IR 05000445/2007006

IR 05000445-07-006; on 01/01/2007-05/21/2007; for Comanche Peak Steam Electric Station, Unit 1. Integrated Resident and Regional Report of Steam Generator and Reactor Vessel Closure Head Replacement Activities

ML071880010
Person / Time
Site:	Comanche Peak
Issue date:	07/06/2007
From:	Johnson C E NRC/RGN-IV/DRP/RPB-A
To:	Blevins M TXU Power
References
IR-07-006
Download: ML071880010 (43)

Text

July 6, 2007

Mike Blevins, Senior Vice President and Chief Nuclear Officer TXU Power ATTN: Regulatory Affairs Comanche Peak Steam Electric Station P.O. Box 1002 Glen Rose, TX 76043

SUBJECT: COMANCHE PEAK STEAM ELECTRIC STATION - NRC INTEGRATEDINSPECTION REPORT 05000445/2007006

Dear Mr. Blevins:

On May 21, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection atyour Comanche Peak Steam Electric Station, Unit 1, facility. No inspection of Unit 2 was performed under this report number. This inspection was conducted due to the Unit 1 steam generator and reactor vessel head replacement activities. The enclosed inspection report documents the inspection findings, which were discussed on May 21, 2007, with Mr. R. Floresand other members of your staff.This inspection examined activities conducted under your licenses as they related to safety andcompliance 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. This inspection covers steam generator and reactor vessel head replacement activities.Based on the results of this inspection, no findings of significance were identified.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and itsenclosure will be made available electronically for public inspection in the NRC PublicDocument 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.gov/reading-rm/adams.html (the Public Electronic Reading Room).Should you have any questions concerning this inspection, we will be pleased to discuss themwith you.

Sincerely,/RA/Claude Johnson, Chief Project Branch A Division of Reactor Projects TXU Power-2-Docket Nos.:50-445License Nos.:NPF-87

Enclosure:

NRC Inspection Report 05000445/2007006

w/Attachment:

Supplemental Informationcc w/

Enclosure:

Fred W. Madden, Director Regulatory Affairs TXU Power P.O. Box 1002 Glen Rose, TX 76043George L. Edgar, Esq.Morgan Lewis 1111 Pennsylvania Avenue, NW Washington, DC 20004Terry Parks, Chief InspectorTexas Department of Licensing and Regulation Boiler Program P.O. Box 12157 Austin, TX 78711The Honorable Walter MaynardSomervell County Judge P.O. Box 851 Glen Rose, TX 76043Richard A. Ratliff, ChiefBureau of Radiation Control Texas Department of Health 1100 West 49th Street Austin, TX 78756-3189Environmental and Natural Resources Policy Director Office of the Governor P.O. Box 12428 Austin, TX 78711-3189 TXU Power-3-Brian AlmonPublic Utility Commission William B. Travis Building P.O. Box 13326 Austin, TX 78711-3326Susan M. JablonskiOffice of Permitting, Remediation and Registration Texas Commission on Environmental Quality MC-122 P.O. Box 13087 Austin, TX 78711-3087Lisa R. Hammond, ChiefTechnological Hazards Branch National Preparedness Division FEMA Region VI 800 N. Loop 288 Denton, TX 76209 TXU Power-4-Electronic distribution by RIV:Regional Administrator (BSM1)DRP Director (ATH)DRS Director (RJC1)DRS Deputy Director (WBJ)Senior Resident Inspector (DBA)Branch Chief, DRP/A (CEJ1)Senior Project Engineer, DRP/A (TRF)Team Leader, DRP/TSS (CJP)RITS Coordinator (MSH3)DRS STA (DAP)M. Kunowski, OEDO RIV Coordinator (MAK3)ROPreports CP Site Secretary (ESS)SUNSI Review Completed: __CEJ_ADAMS: Yes G No Initials: _CEJ_____ Publicly Available G Non-Publicly Available G Sensitive Non-SensitiveC:\FileNet\ML071880010.wpdRIV:RI:DRP/ASRI:DRP/AC:DRS/EBC:DRS/OBC:DRS/PEBC:DRS/PSBAASanchezDBAllenDAPowersATGodyLJSmithMPShannonE-CEJohnsonE-CEJohnson/RA//RA//RA//RA/7/2/077/2/076/22/076/20/076/21/076/18/07C:DRP/A CEJohnson/RA/7/6/07OFFICIAL RECORD COPYT=Telephone E=E-mail F=Fax U.S. NUCLEAR REGULATORY COMMISSIONREGION IVDockets:50-445Licenses:NPF-87 Report: 05000445/2007006 Licensee:TXU Generation Company LP Facility:Comanche Peak Steam Electric Station, Unit 1 Location:FM-56, Glen Rose, Texas Dates:January 1, 2007 through May 21, 2007Inspectors: D. Allen, Senior Resident InspectorA. Sanchez, Resident Inspector E. Owen, Reactor Inspector, Engineering Branch 1 R. Kopriva, Senior Reactor Inspector, Engineering Branch 1 W. Sifre, Senior Reactor Inspector, Engineering Branch 1 R. Azua, Reactor Inspector, Engineering Branch 1 S. Rutenkroger, Reactor Inspector, Engineering Branch 1 Gilbert L. Guerra, CHP, Health Physicist Donald L. Stearns, Health PhysicistApproved by:Claude Johnson, Chief, Project Branch ADivision of Reactor Projects

Attachment:

Supplemental Information Enclosure-2-

SUMMARY OF FINDINGS

IR 05000445/2007006; 01/01/2007-05/21/2007; Comanche Peak Steam Electric Station, Unit 1. Integrated Resident and Regional Report of Steam Generator and Reactor Vessel Closure

Head Replacement Activities.This report covered a 5-month period of inspection by two resident inspectors, five regionalreactor inspectors and two health physicists. No findings were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using the Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the Significance Determination Process 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 3, dated July 2000.A.

NRC-Identified and Self-Revealing Findings

No findings of significance were identified.

B.Licensee-Identified Violations

None.

Enclosure-4-

REPORT DETAILS

Summary of Plant StatusComanche Peak Steam Electric Station (CPSES) Unit 1 began the period operating atessentially 100 percent power. On February 16, 2007 Unit 1 began a reactor power coastdown.

On February 24, at 12:00 noon, Unit 1 entered Mode 3 to begin the steam generator and reactor vessel head replacement outage, 1RF12. On April 20, the Unit 1 replacement outage ended when the main generator beakers were closed. Unit 1 achieved 100 percent power on April 24. On April 27, reactor power was reduced to approximate 80 percent power for final testing. Unit 1 returned to 100 percent power on April 28 and remained at essentially 100 percent power for the rest of the reporting period.1.REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R02 Evaluations of Changes, Tests, or Experiments

a. Inspection Scope

The inspectors reviewed the effectiveness of the licensee's implementation of changesto the facility structures, systems, and components; risk-significant normal and emergency operating procedures; test programs; and the updated final safety analysis report in accordance with 10 CFR 50.59, "Changes, Tests, and Experiments." The inspectors utilized Inspection Procedure 71111.02, "Evaluation of Changes, Tests, or Experiments," for this inspection.The inspectors reviewed one safety evaluation performed by the licensee since the lastNRC inspection of this area at CPSES, Unit 1. The evaluation was reviewed to verify that licensee personnel had appropriately considered the conditions under which the licensee may make changes to the facility or procedures or conduct tests or experiments without prior NRC approval. The inspectors reviewed two licensee-performed applicability determinations in which licensee personnel determined that evaluations were not required, to ensure that the exclusion of a full evaluation was consistent with the requirements of 10 CFR 50.59. Evaluations and applicability determinations reviewed are listed in the attachment to this report.The inspectors reviewed and evaluated a sample of recent licensee condition reports todetermine whether the licensee had identified problems related to 50.59 evaluations, entered them into the corrective action program, and resolved technical concerns and regulatory requirements. The reviewed condition reports are identified in the

.The inspection procedure specifies a required minimum sample of six licensee safetyevaluations and 12 applicability determinations and screenings (combined). The

-5-inspectors completed review of one licensee safety evaluation and 2 applicabilitydeterminations for this effort. The remaining required samples are documented in NRC Inspection Report 05000445;446/2007002, Section 1R02.

b. Findings

No findings of significance were identified.

