IR 05000362/2010009
| ML111300448 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/10/2011 |
| From: | Ryan Lantz NRC/RGN-IV/DRP/RPB-D |
| To: | Peter Dietrich Southern California Edison Co |
| References | |
| IR-10-009 | |
| Download: ML111300448 (22) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION May 10, 2011
SUBJECT:
SAN ONOFRE NUCLEAR GENERATING STATION - UNIT 3 STEAM GENERATOR REPLACEMENT PROJECT INSPECTION REPORT NO. 05000362/2010009
Dear Mr. Dietrich:
On March 30, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed a steam generator replacement inspection at your San Onofre Nuclear Generating Station, Unit 3 facility.
The enclosed inspection report documents the inspection findings which were discussed on March 30, 2011, with Mr. Rich St. Onge, Director of Nuclear Regulatory Affairs and other members of your staff. This inspection report is applicable to Unit 3 only.
The inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
Based on the results of this inspection, no findings of significance were identified. However, a licensee-identified violation which was determined to be of very low safety significance is listed in this report. The NRC is treating this violation as a noncited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy because of the very low safety significance of the violation and because it is entered into your corrective action program. If you contest the violation or the significance of the noncited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the San Onofre Nuclear Generating Station facility.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure(s), and your response if you choose to provide one, will be made available
Southern California Edison -2-electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy or proprietary, information so that it can be made available to the Public without redaction.
Sincerely,
/RA/
Ryan E. Lantz, Chief Project Branch D Division of Reactor Projects Docket No. 50-362 License No. NPF-15 Enclosure:
NRC Inspection Report 05000362/2010009 w/Attachment: Supplemental Information Distribution via Listserv
Southern California Edison -3-Electronic distribution by RIV:
Regional Administrator (Elmo.Collins@nrc.gov)
Deputy Regional Administrator (Art.Howell@nrc.gov)
DRP Director (Kriss.Kennedy@nrc.gov)
DRP Deputy Director (Troy.Pruett@nrc.gov)
DRS Director (Anton.Vegel@nrc.gov)
Senior Resident Inspector (Greg Warnick@nrc.gov)
Resident Inspector (John.Reynoso@nrc.gov)
Branch Chief, DRP/D (Ryan.Lantz@nrc.gov)
Senior Project Engineer, DRP/D (Don.Allen@nrc.gov)
SONGS Administrative Assistant (Heather.Hutchinson@nrc.gov)
Project Engineer, DRP/D (Zachary.Hollcraft@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Public Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Randy.Hall@nrc.gov)
Branch Chief, DRS/TSB (Michael.Hay@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
OEMail Resource ROPreports RIV/OEDO ET (Stephanie.Bush-Goddard@nrc.gov)
R:\_REACTORS\_SONGS\2011\SO2010009-SGR-JPR.docx ADAMS: No Yes SUNSI Review Complete Reviewer Initials:
Publicly Available Non-Sensitive Non-publicly Available Sensitive RI:RPBD SRI:RPBD RI: R-II/DCP/CPB1 C:RPBD JReynoso GWarnick AMasters RLantz
/RA/DAllen for /RA/DAllen for /RA/ /RA/
4/28/11 4/28/11 5/9/11 5/10/11 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 50-362 License: NPF-15 Report: 05000362/2010009 Licensee: Southern California Edison Co. (SCE)
Facility: San Onofre Nuclear Generating Station, Unit 3 Location: 5000 S. Pacific Coast Hwy San Clemente, California Dates: October 10, 2010 through March 11, 2011 Inspectors: S. Achen, Reactor Inspector I. Anchondo, Reactor Inspector S. Makor, Reactor Inspector A. Masters, Senior Construction Inspector, Region II J. Reynoso, Resident Inspector G. Warnick, Senior Resident Inspector M. Young, Resident Inspector Approved By: Ryan E. Lantz Chief, Project Branch D Division of Reactor Projects-1- Enclosure
SUMMARY OF FINDINGS
IR 05000362/2010009; 10/10/2010 - 03/30/2011; San Onofre Nuclear Generating
Station, Unit 3 Steam Generator Replacement Report; Steam Generator Replacement Activities.
The report covered a 4-month period of inspection by resident and regional inspectors. No findings of significance were identified. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,
Revision 4, dated December 2006.
NRC-Identified Findings and Self-Revealing Findings
No findings of significance were identified.
Licensee-Identified Violations
A violation of very low safety significance, identified by the licensee, has been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and corrective action tracking numbers are listed in Section 4OA7.
