ML093441035
| ML093441035 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/22/2009 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Ridenoure R Southern California Edison Co |
| Hall, J R, NRR/DORL/LPL4, 301-415-4032 | |
| References | |
| TAC ME0768, TAC ME0769 | |
| Download: ML093441035 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 22, 2009 Mr. Ross T. Ridenoure Senior Vice President and Chief Nuclear Officer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128
SUBJECT:
SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 - RELIEF REQUEST ISI-3-29, REQUEST FOR RELIEF FROM INSPECTION REQUIREMENTS OF ASME CODE CASE N-729-1 FOR CONTROL ELEMENT DRIVE MECHANISM PENETRATIONS (TAC NOS. ME0768 AND ME0769)
Dear Mr. Ridenoure:
By letter dated February 27,2009, as supplemented by letter dated October 2,2009, Southern California Edison Company (SCE, the licensee) submitted Relief Request ISI-3-29, requesting relief from the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(6)(ii)(D), for the third 10-year inservice inspection (lSI) interval for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3. The submittal requests U.S. Nuclear Regulatory Commission (NRC) approval for relief from the inspection requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section XI, ASME Code Case N-729-1, Figure 2, as conditionally required by 10 CFR 50.55a(g)(6)(ii)(D), for the examination of reactor pressure vessel upper head penetrations for the control element drive mechanisms (CEDMs).
SCE requested approval of Relief Request ISI-3-29 by December 26,2009, in order to support the start-up of the Unit 2 reactor following completion of the 2009 refueling outage. The NRC staff has reviewed the licensee's relief request and finds the proposed alternative examination acceptable, as documented in the enclosed safety evaluation. The staff concludes that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii), and that the alternative provides reasonable assurance of structural integrity of the CEDM nozzles. Therefore, pursuant to 10 CFR50.55a(a)(3)(ii), the NRC staff authorizes the use of the alternative described in Relief Request ISI-3-29 at SONGS, Units 2 and 3 for the duration of the third 10-year lSI interval, which is scheduled to end on August 17, 2013.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
R. Ridenoure
- 2 A copy of the related Safety Evaluation is enclosed. If you have any questions, please contact Randy Hall at (301) 415-4032 or via e-mail at randy.hall@nrc.gov.
Sincerely,
/J.-I ~ A.-rJ;j-Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-361 and 50-362
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST ISI-3-29 SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 SOUTHERN CALIFORNIA EDISON COMPANY DOCKET NOS. 50-361 AND 50-362
1.0 INTRODUCTION
By letter dated February 27, 2009, as supplemented by letter dated October 2, 2009, (Agencywide Documents Access and Management System (ADAMS) Accession Nos.
ML090620358 and ML092790153, respectively), Southern California Edison Company (SCE, the licensee) submitted a request for relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, ASME Code Case N-729-1, for the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3. The information provided by the licensee in support of the request for relief from ASME Code requirements has been evaluated by the U.S. Nuclear Regulatory Commission (NRC) staff and the basis for disposition is documented below.
2.0 REGULATORY EVALUATION
The inservice inspection (lSI) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," and the applicable edition and addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i). 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1O-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The lSI Code of record for the Enclosure
- 2 third 1O-year lSI interval for SONGS, Units 2 and 3, which started in August 2003 and is scheduled to be complete in August 2013, is the 1995 Edition through the 1996 Addenda of Section XI of the ASME Code.
Pursuant to 10 CFR 50.55a(g)(6)(ii)(D), all licensees of pressurized-water reactors shall augment their lSI program with ASME Code Case N-729-1 subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) by December 31, 2008. Once a licensee implements this requirement, the First Revised Order EA-03-009 no longer applies to that licensee and shall be deemed to be withdrawn.
3.0 TECHNICAL EVALUATION
3.1 System/Component(s) for which Relief is Requested SONGS Unit 2:
Item No. 84.20, Ninety-one (91) Control Element Drive Mechanism (CEDM) penetrations - [Reactor Pressure Vessel Head (RPVH)
Penetrations 1 through 91]
SONGS Unit 3:
Item No. 84.20, Ninety-one (91) Control Element Drive Mechanism (CEDM) penetrations - [Reactor Pressure Vessel Head Penetrations 1 through 91]
3.2 Applicable Code Requirements Relief is being requested from the inspection requirement for the base metal examination volume in Figure 2 of ASME Code Case N-729-1.
3.3 Licensee's Proposed Alternative The licensee seeks relief from Code Case N-729-1 where inspection coverage is limited by inaccessible areas of each of the 91 CEDM penetration nozzles for SONGS, Units 2 and 3.
