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
uly 6, 2007
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION - NRC INTEGRATED INSPECTION REPORT 05000445/2007006
Dear Mr. Blevins:
On May 21, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your 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. Flores and other members of your staff.
This inspection examined activities conducted under your licenses as they related to safety and compliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. 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 its enclosure will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs 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 them with you.
Sincerely,
/RA/
Claude Johnson, Chief Project Branch A Division of Reactor Projects
TXU Power -2-Docket Nos.: 50-445 License Nos.: NPF-87
Enclosure:
NRC Inspection Report 05000445/2007006 w/Attachment: Supplemental Information
REGION IV==
Dockets: 50-445 Licenses: 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, 2007 Inspectors: D. Allen, Senior Resident Inspector A. 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 Physicist Approved by: Claude Johnson, Chief, Project Branch A Division of Reactor Projects Attachment: Supplemental Information
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 regional reactor 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.
NRC-Identified and Self-Revealing Findings
No findings of significance were identified.
Licensee-Identified Violations
None.
REPORT DETAILS
Summary of Plant Status
Comanche Peak Steam Electric Station (CPSES) Unit 1 began the period operating at essentially 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.
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 licensees implementation of changes to 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 last NRC 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 to determine 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 safety evaluations and 12 applicability determinations and screenings (combined). The
inspectors completed review of one licensee safety evaluation and 2 applicability determinations 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
.1 Performance of Nondestructive Examination Activities Other Than Steam Generator
Tube Inspections, Pressurized Water Reactor Vessel Upper Head Penetration Inspections, 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/ Identity Examination Examination Component Type Method Reactor Coolant New Steam Generator (#3) Volumetric Ultrasonic System to Hot Leg Weld Reactor Coolant New Steam Generator (#3) Volumetric Ultrasonic System to Cold Leg Weld Auxiliary Feedwater Cap Weld FW-TUX-42 Volumetric Ultrasonic System Radiography Auxiliary Feedwater Cap Weld FW-TUX-36 Volumetric Radiography System Auxiliary Feedwater Cap Weld FW-TUX-38 Volumetric Radiography System Auxiliary Feedwater Cap Weld FW-TUX-7 Volumetric Ultrasonic System
System/ Identity Examination Examination Component Type Method Reactor Coolant New Steam Generator (#4) Volumetric Radiography System to Hot Leg Weld Reactor Coolant New Steam Generator (#4) Volumetric Radiography System to Cold Leg Weld Reactor Coolant New Steam Generator (#4) Surface Penetrant System to Hot Leg Weld Reactor Coolant New Steam Generator (#4) Surface Penetrant System to Cold Leg Weld Auxiliary Feedwater Cap Weld FW-TUX-7 Surface Magnetic System Particle Component Cooling Vertical Spring Can Visual Visual (VT-3)
Water System H1: CC-1-RB-049 Component Cooling Welded Attachment Visual Visual (VT-1)
Water System H1WA: CC-1-RB-049 Component Cooling Vertical Spring Can Visual Visual (VT-3)
Water System H1: CC-1-249-701-C53A For each of the observed nondestructive examination activities, the inspectors verified that 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 appropriate nondestructive 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 were dispositioned 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 recordable indications 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.
The inspection procedure further requires verification of one to three welds on Class 1 or 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.
.2 Reactor Vessel Upper Head Penetration Inspection Activities
a. Inspection Scope
The inspection requirements for this section parallel the inspection requirement steps in Section 02.01. The inspectors reviewed records of completed nondestructive examinations, including the eddy current and ultrasonic examination data analyses process used on the reactor vessel upper head penetrations during their preservice inspections.
Additionally, the nondestructive examination procedures used to perform the above examinations 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.
.3 Boric Acid Corrosion Control Inspection Activities (Pressurized Water Reactors)
a. Inspection Scope
The inspectors evaluated the implementation of the licensees boric acid corrosion control 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 control walkdown visual examination activities through either direct observation or record review. The inspectors reviewed the documentation associated with the licensees boric
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 containing boric acid during walkdowns of the containment building and the auxiliary building.
The inspection procedure requires verification that visual inspections emphasize locations where boric acid leaks can cause degradation of safety significant components. The inspectors verified through direct observation and program/record review that the licensees 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 evaluations performed 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.