REPORT DETAILS
Summary of Plant Status
Unit 3 began the inspection period shutdown for a scheduled refueling outage (U3R16) and steam generator replacement. The Unit 3 refueling outage was completed on February 18, 2011, and the plant returned to full power on March 3,
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R18 Permanent Modification
a. Inspection Scope
The inspectors observed and reviewed key affected parameters associated with cable splicing after Unit 3 containment restoration. The inspectors reviewed cable splicing procedures and observed cable splicing, cable tray installation and testing. Inspectors observed technicians performing a review of the procedure steps during the installation of the environmental qualification splices.
The inspectors verified that: modification preparation, staging, and implementation of the splicing activities did not impair emergency/abnormal operating procedure actions, key safety functions, or operator response to loss of key safety functions; post-modification testing will maintain the plant in a safe configuration during testing by verifying that unintended system interactions will not occur; systems, structures and components performance characteristics still meet the design basis; the modification design assumptions were appropriate; the modification test acceptance criteria will be met; and licensee personnel identified and implemented appropriate corrective actions associated with permanent plant modifications.
Specific documents reviewed during this inspection are listed in the attachment.
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA5 Other Activities
.1 Design and Planning Inspections
a. Inspection Scope
This inspection report documents inspection activities related to the San Onofre Nuclear Generating Station, Unit 3, steam generator replacement project.
These steam generator replacement inspection activities are not part of the normal baseline inspection program, but are performed on an as-needed basis. Therefore, no sample size is specified. The inspectors completed the applicable portion of Inspection Procedure IP 50001, Steam Generator Replacement Inspection, including the post installation verification and testing inspections.
Engineering and Technical Support Inspections to review engineering and technical support activities were performed prior to, and during, the steam generator replacement outage by resident and regional inspectors. Inspectors reviewed key design aspects and modifications associated with steam generator replacement. The inspectors reviewed the root cause evaluation and the extent of condition of replacement steam generator divider plate-to-channel head weld joint cracking.
The inspectors also reviewed safety analysis changes, permanent plant modifications (engineering change packages) and documentation, including safety screens and evaluations, to verify that they were performed in accordance with regulatory requirements and plant procedures. The inspectors also reviewed manufacturer records of parts and tubes of the replacement steam generators and reviewed preservice baseline eddy current examination results of new tubes. Additional activities were also performed during this inspection and applicable inspections are documented in NRC Inspection Report 05000361/2010005, 05000362/2010005; Sections 1R08, 1R18 and 1R20.
Specific documents reviewed during this inspection are listed in the attachment.
Radiation Protection Program An inspection to review radiation protection controls was performed during the steam generator replacement outage by regional inspectors. The results of the inspection are documented in NRC Inspection Report 05000361/2010005, 05000362/2010005; Sections 2RS01 and 2RS02.
Security Considerations and Adverse Impact to Other Unit The inspectors made frequent observations of security practices to verify that the licensee provided appropriate support for affected vital and protected area barriers during outage activities. The inspectors also checked for potential adverse impacts to Unit 2 (the nonoutage unit) caused by outage activities, equipment configurations, etc.
The inspectors reviewed steam generator replacement activities associated with risk management to minimize any adverse impact on the operating unit and common systems. Additional activities were also performed during this inspection and applicable inspection results are documented in NRC Inspection Report 05000361/2010005, 05000362/2010005; Section 1R13.
Specific documents reviewed during this inspection are listed in the attachment.
b. Findings
No findings were identified.
.2 Steam Generator Removal and Replacement Inspections
a. Inspection Scope
The inspectors used the guidance in Inspection Procedure 50001 to select and perform the following inspection activities, consistent with the safety significance and inspection resources and as part of normal baseline inspection efforts.
Welding and Nondestructive Examination Activities Inspectors reviewed or observed the following welding and nondestructive examination activities during the steam generator replacement outage:
- Qualification certifications for the Non-Destructive Examinations (NDE)examiners
- Contractors and Southern California Edison processes to verify they meet ASME code requirements
- Radiography films for several large bore pipe welds to verify the welds meet quality requirements
- Cable tray restoration activities, including post maintenance testing
- Results of mechanical snubber functional testing and associated nonconformance reports and nuclear notifications for safety related snubbers
- Review of the NDE of replacement tubes pre-service inspection results and baseline eddy current examination of tubes These activities are documented in this report and in NRC Inspection Report 05000361/2010005, 05000362/2010005; Section 1R08.
Specific documents reviewed during this inspection are listed in the attachment.