1R08 Inservice Inspection Activities (71111.08).1Performance of Nondestructive Examination Activities Other Than Steam GeneratorTube Inspections, Pressurized Water Reactor Vessel Upper Head PenetrationInspections, Boric Acid Corrosion Control

a. Inspection Scope

The inspectors used Inspection Procedure 71111.08, "Inservice Inspection Activities,"for this inspection. The inspection procedure requires the review of Nondestructive Examination (NDE) activities consisting of two or three different types (i.e., volumetric, surface, or visual). The inspectors observed and reviewed the performance of radiographic and ultrasonic examinations (volumetric) of welds on the Unit 1 new steam generator (3 and 4) to the reactor coolant loops (hot and cold legs), and auxiliary feedwater piping (FW-TUX-42, 36, 38 and 7). Additionally, the inspectors observed dye penetrant and magnetic particle examinations of welds (surface) on new steam generator No. 4 to reactor coolant loop welds (hot and cold legs) and auxiliary feedwater piping (FW-TUX-7) respectively. In addition, the inspectors observed four visual (VT-1 and VT-3) examinations performed on component supports. The table below identifies the above examinations which were conducted using five methods and three examination types.System/ComponentIdentityExaminationTypeExaminationMethodReactor CoolantSystemNew Steam Generator (#3)to Hot Leg WeldVolumetricUltrasonicReactor CoolantSystemNew Steam Generator (#3)to Cold Leg WeldVolumetricUltrasonicAuxiliary FeedwaterSystemCap Weld FW-TUX-42VolumetricUltrasonicRadiographyAuxiliary FeedwaterSystemCap Weld FW-TUX-36VolumetricRadiographyAuxiliary FeedwaterSystemCap Weld FW-TUX-38VolumetricRadiographyAuxiliary FeedwaterSystemCap Weld FW-TUX-7VolumetricUltrasonic System/ComponentIdentityExaminationTypeExaminationMethodEnclosure-6-Reactor CoolantSystemNew Steam Generator (#4)to Hot Leg WeldVolumetricRadiographyReactor CoolantSystemNew Steam Generator (#4)to Cold Leg WeldVolumetricRadiographyReactor CoolantSystemNew Steam Generator (#4)to Hot Leg WeldSurfacePenetrantReactor CoolantSystemNew Steam Generator (#4)to Cold Leg WeldSurfacePenetrantAuxiliary FeedwaterSystemCap Weld FW-TUX-7SurfaceMagneticParticleComponent CoolingWater System Vertical Spring CanH1: CC-1-RB-049VisualVisual (VT-3)Component CoolingWater System Welded AttachmentH1WA: CC-1-RB-049VisualVisual (VT-1)Component CoolingWater System Vertical Spring CanH1: CC-1-249-701-C53AVisualVisual (VT-3)For each of the observed nondestructive examination activities, the inspectors verifiedthat the examinations were performed in accordance with the specific site procedures and the applicable American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code requirements.During review of each examination, the inspectors verified that appropriatenondestructive examination procedures were used, examinations and conditions were as specified in the procedure, and test instrumentation or equipment was properly calibrated and within the allowable calibration period. The inspectors also verified the nondestructive examination certifications of the personnel who performed the above volumetric, surface, and visual examinations. Finally, the inspectors observed that indications identified during the ultrasonic, radiographic, and visual examinations weredispositioned in accordance with the ASME-qualified nondestructive examination procedures used to perform the examinations.The inspection procedure requires review of one or two examinations with recordableindications that were accepted for continued service to ensure that the disposition was made in accordance with the ASME Code. The inspectors verified that two laminar flaws discovered on the original dissimilar metal welds of the Pressurizer Safety Valve B line (TBX-1-4501-12OL and TBX-1-4501-13OL) were acceptable in accordance with the standards of the ASME Code.

-7-The inspection procedure further requires verification of one to three welds on Class 1or 2 pressure boundary piping to ensure that the welding process and welding examinations were performed in accordance with the ASME Code. The inspectors verified through observation and record review that the auxiliary feedwater pipe cap welds (FW-TUX-42, 38 and 7) and the welding that was performed on the Unit 1 nuclear steam supply system to join the new steam generators (3 and 4) to their associated reactor coolant loops, in the field, were performed in accordance with Sections IX and XI of the, 1998 Edition of the ASME Code. This included review of welding material issue slips to establish that the appropriate welding materials had been used and verification that the welding procedure specification had been properly qualified. The inspectors completed one sample.

b. Findings

No findings of significance were identified..2Reactor Vessel Upper Head Penetration Inspection Activities

a. Inspection Scope

The inspection requirements for this section parallel the inspection requirement steps inSection 02.01. The inspectors reviewed records of completed nondestructive examinations, including the eddy current and ultrasonic examination data analysesprocess used on the reactor vessel upper head penetrations during their preservice inspections.Additionally, the nondestructive examination procedures used to perform the aboveexaminations were reviewed to assure that they were consistent with ASME Code requirements, and the equipment and calibration requirements were appropriately identified and demonstrated. The inspectors completed one sample.

b. Findings

No findings of significance were identified..3Boric Acid Corrosion Control Inspection Activities (Pressurized Water Reactors)

a. Inspection Scope

The inspectors evaluated the implementation of the licensee's boric acid corrosioncontrol program for monitoring degradation of those systems that could be deleteriously affected by boric acid corrosion.The inspection procedure requires review of a sample of boric acid corrosion controlwalkdown visual examination activities through either direct observation or record review. The inspectors reviewed the documentation associated with the licensee's boric

-8-acid corrosion control walkdown, as specified in Station Administrative Manual (STA)Procedure STA-737, "Boric Acid Corrosion Detection and Evaluation," Revision 4.

Samples of documented visual inspection records of inspection walkdowns performed on components and equipment during the previous Refueling Outage 1RF11, and this refueling outage, were reviewed by the inspectors. Additionally, the inspectors performed independent observations of piping containingboric acid during walkdowns of the containment building and the auxiliary building. The inspection procedure requires verification that visual inspections emphasizelocations where boric acid leaks can cause degradation of safety significant components. The inspectors verified through direct observation and program/record review that the licensee's boric acid corrosion control inspection efforts are directed towards locations where boric acid leaks can cause degradation of safety-related components.

The inspection procedure requires both a review of one to three engineering evaluationsperformed for boric acid leaks found on reactor coolant system piping and components, and one to three corrective actions performed for identified boric acid leaks. There were no applicable corrective action documents generated since the last inspection period that required formal engineering evaluation (e.g., that resulted in a separate design or structural engineering analysis to determine continued operability). The inspectors reviewed Smart Forms (SMF) documenting minor valve packing leaks on valves in the safety injection system. The planned corrective actions were adequate in each case.

The inspectors completed one sample.

b. Findings

No findings of significance were identified..4Steam Generator Tube Inspection Activities

a. Inspection Scope

The inspectors verified through records review that licensee personnel and contractorsused properly qualified eddy current probes and equipment for the expected types of tube degradation to assure proper identification and evaluation of indications for the new baseline data. The inspectors verified that the licensee analysts reviewed the areas of potential degradation, based on site-specific and industry experience, to assure proper use of this information. The inspectors reviewed the repair criteria used to assure compliance with technical requirements. The inspectors also verified the licensee's eddy current examination scope and expansion criteria met the Technical Specifications, industry guidelines, and commitments to the NRC.Regarding plugging and in-situ pressure testing, because the steam generators werenew replacement components, the licensee had no need for plugging and in-situ pressure testing onsite. The vendor had plugged one tube in Steam Generator No. 3 prior to its delivery onsite due to a tube bulge in the tubesheet region during fabrication.

-9-The inspectors completed one sample.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation (71111.13)

a. Inspection Scope

The inspectors reviewed selected activities regarding risk evaluations and overall plantconfiguration control. The inspectors discussed emergent work issues with work control personnel and reviewed the potential risk impact of these activities to verify that the work was adequately planned, controlled, and executed. The activities reviewed were associated with:Probability Risk Analysis Report related to the multiple crane operations insidethe Unit 1 containment building during 1RF12, on February 23, 2007Defense in depth contingency plan, 1RF-22, for maintaining Unit 1 containmentpressure while the containment liner is removed and fuel is being unloaded with24 or less fuel assemblies remaining in the core, on February 26, 2007The inspectors completed two samples.

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications (71111.17B)

a. Inspection Scope

The inspectors reviewed eleven permanent plant modification packages and associateddocumentation, such as implementation reviews, safety evaluation applicability determinations, and screenings, to verify that they were performed in accordance with regulatory requirements and plant procedures. The inspectors also reviewed the procedures governing plant modifications to evaluate the effectiveness of the program for implementing modifications to risk-significant systems, structures, and components, such that these changes did not adversely affect the design and licensing basis of the facility. Procedures and permanent plant modifications reviewed are listed in the to this report. Further, the inspectors interviewed the cognizant design and system engineers for the identified modifications as to their understanding of the modification packages and process.