.4 Steam Generator Tube Inspection Activities
a. Inspection Scope
The inspectors verified through records review that licensee personnel and contractors used 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 licensees 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 were new 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.
The inspectors completed one sample.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Evaluation
a. Inspection Scope
The inspectors reviewed selected activities regarding risk evaluations and overall plant configuration 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:
C Probability Risk Analysis Report related to the multiple crane operations inside the Unit 1 containment building during 1RF12, on February 23, 2007 C Defense in depth contingency plan, 1RF-22, for maintaining Unit 1 containment pressure while the containment liner is removed and fuel is being unloaded with 24 or less fuel assemblies remaining in the core, on February 26, 2007 The inspectors completed two samples.
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed eleven permanent plant modification packages and associated documentation, 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.
The inspectors evaluated the effectiveness of the licensees corrective action process to identify 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 of six permanent plant modifications. The inspectors completed review of eleven permanent plant modifications.
b. Findings
No findings of significance were identified.
1R19 Postmaintenance Testing
.1 Steam Generator and Reactor Vessel Head Replacement
a. Inspection Scope
The inspectors witnessed or reviewed the results of the postmaintenance tests for the following replacement outage activities:
C Control rod drive mechanism (CRDM) ventilation testing following the complete redesign 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, 2007 C CRDM testing following reactor vessel head replacement, in accordance with procedure IPO-011A, Plant Restart and Testing Following Steam Generator Replacement, Revision 0, reviewed on April 19, 2007 C Steam generator blowdown system flow and vibration testing following the replacement 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, 2007 C Transfer of feedwater bypass control to main feedwater control testing following maintenance 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, 2007 C Electrical load swing testing to ensure reactor control system interaction and tuning 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
C Steam generator steam flow calibration following replacement of steam generators, in accordance with procedure IPO-011A, Plant Restart and Testing Following Steam Generator Replacement, Revision 0, observed and reviewed on April 24, 2007 C Steam Generator Water Level Control System response testing following adjustments 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, 2007 C Large load (275 MWe) reduction test following replacement outage activities, in accordance with procedure IPO-011A, Plant Restart and Testing Following Steam Generator Replacement, Revision 0, observed and reviewed on April 30, 2007 C Reactor coolant system flow measurement test following the replacement of the Unit 1 steam generators, in accordance with procedure number INC-7018A, Reactor Coolant System Flow Measurement, Revision 3, reviewed on May 2, 2007 In each case, the associated work orders and test procedures were reviewed in accordance 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.
.2 Containment Alternate Access
Containment Integrated Leak Rate Test Procedure Review (70307)
a. Inspection Scope
The inspectors reviewed the licensees containment integrated leak rate test procedure to 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.
Containment Integrated Leak Rate Surveillance (70313)
a. Inspection Scope
The inspectors verified through observation, records review, and independent calculations 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 licensee had 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 licensees 1RF12 activities to ensure that risk was considered when developing and when deviating from the outage schedule, the plant configuration was controlled in consideration of facility risk, mitigation strategies were properly implemented, and Technical Specification requirements were implemented to maintain the appropriate defense-in-depth. The inspectors reviewed and/or observed the following items, listed below, as they pertained to the steam generator and reactor vessel head replacement. Coverage of the full scope of Inspection Procedure 71111.20 is documented in Inspection Reports 05000445/446-2007002 and 05000445/446-2007003.
C Unit shutdown and cooldown C Reduced reactor coolant inventory activities
C Defense in depth and mitigation strategy implementation C Containment closure capability C Refueling activities that included fuel offloading, fuel transfer, and core reloading C Implementation of procedures for foreign material exclusion C Electrical power source arrangement C Containment cleanup and inspection C Unit heatup and startup C Licensee identification and resolution of problems related to refueling activities
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications
a. Inspection Scope
The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR), plant drawings, 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.