Containment Opening Restoration Activities The inspectors reviewed licensee activities related to construction activities associated with material, design, fabrication, installation, examination and testing of the containment temporary opening and restoration.
The inspectors completed the following inspection activities:
- Observation of restoration activities and review of the modification packages related to equipment hatch supports, containment liner and containment reinforcement bars
- Verification of the ASME code versions and sections to ensure compliance and correct application to the code and other industry standards
- Review of key containment design aspects found in the updated final safety analysis report to confirm there are no deviations from the safety analysis and licensing basis, including review of 10 CFR 50.59 screenings and evaluations
- Review of the liner plate welds strength analysis to verify that the restored liner will be as strong as the original uncut liner
- Review of the containment concrete pour analysis to ensure there will be no adverse impact on the liner plate
- Review of the design requirements of liner plate stiffeners to ensure the design requirements were adequate to withstand the loads imposed during concrete placement operations
- Visual inspections of liner plate stiffeners to confirm proper structural design requirements were implemented
- Confirming the use of dedicated new anchor heads during tendon restoration or if reused anchor heads were placed in service that material integrity was verified by inspection
- Confirmed that new tendon strands were used and adequate to support containment structural requirements
- Review of the tendon duct or sheathing restoration process to ensure the tendon activities were adequate, including grease fill and retensioning, to ensure containment structural integrity was not impacted, including visual inspections of the installed vertical and horizontal replacement tendon sheathing to confirm leak tightness
- Review of the engineering evaluations associated with containment opening repairs to re-enforcement bars including CadWeld joint design and weld splices
- Confirmed use of weld splices in place of CadWelds satisfied code requirements to ensure all aspects of the original containment structural design were met
- Review of the CadWeld purchasing, installation and testing specifications
- Observations of CadWeld sister splices in the field including verification of sister splices made by each welder for the CadWeld splices (under the same conditions and location) and subsequent tensile-testing to ensure adequate tensile strength of the joint was achieved
- Concrete batch plant operations including material storage and handling of concrete components
- Material test results (cement, fine and coarse aggregate, water, and admixtures)
- Concrete mix and proportion data, including batching results
- Concrete transportation and placement Relative to installation of concrete, the inspectors witnessed placement of concrete in the containment wall to restore the temporary construction opening. The inspectors examined the reinforcing steel to ensure it was installed in accordance with design requirements, was properly cleaned, and observed the concrete forms to ensure tightness and cleanliness. The inspectors reviewed placement activities to ensure that activities pertaining to concrete delivery time, free fall, flow distance, layer thickness and concrete consolidation conformed to industry standards established by the American Concrete Institute (ACI). Concrete batch tickets were examined to ensure that the specified concrete was being delivered to the site. The inspectors also observed testing of the plastic concrete for slump and temperature, including test preparation and molding of the concrete cylinders. Reviews were performed to ensure concrete testing was performed and the cylinders were molded in accordance with applicable American Society Testing and Materials (ASTM) requirements. In addition, the inspectors reviewed activities to ensure that concrete testing was performed by qualified personnel and that concrete placement activities were continuously monitored by licensee and contractor quality control and quality assurance personnel.
The inspectors examined the concrete batch plant to verify proper storage and separation of materials and temperature controls. The inspectors reviewed results of quality control acceptance testing performed on materials (cement, fine and coarse aggregate, and admixtures) used for batching. The inspectors also reviewed records documenting inspection of the concrete batch plant and the concrete truck mixers.
Activities were reviewed to determine if the contractors inspection of the trucks and batch plant were performed in accordance with the guidance of the National Ready Mix Concrete Association, if the batch plant scales were calibrated in accordance with National Ready Mix Concrete Association recommendations, and if mixer efficiency tests were performed on the truck mixers in accordance with ASTM C-94. The inspectors reviewed the concrete mix data to ensure that mix proportions for delivered concrete were selected based on trial concrete mix results, that quality control acceptance criteria
for the plastic concrete were based on the trial mixes, and that the trial mix met concrete strength requirements.
Specific documents reviewed during this inspection are listed in the attachment.
Lifting, Rigging and Steam Generator Movement and Reconnection Activities The inspectors observed and reviewed activities throughout the refueling outage associated with heavy lifting and rigging. The inspectors observed the implementation and reviewed documentation related to several structural modifications associated with the heavy lifting activities.