-10-The inspectors evaluated the effectiveness of the licensee's corrective action process toidentify and correct problems concerning the performance of permanent plant modifications by reviewing a sample of related condition reports. The reviewed condition reports are identified in the Attachment.The inspection procedure specifies inspector-review of a required minimum sample ofsix permanent plant modifications. The inspectors completed review of eleven permanent plant modifications.

b. Findings

No findings of significance were identified.

1R19 Postmaintenance Testing (71111.19).1Steam Generator and Reactor Vessel Head Replacement

a. Inspection Scope

The inspectors witnessed or reviewed the results of the postmaintenance tests for thefollowing replacement outage activities:Control rod drive mechanism (CRDM) ventilation testing following the completeredesign and replacement of the old Unit 1 CRDM ventilation fan system, in accordance with Integrated Plant Operating Procedures Manual (IPO) IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement,"

Revision 0, reviewed on April 19, 2007CRDM testing following reactor vessel head replacement, in accordance withprocedure IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement," Revision 0, reviewed on April 19, 2007Steam generator blowdown system flow and vibration testing following thereplacement of steam generators, in accordance with procedure IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement,"

Revision 0, observed and reviewed on April 20, 2007Transfer of feedwater bypass control to main feedwater control testing followingmaintenance and tuning activities, in accordance with procedure IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement,"

Revision 0, observed and reviewed on April 20, 2007Electrical load swing testing to ensure reactor control system interaction andtuning following the replacement of the Unit 1 steam generators, in accordance with procedure IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement," Revision 0, observed and reviewed on April 20, 2007

-11-Steam generator steam flow calibration following replacement of steamgenerators, in accordance with procedure IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement," Revision 0, observed and reviewed on April 24, 2007Steam Generator Water Level Control System response testing followingadjustments and tuning activities, in accordance with procedure IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement,"

Revision 0, observed and reviewed on April 30, 2007Large load (275 MWe) reduction test following replacement outage activities, inaccordance with procedure IPO-011A, "Plant Restart and Testing Following Steam Generator Replacement," Revision 0, observed and reviewed on April 30, 2007Reactor coolant system flow measurement test following the replacement of theUnit 1 steam generators, in accordance with procedure number INC-7018A, "Reactor Coolant System Flow Measurement," Revision 3, reviewed on May 2, 2007In each case, the associated work orders and test procedures were reviewed inaccordance with the inspection procedure to determine the scope of the maintenance activity and to determine if the testing was adequate to verify equipment operability.

The inspectors also reviewed Chapter 14, "Initial Test Program" of Updated Final Safety Analysis Report to help determine the adequacy of the testing.The inspectors completed nine samples.

b. Findings

No findings of significance were identified..2Containment Alternate AccessContainment Integrated Leak Rate Test Procedure Review (70307)

a. Inspection Scope

The inspectors reviewed the licensee's containment integrated leak rate test procedureto verify that the test complies with regulatory requirements, guidance, and licensee commitments to evaluate the technical adequacy to determine containment leak tight integrity. The inspectors ensured that the procedure contained sufficiently detailed guidance for:

(1) the alignment and operation of all systems and equipment inside and penetrating containment,

(2) inspections of the accessible portions of containment,

(3) verification of equipment calibration, and

(4) appropriate success criteria.

b. Findings

No findings of significance were identified.

-12-Containment Integrated Leak Rate Surveillance (70313)

a. Inspection Scope

The inspectors verified through observation, records review, and independentcalculations whether the containment integrated leak rate test was being properly conducted. In addition, the inspectors independently verified the acceptability of the test results through real time observations and analysis and further in-depth independent analysis. The inspectors:

(1) ensured that the alignment and operation of all systems and equipment inside and penetrating containment was appropriate,

(2) conducted inspections of the accessible portions of containment,

(3) verified equipment calibration, and

(4) ensured appropriate success criteria were being followed per the approved procedure.

b. Findings

No findings of significance were identified.

Containment Leak Rate Test Results Evaluation (70323)

a. Inspection Scope

The inspectors verified through direct observation and records review that the licenseehad adequately performed, reviewed, and evaluated the as-found and as-left containment integrated leak rate test.

This review was to ensure that the containment building function was not impacted by the temporary opening which allowed for the replacement of the steam generators and the reactor vessel head.

b. Findings

No findings of significance were identified.

1R20 Refueling and Outage Activities

a. Inspection Scope

The inspectors evaluated licensee's 1RF12 activities to ensure that risk was consideredwhen developing and when deviating from the outage schedule, the plant configurationwas controlled in consideration of facility risk, mitigation strategies were properlyimplemented, and Technical Specification requirements were implemented to maintainthe appropriate defense-in-depth. The inspectors reviewed and/or observed thefollowing items, listed below, as they pertained to the steam generator and reactorvessel head replacement. Coverage of the full scope of Inspection Procedure 71111.20is documented in Inspection Reports 05000445/446-2007002 and05000445/446-2007003.Unit shutdown and cooldownReduced reactor coolant inventory activities

-13-Defense in depth and mitigation strategy implementationContainment closure capabilityRefueling activities that included fuel offloading, fuel transfer, and core reloadingImplementation of procedures for foreign material exclusionElectrical power source arrangementContainment cleanup and inspectionUnit heatup and startupLicensee identification and resolution of problems related to refueling activities

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications (71111.23)

a. Inspection Scope

The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), plantdrawings, procedure requirements, Technical Specification and Technical Requirements Manual to ensure that the below listed temporary modification was properly implemented. The inspectors:

(1) verified that the modification did not have an affect on system operability/availability;

(2) verified that the installation was consistent with the modification documents;

(3) ensured that the post-installation test results were satisfactory and that the impact of the temporary modification on permanently installed SSCs were supported by the test;

(4) verified that the modification was identified on control room drawings and that appropriate identification tags were placed on the affected equipment; and

(5) verified that appropriate safety evaluations were completed.

The inspectors verified that licensee identified and implemented any needed corrective actions associated with temporary modification.Unit 1 Containment Alternate Access, for steam generator and reactor vesselhead replacement, in accordance with Final Design Authorization (FDA)

FDA-2005-000658-01-02, observed, and reviewed February 24, 2007 through April 6, 2007The inspectors completed one sample.

b. Findings

No findings of significance were identified.

-14-2.RADIATION SAFETYCornerstone: Occupational Radiation Safety2OS1Access Control To Radiologically Significant Areas (71121.01)

a. Inspection Scope

This area was inspected to assess the licensee's performance in implementing physicaland administrative controls for airborne radioactivity areas, radiation areas, high radiation areas, and worker adherence to these controls. The inspectors used the requirements in 10 CFR Part 20, the Technical Specifications, and the licensee's procedures required by Technical Specifications as criteria for determining compliance. During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation protection workers.Additionally, using Inspection Procedure 71121.01, "Access Control to RadiologicallySignificant Areas," the inspectors reviewed activities associated with the steam generator and reactor vessel head replacement to fulfill the inspection requirements of Inspection Procedure 50001, "Steam Generator Replacement Inspection," and Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection."