C Unit 1 Containment Alternate Access, for steam generator and reactor vessel head replacement, in accordance with Final Design Authorization (FDA)
FDA-2005-000658-01-02, observed, and reviewed February 24, 2007 through April 6, 2007 The inspectors completed one sample.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control To Radiologically Significant Areas (71121.01)
a. Inspection Scope
This area was inspected to assess the licensees performance in implementing physical and 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 licensees 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 Radiologically Significant 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:
C Controls (surveys, posting, and barricades) of radiation, high radiation, or airborne radioactivity areas C Radiation work permits, procedures, engineering controls, and air sampler locations C Conformity of electronic personal dosimeter alarm set points with survey indications and plant policy; workers knowledge of required actions when their electronic personnel dosimeter noticeably malfunctions or alarms C Barrier integrity and performance of engineering controls in airborne radioactivity areas C Self-assessments, audits, licensee event reports, and special reports related to the access control program since the last inspection C Corrective action documents related to access controls C Licensee actions in cases of repetitive deficiencies or significant individual deficiencies C Radiation work permit briefings and worker instructions
C Adequacy of radiological controls, such as required surveys, radiation protection job coverage, and contamination control during job performance C Dosimetry placement in high radiation work areas with significant dose rate gradients C Changes in licensee procedural controls of high dose rate - high radiation areas and very high radiation areas C Controls for special areas that have the potential to become very high radiation areas during certain plant operations C Posting and locking of entrances to all accessible high dose rate - high radiation areas and very high radiation areas C Radiation worker and radiation protection technician performance with respect to radiation protection work requirements The samples completed for Inspection Procedure 71121.01 will be tracked in Section 2OS1 of NRC Inspection Report 05000445;446/2007003.
b. Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (71121.02)
a. Inspection Scope
The inspectors assessed licensee performance with respect to maintaining individual and collective radiation exposures as low as is reasonably achievable (ALARA). The inspectors used the requirements in 10 CFR Part 20 and the licensees 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, 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 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
- Site specific trends in collective exposures, plant historical data, and source-term measurements
- Site specific ALARA procedures
- ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements
- Intended versus actual work activity doses and the reasons for any inconsistencies
- Integration of ALARA requirements into work procedure and radiation work permit documents
- Shielding requests and dose/benefit analyses
- Post-job work activity reviews
- Assumptions and basis for the current annual collective exposure estimate, the methodology 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 unexpected changes in scope or emergent work were encountered
- Exposure tracking system
- Use of engineering controls to achieve dose reductions and dose reduction benefits afforded by shielding
- Workers use of the low dose waiting areas
- Records detailing the historical trends and current status of tracked plant source terms 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 exposure reduction initiatives
- Specific sources identified by the licensee for exposure reduction actions and priorities established for these actions, and results achieved against since the last refueling cycle
- Radiation worker and radiation protection technician performance during work activities in radiation areas, airborne radioactivity areas, or high radiation areas
- Self-assessments, audits, and special reports related to the ALARA program since October 2006
- Resolution through the corrective action process of problems identified through post-job reviews and post-outage ALARA report critiques
- Corrective action documents related to the ALARA program and follow-up activities such as initial problem identification, characterization, and tracking
- Effectiveness of self-assessment activities with respect to identifying and addressing repetitive deficiencies or significant individual deficiencies The samples completed for Inspection Procedure 71121.02 will be tracked in Section 2OS2 of NRC Inspection Report 05000445;446/2007003.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
.1 Inservice Inspection
a. Inspection Scope
The inspection procedure requires review of a sample of problems associated with inservice 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 dealt with inservice inspection activities and found that the corrective actions were appropriate. 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.
.2 Steam Generator and Reactor Vessel Head Replacement Inspection (50001, 71007)
a. Inspection Scope
The inspectors reviewed a sample of the problems identified and documented in the licensees 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,
that discussed issues identified in the contractors corrective action program, to ensure that items were entered into the licensees 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.
4OA5 Other Activities
.1 Steam Generator Replacement Inspection
Design and Planning Inspections (Section 02.02)
a. Inspection Scope
The inspectors used the guidance in Inspection Procedure 50001, Steam Generator Replacement Inspection, Section 02.02, and inspection procedures referenced therein, to perform the steam generator removal and replacement activities listed below.
Engineering and Technical Support The inspection activities specified by Section 02.02.a, Steam Generator Removal and Replacement 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 Rigging In accordance with Section 02.02.b, the inspectors reviewed the applicable engineering design, 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 Protection In accordance with Section 02.02.c, the inspectors reviewed radiation protection program 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.
Security Considerations and Adverse Impact to the Other Unit In accordance with Section 02.02.d, the inspectors interviewed security specialists and officers 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, and inspection procedures referenced therein, to perform the steam generator removal and replacement activities listed below.