The inspectors also observed and reviewed engineering evaluations concerning the removal and reinstallation of the following structural modifications:
- Construction of the outside lift system and runway
- Lifting and rigging preparations associated with old steam generators removal
- Interference removal and replacement of new steam generators
- Temporary handling equipment construction and removal
- Structural supports to facilitate steam generator replacement The results of these inspections are documented in NRC Inspection Report 05000361/2010005, 05000362/2010005; Section 1R18.
The following activities were also reviewed:
- Reactor cavity decking construction and removal
- Movement and reconnection of replacement steam generators
- Steam generator hold down / skirt bolts material condition
- Transfer of old steam generators to temporary storage Specific documents reviewed during this inspection are listed in the attachment.
Outage Operating Conditions The inspectors used Inspection Procedure 71111.20 to verify proper outage conditions such as defueling, reactor coolant drain down, system isolation and equipment tagging.
The results of these inspections are documented in NRC Inspection Report 05000361/2010005, 05000362/2010005; Section 1R20.
Post-Installation Verification and Testing Selective inspections and reviews were conducted by the inspectors on steam generator post-installation verification and testing including:
- Reactor coolant system leakage testing and water inventory balance
- Steam Generator secondary side leakage testing
- Hot gap measurement for replacement steam generator snubbers and restraints
- Distributed control system functional testing main turbine and feedwater controls Specific documents reviewed during this inspection are listed in the attachment.
Containment Testing The inspectors performed system walk downs, valve alignment checks, and observed the initiation, and conclusion of the Unit 3 containment integrated leakage rate test. The inspectors reviewed the test procedure and verified that the potential leakage from containment at the design basis accident pressure remained within the limits stated in technical specifications. The results of these inspections are documented in NRC Inspection Report 05000361/2011002, 05000362/2011002; Sections 1R04 and 1R19.
Steam Generator Secondary Side Leakage The inspectors reviewed the secondary side leakage test results, and conducted a visual inspection of the steam generators secondary side after the steam generators were filled to check for leakage.
Specific documents reviewed during this inspection are listed in the attachment.
Calibration and Testing of Instrumentation The inspectors observed and evaluated the calibration and testing of instrumentation for both the primary and secondary side systems impacted by the steam generator replacement in the following areas:
- Functional testing of spliced resistance temperature detector to reactor coolant system hot leg loops. These inspections are documented in NRC Inspection Report 05000361/2011002, 05000362/2011002; Section 1R20
- Calibration and testing of steam generator E089, wide range water level transmitters following steam generator replacement. These inspections are documented in NRC Inspection Report 05000361/2011002, 05000362/2011002; Section 1R19
- Setpoints and calibration changes to replacement steam generator level instruments. These inspections are documented in NRC Inspection Report 05000361/2011002, 05000362/2011002; Section 1R19 Specific documents reviewed during this inspection are listed in the attachment.
Foreign Materials Control The inspectors performed frequent observations of the steam generator replacement activities to verify the licensee was implementing proper foreign materials controls. In particular, the inspectors observed controls related to reactor coolant system and secondary side openings.
Specific documents reviewed during this inspection are listed in the attachment.
Temporary Services The inspectors reviewed the work package and drawings, and then observed the installation, use, and removal of temporary services in the containment building during the outage. Instructions for the use and controls for construction power, acetylene, oxygen, and argon were reviewed, and the actual installation of each system was compared to the approved system sketches.
Specific documents reviewed during this inspection are listed in the attachment.
Startup and Power Ascension The inspectors reviewed the startup and power ascension procedures to determine if the procedure adequately verified proper performance of the components affected by steam generator replacement and outage maintenance activities. The results of these inspections are documented in NRC Inspection Report 05000361/2011002, 05000362/2011002; Section 1R20.
Specific documents reviewed during this inspection are listed in the attachment.
b. Findings
1. Adequacy of Model Inputs Used in Restoration of Nuclear Concrete Containment
Structures
Introduction.
The inspectors identified an unresolved item regarding the licensees engineering modeling inputs related to restoration of Unit 2 and 3 containment buildings.
During the inspection, insufficient information was available to determine if newer industry standards are appropriate to apply in the licensee finite element modeling of concrete stresses. The inspectors also questioned if the alternate modeling was allowed by NRC approved codes. The licensee considered using the inputs from the newer industry standards as an analytical refinement and not a methodology change. The inspectors could not determine if the licensee properly applied the Title 10 CFR 50.59 process to conclude the new analysis did not involve a departure from approved methods of evaluation. To address the concern, a Technical Interface Agreement (TIA 2011-008) was initiated with the NRC Office of Nuclear Reactor Regulation.
