IR 05000445/2007006
| ML071880010 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 07/06/2007 |
| From: | Clay Johnson NRC/RGN-IV/DRP/RPB-A |
| To: | Blevins M TXU Power |
| References | |
| IR-07-006 | |
| Download: ML071880010 (43) | |
Text
July 6, 2007Mike 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 76043SUBJECT: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 Information
REGION 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 ProjectsAttachment: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,
- (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:
- (2) certified mill test reports for the welding material for the reactor vessel head cladding;
-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
- T. Clouser, Manager, Shift Operations
- J. Curtis, Radiation Protection Manager, Radiation and Industrial Safety
- 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
- 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
- P. Passalugo, SHAW, ISI Program Lead
- J. Patton, Supervisor, Quality Assurance
- K. Pitilli, Design Engineering Analysis Engineer
- L. Pope, System Engineer
- H. Quach, AREVA, Principal Engineer
- W. Reppa, JET Manager
- 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