Specifically, the inspectors reviewed the controls in place at the old steam generator and reactor head storage facility. The inspectors inspected the facility, took independent dose rate measurements, and reviewed the licensee's survey plan. See NRC Inspection Report 05000445;446/2007003, Section 2OS1, for additional information.The inspectors reviewed the following items:

Controls (surveys, posting, and barricades) of radiation, high radiation, orairborne radioactivity areasRadiation work permits, procedures, engineering controls, and air samplerlocationsConformity of electronic personal dosimeter alarm set points with surveyindications and plant policy; workers' knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarmsBarrier integrity and performance of engineering controls in airborne radioactivityareasSelf-assessments, audits, licensee event reports, and special reports related tothe access control program since the last inspectionCorrective action documents related to access controls Licensee actions in cases of repetitive deficiencies or significant individualdeficiencies Radiation work permit briefings and worker instructions

-15-Adequacy of radiological controls, such as required surveys, radiation protectionjob coverage, and contamination control during job performance Dosimetry placement in high radiation work areas with significant dose rategradientsChanges in licensee procedural controls of high dose rate - high radiation areasand very high radiation areasControls for special areas that have the potential to become very high radiationareas during certain plant operationsPosting and locking of entrances to all accessible high dose rate - high radiationareas and very high radiation areasRadiation worker and radiation protection technician performance with respect toradiation protection work requirements The samples completed for Inspection Procedure 71121.01 will be tracked inSection 2OS1 of NRC Inspection Report 05000445;446/2007003.

b. Findings

No findings of significance were identified.2OS2ALARA Planning and Controls (71121.02)

a. Inspection Scope

The inspectors assessed licensee performance with respect to maintaining individualand collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors used the requirements in 10 CFR Part 20 and the licensee's procedures required by technical specifications as criteria for determining compliance. During the inspection, the inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers.Additionally, using Inspection Procedure 71121.02, "ALARA Planning and Controls," theinspectors reviewed activities associated with the steam generator and reactor vessel head replacement to fulfill the inspection requirements of Inspection Procedure 50001, "Steam Generator Replacement Inspection," and Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection." Specifically, the inspectors reviewed the controls in place at the old steam generator and reactor head storage facility. The inspectors inspected the facility, took independent dose rate measurements, and reviewed the licensee's survey plan. See NRC Inspection Report 05000445;446/2007003, Section 2OS2, for additional information.The inspectors reviewed the following items:

• Outage (1RF12) work activities and associated work activity exposure estimates,which were likely to result in the highest personnel collective exposures

-16-*Site specific trends in collective exposures, plant historical data, and source-termmeasurements*Site specific ALARA procedures

• ALARA work activity evaluations, exposure estimates, and exposure mitigationrequirements*Intended versus actual work activity doses and the reasons for anyinconsistencies *Integration of ALARA requirements into work procedure and radiation workpermit documents*Shielding requests and dose/benefit analyses

• Post-job work activity reviews

• Assumptions and basis for the current annual collective exposure estimate, themethodology for estimating work activity exposures, the intended dose outcome, and the accuracy of dose rate and man-hour estimates*Method for adjusting exposure estimates, or re-planning work, when unexpectedchanges in scope or emergent work were encountered*Exposure tracking system

• Use of engineering controls to achieve dose reductions and dose reductionbenefits afforded by shielding*Workers use of the low dose waiting areas

• Records detailing the historical trends and current status of tracked plant sourceterms and contingency plans for expected changes in the source term due to changes in plant fuel performance issues or changes in plant primary chemistry *Source-term control strategy or justifications for not pursuing such exposurereduction initiatives*Specific sources identified by the licensee for exposure reduction actions andpriorities established for these actions, and results achieved against since the last refueling cycle*Radiation worker and radiation protection technician performance during workactivities in radiation areas, airborne radioactivity areas, or high radiation areas *Self-assessments, audits, and special reports related to the ALARA programsince October 2006*Resolution through the corrective action process of problems identified throughpost-job reviews and post-outage ALARA report critiques

-17-*Corrective action documents related to the ALARA program and follow-upactivities such as initial problem identification, characterization, and tracking *Effectiveness of self-assessment activities with respect to identifying andaddressing repetitive deficiencies or significant individual deficiencies The samples completed for Inspection Procedure 71121.02 will be tracked inSection 2OS2 of NRC Inspection Report 05000445;446/2007003.

b. Findings

No findings of significance were identified.4.OTHER ACTIVITIES

4OA2 Problem Identification and Resolution

.1Inservice Inspection (71111.08)

a. Inspection Scope

The inspection procedure requires review of a sample of problems associated withinservice inspections documented by the licensee in the corrective action program for appropriateness of the corrective actions.The inspectors reviewed eight corrective action documents (Smart Forms) which dealtwith inservice inspection activities and found that the corrective actions wereappropriate. From this review the inspectors concluded that the licensee had an appropriate threshold for entering issues into the corrective action program and has procedures that direct a root cause evaluation when necessary. The licensee also had an effective program for applying industry operating experience.

b. Findings

No findings of significance were identified..2Steam Generator and Reactor Vessel Head Replacement Inspection (50001, 71007)

a. Inspection Scope

The inspectors reviewed a sample of the problems identified and documented in thelicensee's corrective action program for appropriateness of the corrective actions. The inspector reviewed over eighty corrective action documents which were related to the steam generator and reactor vessel head replacement project and found that the corrective actions were appropriate. The review concluded that the licensee had an appropriate threshold for entering issues into the corrective action program and has procedures that deal with resolution of the issues, even directing a root cause evaluation if necessary. The inspectors also attended numerous contractor overview meetings,

-18-that discussed issues identified in the contractor's corrective action program, to ensurethat items were entered into the licensee's corrective action program as necessary. The inspectors also determined that the licensee effectively sought and implemented industry operating experience.

b. Findings

No findings of significance were identified.4OA5Other Activities.1Steam Generator Replacement Inspection (50001)Design and Planning Inspections (Section 02.02)

a. Inspection Scope

The inspectors used the guidance in Inspection Procedure 50001, "Steam GeneratorReplacement Inspection," Section 02.02, and inspection procedures referenced therein, to perform the steam generator removal and replacement activities listed below.Engineering and Technical SupportThe inspection activities specified by Section 02.02.a, "Steam Generator Removal andReplacement Inspections," were accomplished in accordance with Inspection Procedures 71111.02, "Evaluation of Changes, Tests, or Experiments", and 71111.17, "Permanent Plant Modifications." These inspections are documented in Sections 1R02 and 1R17 of this inspection report.Lifting and RiggingIn accordance with Section 02.02.b, the inspectors reviewed the applicable engineeringdesign, modification, and analysis associated with steam generator lifting and rigging including:

(1) crane and rigging equipment,

(2) steam generator component drop analysis,

(3) safe load paths, and

(4) load lay-down areas. The inspection focused on the impact of load handling activities on reactor core or spent fuel and its cooling, and plant support systems for Unit 1 and common systems for the operation of Unit 2. Radiation ProtectionIn accordance with Section 02.02.c, the inspectors reviewed radiation protectionprogram controls, planning, and preparation in:

(1) as low as reasonably achievable planning,

(2) dose estimates and tracking,

(3) exposure and contamination controls,

(4) radioactive material management,

(5) radiological work plans and controls,

(6) emergency contingencies, and

(7) project staffing and training plans. The results are documented in Sections 2OS1 and 2OS2 above, as well as in NRC Inspection Report 05000445;446/2007003, Sections 2OS1 and 2OS2.

-19-Security Considerations and Adverse Impact to the Other UnitIn accordance with Section 02.02.d, the inspectors interviewed security specialists andofficers specifically assigned to the steam generator and reactor vessel head replacement project. The inspectors also made frequent observations of security practices during all stages of the project to verify vital and protected barriers were not affected or compromised. The inspectors also reviewed impacts to Unit 2 (operating unit) stemming from the replacement project as activities and schedules changed.

b. Findings

No findings of significance were identified.

Steam Generator Removal and Replacement Inspections (Section 02.03)

a. Inspection Scope

The inspectors used the guidance in Inspection Procedure 50001, Section 02.03, andinspection procedures referenced therein, to perform the steam generator removal and replacement activities listed below.Welding and Nondestructive Examination ActivitiesIn accordance with Section 02.03.a, inspections were conducted to review welding andNDE activities including:

(1) special procedures,

(2) training and qualifications,

(3) radiography results and work packages,

(4) completion of preservice NDE requirements for welds, and

(5) completion of baseline eddy current examination of new steam generator tubes. This inspection was performed as part of Inspection Procedure 71111.08, in Section 1R08 of this report.Lifting and Rigging ActivitiesIn accordance with Section 02.03.b, and Inspection Procedure 71111.23, "TemporaryPlant Modifications," the inspectors observed and reviewed several activities associated with lifting and rigging. The inspectors observed and reviewed preparations, crane and rigging inspections, testing, and equipment lay-down areas associated with the following activities:*Construction of the outside lift system*Inspection and testing of the outside lift system

• Temporary lift device (inside containment) construction and removal

• Reactor cavity and containment decking (for storage) construction and removal

• Old steam generator removal

• New steam generator installationMajor Structural Modifications and Containment Access and IntegrityIn accordance with Section 02.03.c and .d and Inspection Procedures 71111.17 and71111.23, the inspectors observed the implementation, restoration, where applicable, and removal of the installation of the following structural modifications to support the two steam generator replacement activities listed below.