Welding and Nondestructive Examination Activities In accordance with Section 02.03.a, inspections were conducted to review welding and NDE 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 Activities In accordance with Section 02.03.b, and Inspection Procedure 71111.23, Temporary Plant 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 installation Major Structural Modifications and Containment Access and Integrity In accordance with Section 02.03.c and .d and Inspection Procedures 71111.17 and 71111.23, 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.
- Complete removal of the upper steam generator snubber supports
- Temporary alternate containment access The testing activities for the repair and recovery of the containment building can be found in Sections 1R19, while more information concerning the temporary modification can be found in Section 1R23 of this report.
Unit 1 Outage Operating Conditions The inspectors used Section 02.03.e and Inspection Procedure 71111.20, Refueling and Outage Activities, to complete this inspection. Section 1R20 contains a more detailed explanation of what was observed and reviewed.
Radiation Protection Controls This inspection was performed during the outage by regional inspectors and the results are documented in Sections 2OS1 and 2OS2 of this report.
Foreign Material Controls The inspectors followed the guidance contained in Section 02.03.e. The inspectors reviewed and observed procedural controls, field observations, and the licensees 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 Services In accordance with Sections 02.03.e, the inspectors reviewed work orders, procedures and 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 Head Storage Facility In accordance with Section 02.03.f, the inspectors reviewed the licensees radiological safety 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.
Post-installation Verification and Testing Inspection (Section 02.04)
a. Scope
The inspectors used the guidance in Inspection Procedure 50001, Section 02.04, and inspection 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.
.2 Reactor Vessel Head Replacement Inspection
Design and Planning Inspections (Section 02.02)
a. Inspection Scope
The inspectors used the guidance in Inspection Procedure 71007, Reactor Vessel Head 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 Support The 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 Rigging In accordance with Section 02.02.b of Inspection Procedure 71007, the inspectors reviewed 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 Protection The 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,
- (6) emergency contingencies, and
- (7) project staffing and training plans are documented in 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 Unit In accordance with Section 02.02.d, the inspectors interviewed security specialists and officers 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, and inspection procedures referenced therein, to perform the following reactor vessel head fabrication inspection activities.
Heat Treatment The inspectors verified that the material heat treatment used to enhance the mechanical properties 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 Examination Inspections were conducted to ensure the manufacturing control plan included provisions for monitoring NDE, and to ascertain that the NDE was performed in accordance with applicable code, material specification, and contract requirements.
Welding The inspectors reviewed the documentation for the weld overlay welding operations that established 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;
requisitions, traceability of weld material rods, weld procedure qualification, welder qualifications, 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.
Procedures Inspections were completed to ensure that repair procedures had been established and that 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 Reconciliation The 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 Program Inspections were conducted to ensure that machining was carried out under a controlled system 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 manufacturers 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 Inspection The inspectors verified that the original ASME Code,Section III, data packages for the replacement 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.
Reactor Vessel Head Removal and Replacement (Section 02.04)
Lifting and Rigging Activities In accordance with Section 02.04.a, and Inspection Procedure 71111.23, Temporary Plant 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 set Major Structural Modifications This inspection was not applicable due to the lack of major structural modifications. The only 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 Integrity The inspection is documented in Sections 1R19, 1R23, and Section 4OA5.1.
Unit 1 Outage Operating Conditions The inspectors used Section 02.04.d and Inspection Procedure 71111.20, Refueling and Outage Activities, to complete this inspection. Section 1R20 contains a more detailed explanation of what was observed and reviewed.
Radiation Protection Controls This inspection was performed during the outage by regional inspectors and the results are documented in Sections 2OS1 and 2OS2 of this report.
Foreign Material Controls The inspectors followed the guidance contained in Section 02.04.d. The inspectors reviewed and observed procedural controls, field observations, and the licensees 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 Services In accordance with Sections 02.04.d, the inspectors reviewed work orders, procedures and 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 Head Storage Facility In accordance with Section 02.04.e, the inspectors reviewed the licensees radiological safety 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, and inspection 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.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On February 9, 2007, the inspectors presented the safety evaluation and permanent plant 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, Steam Generator 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.
On May 21, 2007, the inspectors presented the resident inspection results to Mr. 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
Attachment
- D. OConnor, 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
None
Opened and Closed
None
Closed
None
Discussed
None Attachment