SCE proposes to examine each CEDM nozzle, the inspection coverage requirement of dimension "au in Code Case N-729-1, Figure 2, from above the top of the attachment weld to as far down the nozzle as physically possible. The required coverage distance per Code Case N-729-1 is 1.5 inches for Incidence Angle e s 30° and for all nozzles greater than 4.5-inch outside diameter (OD), or 1 inch for Incidence Angle e) 30°, or to the end of the tube, whichever is less. The licensee stated the distance shall be at least the minimum inspection distance below the bottom of the attachment weld as stated in Table 1.
Table 1: Proposed Minimum Inspection Distance for CEDM Penetration Nozzles CEDM penetration nozzle number(s)
Proposed coverage below the J-weld for non-circumferential flaws (in)
Proposed coverage below the J-weld for all flaw orientations (in) 1 0.44 0.19 2 through 35 0.43 0.18 36 through 87 0.42 0.17 88 through 91 0.35 0.10
- 3 3.4 Licensee's Basis for Requesting Relief The licensee stated the material near the bottom of each CEDM nozzle cannot be inspected due to the presence of a CEDM extension shaft guide cone threaded to the inside diameter (ID) surface of the nozzle. Compliance with the Code Case requirement requires the reactor vessel head to be redesigned. SONGS has ordered replacement heads for both Units and plans to have them installed during the Cycle 17 refueling outages, currently scheduled for the fall of 2011 and 2012, respectively. SCE is working with the manufacturer of the new head to incorporate design changes that would improve the area of inspection coverage in order to meet the requirements of ASME Code Case N-729-1.
In a letter dated February 9, 2004 (ADAMS Accession No. ML040480386), the licensee provided a response to a previous NRC request for additional information that was related to a prior relaxation request dated December 9,2003 (ADAMS Accession No. ML033450462),
regarding the CEDM extension shaft guide cone threads, which was approved by the NRC staff on March 19, 2004 (ADAMS Accession No. ML040860761). The licensee stated that guide cones are threaded into the ID of all 91 CEDM penetrations. The guide cone threads are staked with a set screw which is plug welded to preclude unthreading of the cone during operation. In addition, the licensee stated that there are two 1-inch-long fillet welds between the top of the tapered portion of the guide cone and the bottom of the CEDM nozzles. The licensee stated that removal of each guide cone would require destructive removal of three welds and the stake, then unthreading the guide cone. The licensee estimated the time to perform this labor would be at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per guide cone in a radiation field of approximately 4 roentgen equivalent man per hour (rem/hr).
In addition, recent qualification of the licensee's examination contractor has established a new limitation against using circumferentially oriented Time of Flight Diffraction (TOFD) for detection of circumferential flaws which impacts the lower extent to w~lich fully qualified examination coverage can be credited. This is due to the loss of axial TOFD transducer coupling approximately 0.250 inches above the CEDM counterbore.
3.5 Duration of Proposed Alternative The proposed alternative will apply to the existing RPVH for the remainder of the current SONGS, Units 2 and 3, third 1O-year lSI interval. The third 1O-year interval began on August 18, 2003, and is scheduled to end on August 17, 2013.
3.6 Staff Evaluation The NRC staff's review of this request was based on 10 CFR 50.55a(a)(3)(ii) which states that:
Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Within the context of the licensee's proposed alternative examination of the CEDM penetration nozzles, the licensee has demonstrated the hardship that would result from implementing
- 4 examinations to the bottom end of these nozzles. The licensee estimated the dose to be four person-rem per nozzle to remove the guide cones. The licensee also stated additional dose and time would be required to replace the guide cones. The licensee stated that grinding and drilling operations required to remove the guide cones would degrade the CEDM penetrations with respect to primary water stress-corrosion cracking (PWSCC) resistance. Further, any alternative surface examinations to compensate for the physical limitations to obtain full volumetric inspection coverage would need to be manually performed in a 4 rem/hr radiological close field.
The phenomenon of concern is PWSCC in the J-groove welds and the nozzles, which normally initiates in areas of highest stress. The highest residual stresses in CEDM penetrations are found in the area adjacent to the J-groove attachment weld. Therefore, it is most likely that PWSCC will initiate in this area.
SONGS previously performed analysis to characterize the potential growth of postulated cracks in the uninspected areas. These results were given in Westinghouse Report WCAP-15819, Revision 1, "Structural Integrity Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: San Onofre Units 2 and 3" (ADAMS Accession Nos.