-20-*Complete removal of the upper steam generator snubber supports*Temporary alternate containment accessThe testing activities for the repair and recovery of the containment building can befound in Sections 1R19, while more information concerning the temporary modification can be found in Section 1R23 of this report.Unit 1 Outage Operating ConditionsThe inspectors used Section 02.03.e and Inspection Procedure 71111.20, "Refuelingand Outage Activities," to complete this inspection. Section 1R20 contains a more detailed explanation of what was observed and reviewed.Radiation Protection ControlsThis inspection was performed during the outage by regional inspectors and the resultsare documented in Sections 2OS1 and 2OS2 of this report.Foreign Material ControlsThe inspectors followed the guidance contained in Section 02.03.e. The inspectorsreviewed and observed procedural controls, field observations, and the licensee's Plant Event Review Committee (PERC) meetings and various other meetings discussing the foreign material control issues. The inspectors paid particular attention to the reactor coolant and secondary side openings.Temporary ServicesIn accordance with Sections 02.03.e, the inspectors reviewed work orders, proceduresand observed activities, and performed walkdowns of temporary systems in the containment building. The inspectors also reviewed the fire protection, and industrial safety aspects for alternate construction power, and welding activities.Radiological Safety Plans for the Old Steam Generator and Reactor Vessel HeadStorage FacilityIn accordance with Section 02.03.f, the inspectors reviewed the licensee's radiologicalsafety plans for the storage facility. The inspectors also performed a complete walkdown of the storage facility. This inspection area was also reviewed by regional health physicists inspectors and is documented in Sections 2OS1 and 2OS2 of this report.

b. Findings

No findings of significance were identified.

-21-Post-installation Verification and Testing Inspection (Section 02.04)

a. Scope

The inspectors used the guidance in Inspection Procedure 50001, Section 02.04, andinspection procedures referenced therein, to perform the steam generator removal and replacement activities. Selective inspections were performed in the following areas:

(1) containment testing,

(2) post-installation inspections and verifications program and its implementation,

(3) conduct or RCS leakage testing,

(4) conduct of the SG secondary side leakage testing,

(5) calibration and testing of instrumentation affected by the SG replacement,

(6) procedures required to confirm design and to establish baseline measurements and conduct of testing, and

(7) pre-service inspection of new welds.

Specific items reviewed are documented in Section 1R19.

b. Findings

No findings of significance were identified..2Reactor Vessel Head Replacement Inspection (71007)Design and Planning Inspections (Section 02.02)

a. Inspection Scope

The inspectors used the guidance in Inspection Procedure 71007, "Reactor VesselHead Replacement Inspection," Section 02.02, and inspection procedures referenced therein, to perform the reactor vessel head removal and replacement activities listed below.Engineering and Technical SupportThe inspection activities specified by Section 02.02.a of Inspection Procedure 71007,were accomplished in accordance with Inspection Procedures 71111.02, "Evaluation of Changes, Tests, or Experiments", and 71111.17, "Permanent Plant Modifications."

These inspections are documented in Sections 1R02 and 1R17 of this report.Lifting and RiggingIn accordance with Section 02.02.b of Inspection Procedure 71007, the inspectorsreviewed the applicable engineering design, modification, and analysis associated with Reactor Vessel Head lifting and rigging including:

(1) crane and rigging equipment,

(2) Steam Generator component drop analysis,

(3) safe load paths, and

(4) load lay-down areas. The inspection focused on the impact of load handling activities on reactor core or spent fuel and its cooling, and plant support systems for the reactor unit and common systems for the other operation unit at the site. Radiation ProtectionThe review of radiation protection program controls, planning, and preparation in:

(1) ALARA planning,

(2) dose estimates and tracking,

(3) exposure and contamination controls,

(4) radioactive material management,

(5) radiological work plans and controls,

-22-(6) emergency contingencies, and

(7) project staffing and training plans are documentedin Section 2OS1 and Section 2OS2 above, as well as in NRC Inspection Report 05000445;446/2007003, Section 2OS1 and Section 2OS2.Security Considerations and Adverse Impact to the Other UnitIn accordance with Section 02.02.d, the inspectors interviewed security specialists andofficers specifically assigned to the steam generator and reactor vessel head replacement project. The inspectors also made frequent observations of security practices during all stages of the project to verify vital and protected barriers were not affected or compromised. The inspectors also reviewed impacts to Unit 2 (operating unit) stemming from the replacement project as activities and schedules changed.

b. Findings

No findings of significance were identified.

Reactor Vessel Head Fabrication Inspections at Licensee Facility (Section 02.03)

a. Inspection Scope

The inspectors used the guidance in Inspection Procedure 71007, Section 02.03, andinspection procedures referenced therein, to perform the following reactor vessel head fabrication inspection activities.Heat TreatmentThe inspectors verified that the material heat treatment used to enhance the mechanicalproperties of the Reactor Vessel head material carbon, low alloy, and high alloy chromium steels was conducted per the ASME Code,Section III requirements. Also,the inspections were performed to verify that adequate heat treatment procedures were available to assure that the following requirements were met:

(1) furnace atmosphere,

(2) furnace temperature distribution and calibration of measuring and recording devices,

(3) thermocouple installation,

(4) heating and cooling rates,

(5) quenching methods, and

(6) record and documentation requirements.Nondestructive ExaminationInspections were conducted to ensure the manufacturing control plan includedprovisions for monitoring NDE, and to ascertain that the NDE was performed in accordance with applicable code, material specification, and contract requirements.WeldingThe inspectors reviewed the documentation for the weld overlay welding operations thatestablished a layer of stainless steel cladding on the inside of the reactor vessel head to determine if it was accomplished per design. The inspectors also selected a sample of dome-to-flange and control rod drive mechanism (CRDM) flange-to-nozzle welds and reviewed the following items:

(1) certified mill test reports of the dome, flange, weld material rods, and CRDM nozzles;

(2) certified mill test reports for the welding material for the reactor vessel head cladding;

(3) cladding weld records, weld rod material control

-23-requisitions, traceability of weld material rods, weld procedure qualification, welderqualifications, and nonconformance reports;

(4) CRDM nozzle cladding welding inspection records, weld rod material control requisitions, traceability of weld material rods, weld procedure qualification, welder qualifications, and nonconformance reports;

(5) CRDM to nozzle welding and welds inspection records, weld rod material control requisitions, traceability of weld material rods, weld procedure qualification, welder qualifications, and nonconformance reports; and

(6) NDE procedures, NDE records of the welds, NDE personnel qualifications, and certification of NDE solvents.ProceduresInspections were completed to ensure that repair procedures had been established andthat these procedures were consistent with applicable ASME Code, material specification, and contract requirements by verifying:

(1) repair welding was conducted in accordance with procedures qualified to Section IX of the ASME Code,

(2) all welders had been qualified in accordance with Section IX of the ASME Code,

(3) records of the repair were maintained, and

(4) that requirements had been established for the preparation of certified material test reports and that the records of all required examinations and tests were traceable to the procedures to which they were performed.Code ReconciliationThe inspectors reviewed the required documentation, supplemental examinations,analysis, and ASME Code documentation reconciliation to ensure that the original ASME Code N-Stamp remains valid, and that the replacement head complies with appropriate NRC rules and industry requirements. The inspectors also ensured that the design specification was reconciled and a design report was prepared for the reconciliation of the replacement head, verifying that they were certified by professional engineers competent in ASME Code requirements.Quality Assurance ProgramInspections were conducted to ensure that machining was carried out under a controlledsystem of operation, a drawing/document control system was in use in the manufacturing process, and that part identification and traceability was maintained throughout processing and was consistent with the manufacturer's Quality Assurance program. In addition, the inspectors ensured that only the specified drawing and document revisions were available on the shop floor and were being used for fabrication, machining, and inspection through review of applicable procedures.Compliance InspectionThe inspectors verified that the original ASME Code,Section III, data packages for thereplacement Reactor Vessel head were supplemented by documents included in the ASME Code Section XI, (preservice inspection) data packages; examined selected manufacturing and inspection records of the finished machined Reactor Vessel head; and verified compliance with applicable documentation requirements.

b. Findings

No findings of significance were identified

.

-24-Reactor Vessel Head Removal and Replacement (Section 02.04)Lifting and Rigging ActivitiesIn accordance with Section 02.04.a, and Inspection Procedure 71111.23, "TemporaryPlant Modifications," the inspectors observed and reviewed several activities associated with lifting and rigging. The inspectors observed and reviewed preparations, crane and rigging inspections, testing, and equipment lay-down areas associated with following activities:*Construction of the outside lift system*Inspection and testing of the outside lift system

• Temporary palfinger crane for servicing and preparing the reactor vessel head

• Reactor cavity and containment decking (for storage) construction and removal

• Old reactor vessel head removal

• New reactor vessel head installation, and vessel setMajor Structural ModificationsThis inspection was not applicable due to the lack of major structural modifications. Theonly modification was the installation of the new control rod drive mechanism vent fan modification. The modification was reviewed as part of the 71111.02, "Evaluations of Changes, Tests, or Experiments," and 71111.17, "Permanent Plant Modification,"

inspection. This modification item is documented in Section 1R02 and 1R17.Containment Access and IntegrityThe inspection is documented in Sections 1R19, 1R23, and Section 4OA5.1.

