Engineered Safety Features Response Times, Revision 1 Technical Specifications

3.4.6.2.f

,4.5.2.e and 4.0.5 OR14 Local Leak Rate Test Summary,

412812011 RE1707-B-R, Shutdown Margin Verification ,

5117111 Subcritical Rod Worth Measurement Data Analysis System Results

511712011 Westinghouse Letter

NAH-1 1-42, Cycle 15 Subcritical Physics Testing,

611612011 Westinghouse Letter

LTR-NRC-O8-13, SER Compliance with

WCAP-16260-P-A, 4115l20OB

WCAP-16260-P-A, The Spatially Corrected Inverse Count Rate (SCICR) Method for Subcritical Reactivity Measurement, September

2005 Station Operating Logs - various Section 2RS01: Radioloqical Hazard Assessment and Exposure Gontrols HD0958.03, Personnel Survey and Decontamination Techniques

HD0958.04, Posting of Radiologically Controlled Areas HD0958.17, Performance of Routine Radiological Surveys HN0958.25, High Radiation Area Controls HD0958.30, Inventory and Control of Locked or Very High Radiation Area Keys and Locksets Condition Reports

1638564,

1640938,

1644445,

1640938,

1626367 ,

1612661 ,

1640268,

1650347 ,

1618932,

1623142,

1629512,

1649794,

1604791,

1642643, 1

626363, 1

639705, 1 651 585,

1651072, 1 651 584 Attachment Section 2RS02: Occupational

ALARA Planninq and Controls

RP-AA-1 04, ALARA Program

RP-AA-1 04-1 000, ALARA lmplementing

Procedure

RP-AA-1 01-2004, Method for Monitoring and Assigning Effective Dose Equivalent for High Dose Gradient Work Condition Reports: See Section 2RS01 Section 2RS03: In-Plant Airborne Radioactivitv Gontrol and Mitiqation

HD0955.01, Analysis of Smears and Air Samples HD0958.01, Air Sampling HD0965.12, Respiratory Protection lssue and Use Condition Reports: See Section 2RS01 Section 2RS04: Occupational Dose Assessment

HD0955.54, Operation of the TSA Model

SPM-906 Portal Monitor HD0955.62, Use of the Argos 4AlB HD0958.1 9, Evaluation of Dosimetry Abnormalities

HD0958.27, Dose Assessment for Personnel Contaminations

HN0958.39, Multi-Badge Control & Exposure Tracking HD0958.41, Blind Spiking of TLDs HD0958.42, Determination and Controlof Dose to an Embryo/Fetus

HD0958.49, Response Protocols for Whole Body Counting and Personnel Contamination Monitoring

HD0961.29, Internal Dosimetry Assessment

HD0963.28, Calibration and Troubleshooting of MGP Instruments

DMC 2000 Dosimeters

HD0992.02, lssuance and Control of Personnel Monitoring Devices

RP-AA-1 01-2004, Method for Monitoring and Assigning Effective Dose Equivalent for High Dose Gradient Work Condition Reports: See Section 2RS01 Miscellaneous Documents:

NVLAP Certification Records, Personnel Dosimetry Performance Testing Annual Review Report of the 2010 10 CFR Part 61 Radionuclide Analysis Electronic Dosimeter Dose/Dose Rate Alarm Reports, January - May 2011 Top Ten Individual Exposure Records for 2Q11 Portable HEPA Inventory

& Test Records EPRI Standard Radiation Monitoring Program Data Summary for primary piping Reactor Coolant System

OR-14 Clean Up Data Nuclear Oversight Field Observation

OR-14 Daily Quality Summary Reports HPSTID 09-01 1, Use of Effective Dose Equivalent for Steam Generator Nozzle Dam Work HPSTID 08-13, Calibration of the FastScan WBC System

OR-14 ALARA Plans (AP)fuVork-ln-Prooress (WlP) Reviews: AR 1 1-01, reactor vessel disassembly/re-assembly

AR 1 1-02, steam generator (S/G) eddy current testing/tube plugging

AR 11-03, S/G secondary side maintenance

AP 1 1-11, scaffolding Installation/Removal

AP 1 1-13, reactor vessel bare metal visual inspections

ACI ADAMS ALARA AMS AP AR ASME ASR BACC CAP CB/ET CEB CEDE CR DG DPW ECT EDEX EDG EFW FBL FHB FPP GTAW HEPA IMC IP tsl LHRA MR MSIP MT NCV NDE NFPA NRC NVLAP OR OD ODs OM PAB PARS PCCW PDI PMT POD

LIST OF ACRONYMS

American Concrete Institute Agency-wide

Documents

Access and Management

System As Low As is Reasonably

Achievable

Airborne Monitoring

System ALARA Plans Action Request American Society of Mechanical

Engineers Alkali-silica

Reaction Boric Acid Corrosion

Control (Program)Corrective

Action Program Control Building/Electric

Tunnel Containment

Building Committed

Effective

Dose Equivalent

Condition

Report Diesel Generator Declared Pregnant Workers Eddy Current Testing External Effective

Dose Equivalent

Emergency

Diesel Generator Emergency

Feedwater Fire Brigade Leader Fuel Handling Building Fire Protection

Program Gas Tungsten Arc Welding High Efficiency

Particulate

Air Inspection

Manual Chapter Inspection

Procedure In-service

Inspection

Locked High Radiation

Areas Maintenance

Rule Mechanical

Stress lmprovement

Process Magnetic Particle Test Non-cited

Violation Non-Destructive

Examination

National Fire Protection

Association

U.S. Nuclear Regulatory

Commission

National Voluntary

Laboratory

Accreditation

Program Outage for Refueling Operability

Deficiency

Operability

Determinations

Operations

Management

Primary Auxiliary

Building Publicly Available

Records Primary Component

Cooling Water Performance

Demonstration

Initiative

Post-maintenance

Testing Prompt Operability

Determination

PQR [[]]

PT A-12 Procedure

Qualification

Record Penetrant

Test Pressurized

Water Reactor Reactor Coolant Pump Reactor Coolant System Residual Heat Removal Reactor Pressure Vessel Radiation

Safety Committee Radiation

Work Permit Severe Accident Management

Guidelines

Station Blackout Significance

Determination

Process Spent Fuel Pool Steam Generator Shift Manager Shielded Metal Arc Welding Spent Fuel Pool Scintillation

Portal Monitor Subcritical

Rod Worth Measurement

Structures, Systems or Components

Service Water Service Water Pump Temporary

Instruction

Thermolum

inescent Dosimeter Technical

Specifications

Updated Final Safety Analysis Report Ultrasonic

Testing Very High Radiation

Area Visual Test Work-ln-Progress

Work Order Weld Procedure

Specification

Work Request

PWR [[]]

RCP [[]]

RCS [[]]

RHR [[]]

RPV [[]]

RSC [[]]

RWP [[]]

SAMG [[]]

SBO [[]]

SDP [[]]

SFP [[]]

SG [[]]

SM [[]]

SMAW [[]]

SPF [[]]

SPM [[]]

SRWM [[]]

SSC [[]]

SW [[]]

SWP [[]]

TI [[]]

TLD [[]]

TS [[]]

UFSAR [[]]

UT [[]]

VHRA [[]]

VT W-I-P

WO [[]]

WPS [[]]

WR Attachment