ML040500598 and ML040500602). The minimum inspection distance below the weld that was approved by the NRC staff in a previous safety evaluation dated June 27,2005 (ADAMS Accession No. ML051780416), and is proposed for each CEDIVI nozzle is based on the Appendix C curves provided in WCAP-15819, Revision 1 (Appendix C curves). The staff reviewed several conservatisms that were used in the flaw evaluation and are described in the following paragraphs.
The NRC staff concludes that the methodology used in the development of the Appendix C curves, which postulated the initial crack extending from the expected lower extent of the inspection coverage area to the point where hoop stresses on either the ID or the OD becomes compressive, is conservative. Further, the staff concludes these Appendix C crack growth curves use design weld sizes, which are conservative compared to the as-built weld sizes.
Finally, the staff concludes that the crack growth rate used to develop the Appendix C curves is based on the 75th percentile/95th confidence curve of Electric Power Research Institute report, "Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Material (MRP-55)," dated July 18, 2002 (ADAMS Accession No. ML023010510), was used. The crack growth model equation used had an identical form to that in MRP-55 and is conservative. Therefore, the staff concludes that the Appendix C curves are acceptable to show reasonable assurance of conservative crack growth in the CEDM nozzles.
The staff concludes that these Appendix C curves support that a through-wall axial crack growing from the minimum distance inspected for each CEDM below the top of the weld would take at least one operating cycle to reach the bottom of the weld. This does not include the time that would be required for an axial crack to propagate through the attachment weld and result in a leakage path. Therefore, the staff concludes that given the licensee's proposed axial flaw minimum inspection coverage below the J-groove weld, the Appendix C curves provide reasonable assurance of structural integrity for at least one operational cycle.
- 5 The licensee's previously requested relief dated February 27,2009, was based on coverage provided using only the circumferentially oriented TOFD transducer pair. However, qualification of the examination contractor established a new limitation in using circumferentially oriented TOFD for detection of circumferential flaws. The axial TOFD transducer is used for detection of circumferential flaws; however, coverage below the J-weld using the axially oriented TOFD pair is more limited. This is because the lower transducer loses acoustic coupling approximately 0.250 inches above the counterbore. Based on the above, in its supplemental letter dated October 2, 2009, the licensee has requested relief from the inspection requirements of Figure 2 of Code Case N-729-1 for an additional 0.250 inches above the previously requested lower exam extent for circumferential flaws only.
In the case of a circumferential flaw in the uninspected area, the threaded portion of the extension shaft guide cone would serve to retain potential loose parts. A 360-degree through-wall crack in the narrow uninspected annulus above the guide cone threads could result in separation of the guide cone from the penetration. In such a case, the guide cone would be retained by the control element assembly shroud and associated extension shaft. The NRC staff concludes that a reasonable basis was given to show that this condition is highly unlikely to interfere with control element assembly function or any other reactor coolant system function, and would be observed in the subsequent refueling outage. Given the staff's analysis, the licensee's proposed minimum inspection distances below the J-groove weld for both axial and circumferential flaws, as listed in Table 1, are acceptable.
Based on the above, the NRC staff concludes that the licensee has demonstrated a sound basis for the relief and that the proposed alternative examination is acceptable as it provides reasonable assurance of the structural integrity of the reactor pressure vessel head.
Inspections to comply with Code Case N-729-1 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity of the sUbject components. Complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii), and the alternative provides reasonable assurance of structural integrity. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii), the NRC staff authorizes Relief Request ISI-3-29 at SONGS, Units 2 and 3, for the duration of the third 10-year lSI interval, which is scheduled to end on August 17, 2013.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: M. Audrain Date: December 22, 2009
R. Ridenoure
- 2 A copy of the related Safety Evaluation is enclosed. If you have any questions, please contact Randy Hall at (301) 415-4032 or via e-mail at randy. hall@nrc.gov.
Sincerely, IRN Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-361 and 50-362
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC RidsNrrPMSanOnofre Resource LPLIV r/f RidsNrrLAJBurkhardt Resource RidsAcrsAcnw_MailCTR Resource RidsOgcRp Resource RidsNrrDciCpnb Resource RidsRgn4MailCenter Resource RidsNrrDorlDpr Resource MAudrain, NRR/DCI/CPNB RidsNrrDorlLp/4 Resource LTrocine, EDO RIV ADAMS Accession No.: ML093441035
(*) Concurrence via SE OFFICE NRRlLPL4/PM NRRlLPL4/LA DCI/CPNB/BC NRRlLPL4/BC NRRlLPL4/PM NAME JRHall JBurkhardt TChan (*)
MMarkley JRHall DATE 12/18/09 12/17/09 12/04/09 12/22/09 12/22/09 OFFICIAL AGENCY RECORD