Unit 1 Outage Operating ConditionsThe inspectors used Section 02.04.d and Inspection Procedure 71111.20, "Refuelingand Outage Activities," to complete this inspection. Section 1R20 contains a more detailed explanation of what was observed and reviewed.Radiation Protection ControlsThis inspection was performed during the outage by regional inspectors and the resultsare documented in Sections 2OS1 and 2OS2 of this report.Foreign Material ControlsThe inspectors followed the guidance contained in Section 02.04.d. The inspectorsreviewed and observed procedural controls, field observations, and the licensee's plant event review committee (PERC) meetings and various other meetings discussing the foreign material control issues. The inspectors paid particular attention to the reactor coolant and secondary side openings.

-25-Temporary ServicesIn accordance with Sections 02.04.d, the inspectors reviewed work orders, proceduresand observed activities, and performed walkdowns of temporary systems in the containment building. The inspectors also reviewed the fire protection, and industrial safety aspects for alternate construction power, and welding activities.Radiological Safety Plans for the Old Steam Generator and Reactor Vessel HeadStorage FacilityIn accordance with Section 02.04.e, the inspectors reviewed the licensee's radiologicalsafety plans for the storage facility. The inspectors also performed a complete walkdown of the storage facility.

b. Findings

No findings of significance were identified.

Post-installation Verification and Testing Inspection (Section 02.05)

a. Scope

The inspectors used the guidance in Inspection Procedure 71007, Section 02.05, andinspection procedures referenced therein, to perform the steam generator removal and replacement activities. Selective inspections were performed in the following areas:

(1) containment testing,

(2) post-installation inspections and verifications program and its implementation,

(3) conduct or RCS leakage testing,

(4) conduct of the SG secondary side leakage testing,

(5) calibration and testing of instrumentation affected by the SG replacement,

(6) procedures required to confirm design and to establish baseline measurements and conduct of testing, and

(7) pre-service inspection of new welds.

Specific items reviewed are documented in Section 1R19.

b. Findings

No findings of significance were identified.4OA6Meetings, Including ExitExit Meeting SummaryOn February 9, 2007, the inspectors presented the safety evaluation and permanentplant modifications inspection results to Mr. Steve L. Smith, Site Engineering Director, and other members of the staff who acknowledged those results. No proprietary information was included in this report.On March 29, 2007, the inspectors presented the In-Service Inspection, SteamGenerator and Reactor Vessel Closure Head Replacement Activities inspection results to Mr. Steve L. Smith, Site Engineering Director, and other members of the staff who acknowledged those results. No proprietary information was included in this report.

-26-On May 21, 2007, the inspectors presented the resident inspection results toMr. R. Flores, Vice President Nuclear Operation, and other members of licensee management. No proprietary information was included in this report.ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Abbott, Engineer, Westinghouse

W. Bamford, Engineer, Westinghouse

D. Bersi, Steam Generator Replacement Project, Component Design/Fabrication Lead

M. Blevins, Senior Vice President and Chief Nuclear Officer

J. Brabec, Steam Generator Replacement Project, Installation Manager/Asst. Project Manager

S. Bradley, Supervisor, Health Physics, Radiation Protection & Safety Services

A. Caves, ALARA Coordinator

T. Clouser, Manager, Shift Operations

W. Crosby, Bechtel, NDE Level III

J. Curtis, Radiation Protection Manager, Radiation and Industrial Safety

T. Dorris, Bechtel, Lead Weld Engineer

B. Emanuel, Radiation Protection ALARA

J. Finneran, Steam Generator Replacment Project, Project Engineering Manager

R. Flores, Vice President, Nuclear Operations

J. Gallman, Senior Nuclear Analyst (Work Week Coordinator)

R. Garcia, Supervisor, Radioactive Material Control

D. Haggerty, Project Engineer, Bechtel

N. Harris, Consulting Licensing Analyst

B. Henley, Engineering Consultant (Seismic Analysis)

G. Hietpas, AREVA, Site Director

D. Holland, Senior Nuclear Analyst (Work Week Coordinator)

N. Hood, Project Engineering Manager

T. Hope, Regulatory Performance Manager

M. Kanavos, Plant Manager

S. Karpyak, Risk & Reliability Engineering Supervisor

R. Kidwell, Sr. Nuclear Technologist, Regulatory Affairs

M. Killgore, Engineering Support Director

D. Kissinger, Design Engineering Analysis Engineer

G. Krishnan, Procurement Engineering & Program Manager, SHAW

D. Kross, Director, Maintenance

J. Lamarca, Engineering Smart Team Manager

S. Lantis, Lead Weld Superintendent/Dayshift

B. Lichtenstein, Engineer, Risk and Reliability, Westinghouse

F. Madden, Director, Regulatory Affairs

F. Maddy, JET Engineer

S. Maier, Design Engineering Analysis Manager, Technical Support

B. Mays, Steam Generator Project Manager

E. Meaders, Outage Manager

J. Mercer, Maintenance Rule Coordinator

G. Merka, Regulatory Affairs

J. Meyer, Technical Support Manager

S. Miller, Senior Engineering Analyst, Results Engineering

G. Morini, Westdyne, Project Manager

W. Morrison, Maintenance Smart Team Manager

AttachmentA-2

D. O'Connor, Supervisor, Radiation Protection, Radiation Protection & Safety Services

W. Olsen, Bechtel, Lead Mechanical Weld QC

P. Passalugo, SHAW, ISI Program Lead

J. Patton, Supervisor, Quality Assurance

C. Peters, Bechtel, CAD Weld QC

K. Pitilli, Design Engineering Analysis Engineer

L. Pope, System Engineer

H. Quach, AREVA, Principal Engineer

W. Reppa, JET Manager

J. Rincon, Radiation Protection ALARA

J. Seawright, Consulting Engineer, Regulatory Affairs

R. Segura, Nuclear Analyst Consultant (Electrical Systems)

J. Simmons, Manager, Radiation Protection, Steam Generator Replacement Project

R. Smith, Director, Operations

S. Smith, Site Engineering Director

D. Snow, Regulatory Affairs

D. Sparks, Senior Nuclear Analyst (Work Week Coordinator)

J. Stansbury, Radiation Protection, Sr. Technician

J. Taylor, Engineering Smart Team Manager

D. Tirsun, Engineer, Risk and Reliability, Westinghouse

C. Tran, Engineering Programs Manager

I. Whitt, Engineer, Boric Acid Corrosion Detection Program

D. Wilder, Radiation and Industrial Safety Manager

H. Winn, System Engineer

T. Wright, Bechtel

G. Yezefski, System Engineer

NRC

D. Allen, Senior Resident Inspector

A. Sanchez, Resident Inspector

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened NoneOpened and

Closed

None

Closed

None

Discussed

None

AttachmentA-3

Section 1R02:

Evaluations of Changes, Tests, or Experiments (71111.02)EvaluationsDocument NumberTitle/DescriptionRevision59EV-2005-000224-01-00Main Feedwater Modification Due to ReplacementSteam Generators

010

CFR 50.59 ScreeningsDocument NumberTitle/DescriptionRevision59SC-2005-000658-02-01Rigging and Transport of OSG and RSG1

59SC-2005-000224-01-00Main Feedwater Piping Modifications Due toReplacement Steam Generators

0

Section 1R08:

Inservice Inspection ActivitiesReports:Report No.TitleDate12UT-11Calibration Data Sheet3/2007

2UT-14Calibration Data Sheet3/2007

2UT-17Calibration Data Sheet3/2007

2UT-20Calibration Data Sheet3/2007Procedures:Procedure No.TitleRevision5069-000-4MP-T040-W0109Bechtel Welding Procedure Specification P1-T225069-000-4MP-T040-W0458Bechtel Welding Procedure Specification P8-T(RA)125069-000-4MP-T040-W0576Bechtel Welding ProcedureSpecification P1-AT-Lh(CVH+30

o F)1

Procedures:AttachmentA-4Procedure No.TitleRevision25069-000-4MP-T040-W0579Bechtel Welding ProcedureSpecification P1-T-0(CVN+30)