Discussion. During performance of the inspection, the inspectors reviewed three related engineering calculations and the screening required by 10 CFR 50.59 associated with the SONGS Unit 3 containment restoration. The licensee performed calculations and
evaluations of the structural integrity of the restored containment building. The calculations reviewed in conjunction with the inspection referenced models and equations from two contemporary reports; ACI 209R-92, Prediction of Creep, Shrinkage, and Temperature Effects in Concrete Structures, and ACI 224.2R-92, Cracking of Concrete Members in Direct Tension. These two ACI reports were not referenced in the licensees concrete construction code of record, which is ACI 318-71, Building Code Requirements for Reinforced Concrete, and BC-TOP-5, Prestressed Concrete Nuclear Reactor Containment Structures. These reports were also not referenced in the licensees final safety analysis report. The NRC determined the licensees 10 CFR 50.59 evaluation did not address the referenced models and equation inputs from ACI 209R-92 and ACI 224.2R-92. Upon further questioning by the NRC about the appropriate use of the inputs, engineering personnel concluded that inputs from these reports did not represent a change in methodology to the original approved evaluation, in part, because the inputs were needed to address cracking, as required per the original analysis and was considered a calculation refinement. Pending review of documentation to determine if inputs from these reports do represent a change in methodology as described in the 10 CFR 50.59 evaluation, this issue will remain an Unresolved Item (URI) and tracked as URI 05000362/2010009-01; Adequacy of Model Inputs Used in Restoration of Concrete Containment Structures.
4OA6 Meetings
Exit Meeting Summary
On January 20, 2011, the inspectors presented a brief of the team inspection results to Mr. Mike Wharton, Manager, Steam Generator Replacement Project, and other members of the licensees staff.
On March 11, 2011, the inspectors presented inspection results to Mr. Craig Harberts, Manager, Steam Generator Replacement Project, and other members of the licensees staff.
On March 30, 2011, the inspectors presented the final inspection results to Mr. Rich St.
Onge, Director, Nuclear Regulatory Affairs.
The licensee acknowledged the inspection results and observations presented.
Some proprietary information was reviewed during this inspection but no proprietary information was included in this report.
4OA7 Licensee-Identified Violations
The following finding of very low safety significance was identified by the licensee and is a violation of NRC requirements which met the criteria of Section 2.0 of the NRC Enforcement Policy, for being dispositioned as a noncited violation.
Contrary to 10 CFR, Part 50, Appendix B, Criterion III, Design Control, the licensee failed to assure that design bases were correctly translated into
instructions for adjusting steam generator snubbers to their cold settings. On December 28, 2010, after entry into Mode 6, engineering discovered three of the four steam generator snubbers outside the design cold settings. The snubbers had been adjusted without consideration of the reactor coolant temperature. A subsequent change in reactor coolant temperature caused the snubbers to contract beyond the cold set allowance. The snubbers were reworked and adjusted to acceptable cold settings. The finding was determined to be of very low safety significance because it did not increase the likelihood of a loss of reactor coolant system inventory or degrade the licensees ability to add reactor coolant inventory when needed. The issue was entered into the licensees corrective action program as Nuclear Notifications NNs 201260722 and 201262831.
ATTACHMENT: Supplemental Information
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- D. Axline, Technical Specialist, Nuclear Regulatory Affairs
- D. Bauder, Vice President and Station Manager
- D. Calhoun, Senior Nuclear Engineer
- D. Czapski, Civil Inspector Level III
- C. Harberts, Manager, Steam Generator Replacement Project
- S. Hetherington, Civil Superintendent, Bechtel
- E. Hubley, Director, Maintenance & Construction Services
- L. Kelly, Engineer, Nuclear Regulatory Affairs
- B. Kotteakos, Manager, Vendor Oversight
- M. Lewis, Manager, Health Physics
- A. Matheny, System Engineer
- T. McCool, Plant Manager
- A. Meichler, Mechanical/System Engineering Supervisor
- M. Mihalik, Areva Project Manager, Steam Generator Replacement Project
- R. Nielsen, Supervisor, Nuclear Oversight
- B. Power, Operations Manager, Catalina Pacific
- C. Ryan, Manager, Maintenance & Construction Services
- R. St. Onge, Director, Nuclear Regulatory Affairs
- D. Schaffer, Senior Nuclear Engineer
- D. Todd, Manager, Site Projects Oversight
- M. Wharton, Manager, Steam Generator Replacement Project
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000362/2010009-01 URI Adequacy of Model Inputs Used in Restoration of Nuclear Concrete Containment Structures
Opened and Closed
None Attachment