0ENSA-GP-7.1Distribution and Control of Documents33EPG-9.02CPSES Alloy 600 Management Program0

EPG-703Inservice Inspection Program1

EPG-731ASME Section XI Repair/Replacement Activities1

GWS-1General Welding Standard2

NDE 7.10Steam Generator Tube Selection and Examination11

NDE 3.02 ASME Section XI Magnetic Particle Examination3

PQR 1041Bechtel Welding Procedure Qualification Record10

RT-1NDE Procedure Radiographic Examination10

STA-703Inservice Inspection Program13

STA-731ASME Section XI Repair & Replacement Activities6

STA-733Steam Generator Reliability Program10

STA-760RCS Materials Management Program1

TX-ISI-08VT-1 and

VT-3 Examination Procedure for CPSES6

TX-ISI-11Liquid Penetrant Examination for Comanche PeakSteam Electric Station

11TX-ISI-302Ultrasonic Examination of Austenitic Piping Welds2VL-04-002930ENSA Welding Procedure Specification0

VL-04-002931ENSA Welding Procedure Specification0

VL-04-002933ENSA Welding Procedure Specification0

Procedures:AttachmentA-5Procedure No.TitleRevisionVL-05-000111Acceptance Testing for Weld Overlay Cladding(Welding Strip

ER 309L + Flux)

0VL-05-000120Quality Plan for Closure Head Forging2VL-05-000426Visual Examinations2

VL-05-000679ENSA Welding Procedure Specification2

VL-05-000996ENSA Welding Procedure Specification0

VL-05-001245ENSA Welding Procedure Specification0

VL-05-001329Measurement of Cladding Thickness by Ultrasonics1

VL-05-001484ENSA Welding Procedure Specification1

VL-05-001564Acceptance Testing for Weld Overlay Cladding(Welding Strip

ER 308L + Flux)

2VL-05-002108ENSA Welding Procedure Specification0VL-05-002708Preheating and Hydrogen Bake Requirements0

VL-05-002709Magnetic Particles Examination1

VL-05-002245Post Weld Stress Relief Heat Treatment1

VL-05-003073ENSA Welding Procedure Specification3

VL-05-003074ENSA Welding Procedure Specification3

VL-05-003331Ultrasonic Examination of the CRDMH/CETNA/RVLMSFull Penetration Welds

2VL-06-000078ENSA Welding Procedure Specification1VL-06-000172Radiographic Examinations4

VL-06-000410Liquid Penetrant Examinations2

VL-06-000413ENSA Welding Procedure Specification4

Procedures:AttachmentA-6Procedure No.TitleRevisionVL-06-000599ENSA Welding Procedure Specification1VL-06-000741Preservice Examinations-Manual Ultrasonic Inspectionof the CRD Full Penetration Welds of the Replacement

RPV Head of Comanche Peak Unit 1

1VL-06-000870Non Destructive Examinations after the HydrostaticPressure Testing

2VL-06-000918ENSA Welding Procedure Specification0VL-06-001356Procedure for Ultrasonic Examination of the ReactorVessel Closure Head Penetrations During the Comanche Peak 1 RRVCH Pre-service Inspection

1VL-06-001357Procedure for the Remote Visual Examination of theReactor Pressure Vessel Head During the Comanche Peak 1 RRVCH Pre-service Inspection

1VL-06-001510Preservice Examinations-Examination using DyePenetrants, not Soluble in Water, and Directly Visible by Color Contrast on Replacement RPV Head of Comanche Peak Unit 1

1VL-06-001514Procedure for the Eddy Current Pre-service Inspectionof the Outer Surface of Penetration Nozzle and the J-

Groove Weld (CRD Area) of Comanche Peak 1

RRVCH 1VL-06-001515Procedure for the Eddy Current Pre-service Inspectionof the J-Groove Weld (Cladding Area) of Comanche Peak 1 RRVCH

1VL-06-001516Procedure for the Eddy Current Pre-service Inspectionof the Vent Pipe (J-Groove Weld and Inner Surface) of Comanche Peak 1 RRVCH

1VL-06-001517Procedure for the Eddy Current Pre-service Inspectionof Open Penetration Nozzles of Comanche Peak 1

RRVCH 1

Procedures:AttachmentA-7Procedure No.TitleRevisionVL-06-001804Guidelines for Analyzing Data from PWR ReactorVessel Head Penetrations Using MASERA and

MASERA-TOFD During the Comanche Peak 1

RRVCH Pre-Service Inspection

2VL-06-002182Project M505 - RRVCH Arc Strike Repair Procedure(MRR No. 1571X)

0VL-05-002245Post Weld Stress Relief Heat Treatment0VL-06-003406Comanche Peak Unit 1 Replacement RV ClosureHead - ASME Design Summary

3WD-1Bechtel Welding Standard Documentation of Welds3WLD-103Welder Performance Qualifications6

WCI-606Work Control Process9Design Documents:Document No.TitleRevisionDBD-CS-018Design Criteria for Pipe Stress and Pipe Supports7

2EP-5.13Guidelines for Wall Thinning Evaluation for ASMECode Class 2, 3, and ANSI B31.1 Piping

0900580-07Comanche Peak Unit 1: Operational Qualification forDimetrics Gold Track II Welding System

0Calculations:Calculation No.TitleRevisionCT-2-030Pipe Stress Calculation for Containment Spray PipingStress Problem CT-2-030

2CT-2-031Pipe Stress Calculation for Containment Spray PipingStress Problem CT-2-031

AttachmentA-8Miscellaneous Documents:Document No.TitleRevisionLetterFort Calhoun Station, Unit No. 1 - Relief Request forthe Use of Radiography using Phosphor Imaging Plate

(TAC No. MC8843)5/2006LetterComanche Peak Steam Electric Station (CPSES) Unit1 - Summary of Conference Calls with TXU Energy to Discuss the 2004 Steam Generator Tube Inspections

(TAC No. MC2564)7/2004LetterComanche Peak Steam Electric Station Unit 1 -Summary of the Tenth Refueling Outage (1RF10)

Steam Generator Tube Inservice Inspection (TAC No.

MC4458)7/2005EVAL-2006-000751-01-00Relief Request for use of Phosphor Imaging PlatesEVAL-2003-002426-25-00Evaluation to Allow the use of Digital RadiographicExamination

1/2007TXX-04141Comanche Peak Steam Electric Station (CPSES)Unit 1, Docket No. 50-445 Submittal of Unit 1 Tenth Refueling Outage (1RF10)

GL 95-05 Report7/2004TXX-04157Comanche Peak Steam Electric Station (CPSES)Unit 1 Tenth Refueling Outage (1RF10) GL 95Steam Generator Twelve Month Report8/2004TXX-04172Comanche Peak Steam Electric Station (CPSES)Unit 1, Docket No. 50-445 Submittal of Corrected Unit Tenth Refueling Outage (1RF10)

GL 95-05 Report9/2004TXX-05059Comanche Peak Steam Electric Station (CPSES)Unit 1, Docket No. 50-445 CPSES Response to Request for Additional Information Concerning the Spring 2004 (1RF10) Steam Generator Inservice Inspection Reports3/2005TXX-07013Comanche Peak Steam Electric Station (CPSES)Docket Nos. 50-445 and 50-446 Inspection and Mitigation of Alloy 82/182 Pressurizer Butt Welds

1/2007Smartforms:SMF-2006-002066-00

AttachmentA-9 Codes:ASME Code Secti on III, 1989 EditionASME Code Section IX, 1998 Edition

ASME Code Section XI, 1998 EditionSpecifications:CPSES-P-1079, Rev. 6

Section 1R13: Maintenance Risk Assessments and Emergent Work Evaluation(71111.13)CPSES Containment Crane PlanDesign Basis Document

DBD-ME-006NSSS Upgrade Project Containment Crane Plan

Procedure

MDA-304, Control of Heavy Loads and Critical Lifts Engineering Report: PRA considerations Related to Multiple Crane Operations InsideContainment During 1RF12CPSES 1RF12 Outage Scope Presentation

Section 1R17: Permanent Plant Modifications (71111.17A)Final Design AuthorizationNumberTitleRevision/DateFDA-2005-000224-07-02Instrumentation and Control Change to the

SGLevel Instrumentation System for the SGRP

2FDA-2003-002426-01-00FSAR update for Steam Generator Replacement0FDA-2005-000224-04-02Modify the Main Steam Piping System toSupport Replacement of the Unit1 Steam Generators in 1RF12

2FDA-2005-000658-03-01Design and Construct the Systems andStructures Needed to Move the Old and New Steam Generators and Reactor Vessel Heads

1FDA-2005-000658-02-01Rigging and Transporting of Steam Generatorsand Reactor Vessel Head

1FDA-2005-000658-03-00Roads and Haul Route Engineering Basis0

Final Design AuthorizationNumberTitleRevision/DateAttachmentA-10FDA-2005-000658-01-01Design and construct the systems and structuresrequired to create and restore the Unit-1 Steam Generators and Reactor Vessel Head Replacement Project Containment Alternate

Access.1FDA-2004-002711-01-00Develop design modification for Replacement ofthe Unit-1 Reactor Vessel Closure Head and Control Rod Drive Mechanisms

0FDA-2005-000224-02-01Modify the Auxiliary Feedwater (AFW) PipingSystem to Support Replacement of the Unit 1

Steam Generator in 1RF12

1FDA-2005-000224-01-00Main Feedwater Piping Modifications Due toReplacement Steam Generators

0FDA-2005-000658-02-01Rigging and Transport of OSG and RSG1CalculationsNumberTittleRevision/DateME-CA-0000-5208Main Steam and Feedwater Penetration AreaEnvironmental Analysis

3NUB-099Subcompartment Analysis for Main SteamLine Penetration Area

1NUB-168Steam Generator Main Steamline Break withGradual Pipe Separation at Break

25069-100-COC-1000-0001Evaluation of Buried Utilities and At-GradeStructures Along OSG and RSG Route

1DrawingsNumberTitleSheet No.SK-F16-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 14SK-F18-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 15SK-F01-05-000658-03-01Heavy Haul Route Location PlanSheet 1

DrawingsNumberTitleSheet No.AttachmentA-11SK-F02-05-000658-03-01Replacement Steam Generators andReplacement Reactor Vessel Head Offload AreaSheet 1SK-F03-05-000658-03-01Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 1SK-F04-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 2SK-F05-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 3SK-F06-05-000658-03-01Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 4SK-F07-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 5SK-F08-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 6SK-F09-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 7SK-F10-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 8SK-F11-05-000658-03-01Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 9SK-F12-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 10SK-F13-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 11SK-F14-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 12SK-F15-05-000658-03-00Heavy Haul Route Modifications and ExistingCommodities ProtectionSheet 13SK-F17-05-000658-03-01Replacement Steam Generator Storage Facilityand Replacement Reactor Vessel Head Storage Facility Enclosure Plan and SectionsSheet 1SK-F19-05-000658-03-01Heavy Haul Route and Protection DetailsSheet 125069-100-V14-UA30-00512-001"Rigging International Drawing of RunwaySystem Decking and Handrail,"

DrawingsNumberTitleSheet No.AttachmentA-1225069-100-V14-UA30-00190-001"Rigging International Drawing - Outside LiftSystem (OLS) Load Test

25069-100-V14-UA30-00189-001"Rigging International Drawing - Outside LiftSystem (OLS) Load Test,"

25069-100-V14-UA30-00188-001"Rigging International Drawing - Outside LiftSystem (OLS) Load Test,"

25069-100-V14-UA30-00264-002"Rigging International Drawing - Handling SG'sInside Containment - General Arrangement,"

25069-100-V14-UA30-00263-002"Rigging International Drawing - Handling SG'sInside Containment - General Arrangement,"

25069-100-V14-UA30-00262-002"Rigging International Drawing - Handling SG'sInside Containment - General Arrangement,"

25069-100-V14-UA30-00261-001"Rigging International Drawing - Handling SG'sInside Containment - General Arrangement,"

25069-100-V14-UA30-00513-001"Rigging International Drawing of RunwaySystem Decking and Handrail,"

0ProceduresNumberTitleRevision/DateECE-5.01-03Design Change Notices and Related ProcessDocumentation

10ECE-5.01-04Technical Evaluation of Replacement Items 3ECE-5.01-08Electronic Design Change Process10STA-716Modification Process16

STA-70710CFR50.59 Reviews16

P-2786-31:Rigging International Procedure.31

AttachmentA-13Miscellaneous DocumentsNumberTitleRevision/DateWCAP-16469PComanche Peak Unit 1 Replacement Steam GeneratorProgram NSSS Engineering Report

1DBD-ME-206Auxiliary Feedwater System19

Section 1R19: Postmaintenance Testing (71111.19)Engineering Position Paper -1RF12

IPO-011A Start-up Testing ReviewSmart Forms2007-14192007-1413

2007-1303

2007-0434

Section 1R23: Temporary Plant Modifications (71111.23)Final Design Authorizations (FDA)2005-3364-02-042005-000658-01-02CPSES Design Basis Document

DBD-CS-073, "Concrete Containment Structure," Revision 7

CPSES Design Basis Document

DBD-CS-074, "Containment Liner and Penetrations,"Revision 7CPSES Design Basis Document

DBD-CS-083, "Containment Concrete Internals," Revision 5

CPSES Design Basis Document

DBD-ME-029, "Seismic Qualification of Equipment,"Revision 10Magnetic Particle Nondestructive Examination Report #

MT-095, MT-096

Leak Testing - Vacuum Box Bubble Test Nondestructive Examination Report # VB-002

Bechtel Specification 25069-100-3PS-DG00-Q0002, "Technical Specification for Installation ofQ (Safety Related) Cadweld Splices," Revision 2Bechtel General Construction Procedure 25069-200-GPP-GCPC-00002, "Cadweld RebarSplices/Testing of Cadweld Rebar Splices," Revision 0

AttachmentA-14

Section 2OS1: Access Controls to Radiologically Significant Areas (71121.01)

Audits and Self-AssessmentsSA-2006-042, Steam Generator Replacement Radiation Protection PreparednessProcedures

STA-650General Health Physics Plan, Revision 5STA-653Contamination Control Program, Revision 10

STA-656Radiation Work Control, Revision 12

STA-660Control of High Radiation Areas, Revision 10

RPI-602Radiological Surveillance and Posting, Revision 29

RPI-610Radiography Controls, Revision 6

Section 2OS2: ALARA Planning and Controls (71121.02)Audits and Self-AssessmentsSelf-Assessment Report

SA-2006-042, Steam Generator Replacement Radiation ProtectionPreparednessSelf-Assessment Report

SA-2006-048, Review of CPSES ALARA Program Shielding Requests2007-132007-15

2007-16Radiation Work Permits 2007-13022007-1305

2006-1306ProceduresRPI-606Radiation Work and General Access Permits, Revision 15STA-651ALARA Program, Revision 9

STA-657ALARA Job Planning/Debriefing, Revision 11OtherCPSES ALARA Committee Meeting Minutes- 11/30/06, 12/14/06, 1/11/07, 1/25/07, 2/1/07, 2/8/071RF12 Comanche Peak NSSS Upgrade Project Manual, Radiation Protection Activity Plans AttachmentA-15

Section 4OA2:

Problem Identification and Resolution (71152)Smart Forms

200717861324122709480780066206111671131211550916076006570609

1516130311530864075906490436

1451130111520864075306470434

1438128511500862074506460356

1406124711460850074306400256

1396123011190843073806390098

139312211058083607310635

136412200981081807260627

133012280950079607100627

2006 2492 2291

22

0751Comanche Peak Steam Electric Station "Quality Assurance Oversight Plan fo NSSS UpgradeProject," Revision 1Weekly Quality Performance Meetings AttendedFebruary 28, 2007March 7, 2007

March 14, 2007

March 21, 2007

March 28, 2007

April 4, 2007

April 11, 2007

April 18, 2007

AttachmentA-16

Section 4OA5: Other Activities (50001, 71007)Procedures:PPT-P1-7001, "ILRT Alignment and Equipment Protection," Revision 0PPT-P1-7002, "ILRT Instrumentation System," Revision 0

PPT-S1-7014, "Containment Integrated Leakage Rate Test," Revision 1

LIST OF ACRONYMS

1RF12unit 1, twelfth refueling outageALARAas low as reasonably achievable

ASMEAmerican Society of Mechanical Engineers

CFRCode of Federal RegulationsCPSESComanche Peak Steam Electric Station

CRDMcontrol rod drive mechanismFDAfinal design authorizationIPOintegrated plant operating proceduresNCVnoncited violation

NDEnondestructive examination

NRCNuclear Regulatory Commission

OPToperations testing manual

PERCplant event review committee

SMFsmart form

SOPsystem operating procedure

SSCstructures, systems, or components

STAstation administrative manual

UFSAR updated final safety analysis report