ML070310154

Relief Request for Relief from Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Concerning Reactor Vessel Head Penetration Repairs

ML070310154
Person / Time
Site:	San Onofre
Issue date:	02/15/2007
From:	Terao D NRC/NRR/ADRO/DORL/LPLIV
To:	Rosenblum R Southern California Edison Co
Kalyanam N, NRR/DORL, 415-1480
References
TAC MD1714
Download: ML070310154 (10)

Text

February 15, 2007 Mr. Richard M. Rosenblum 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, UNIT 3 - RE: REQUEST FOR RELIEF FROM REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE CONCERNING REACTOR VESSEL HEAD PENETRATION REPAIRS (TAC NO. MD1714)

Dear Mr. Rosenblum:

By letter dated May 11, 2006, as supplemented by letters dated August 15 and October 20, 2006, Southern California Edison Company (SCE, the licensee) submitted Relief Request ISI-3-22 to use the inside diameter (ID) weld inlay repair process as an alternative to requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for the repair of Reactor Vessel Head Penetration (RVHP) Control Element Drive Mechanism (CEDM) No. 56 at San Onofre Nuclear Generating Station, Unit 3.

The Nuclear Regulatory Commission (NRC) staff has completed its review of this submittal and concludes that the licensees proposed use of the embedded flaw ID weld inlay repair process on RVHP CEDM No. 56 provides an acceptable level of quality and safety for one operational cycle during the third 10-year inservice inspection. The NRC staff authorizes the alternative

R. Rosenblum proposed by SCE in accordance with 50.55a(a)(3)(i) of Title 10 of Code of Federal Regulations.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

The staff's safety evaluation is enclosed.

Sincerely,

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-362

Enclosure:

Safety Evaluation cc w/encl: See next page

R. Rosenblum proposed by SCE in accordance with 50.55a(a)(3)(i) of Title 10 of Code of Federal Regulations.

All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

The staff's safety evaluation is enclosed.

Sincerely,

/RA/

David Terao, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-362

Enclosure:

Safety Evaluation cc w/encl: See next page DISTRIBUTION:

PUBLIC RidsOgcRp LPLIV r/f RidsAcrsAcnwMailCenter RidsNrrDorlLpl4 SSheng, NRR RidsRgn4MailCenter DCullison, RIV Plants RidsNrrPMNKalyanam RidsNrrLALFeizollahi Accession No.: ML070310154

• Minor editorial changes made in staff supplied SE.

OFFICE NRR/LPL4/PM NRR/LPL4/LA DCI/CFEB* OGC - NLO NRR/LPL4/BC NAME NKalyanam LFeizollahi KGruss JRund DTerao DATE 1/31/07 1/31/07 1/8/07 2/12/07 2/15/07 OFFICIAL RECORD COPY

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF ISI-3-22 SOUTHERN CALIFORNIA EDISON SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3 DOCKET NO. 50-362

1.0 INTRODUCTION

By letter dated May 11, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061360065), and supplemented by letters dated August 15 (ADAMS Accession No. ML062290241) and October 20, 2006 (ADAMS Accession No. ML062980205), Southern California Edison (SCE, the licensee) submitted Relief Request ISI-3-22 seeking relief from certain American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) requirements and to employ an embedded flaw inside diameter (ID) structural weld inlay repair process (ID weld repair) to repair Reactor Vessel Head Penetration (RVHP) Control Element Drive Mechanism (CEDM) No. 56 at San Onofre Nuclear Generating Station, Unit 3 (SONGS 3), during the Cycle 14 outage if an inspection indicates that the previously repaired flaw shows crack growth.

The inservice inspection (ISI) of the ASME Code Class 1, Class 2, and Class 3 components is to be performed in accordance with Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," of the ASME Code and applicable edition and addenda as required by 50.55a(g) of Title 10 of the Code of Federal Regulations (10 CFR). Paragraph 50.55a(a)(3) of 10 CFR states, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (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) shall meet the requirements, except the design and access provisions and 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 10-year ISI 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 third 10-year ISI interval for SONGS 3, began in August 2003 and will end in August 2013. The ISI Code of record for the SONGS 3, third 10-year ISI interval is the 1995 Edition with 1996 Addenda. The components (including

supports) may meet the requirements set forth in subsequent editions and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein and subject to Commission approval.

2.0 REGULATORY EVALUATION

2.1 Components for which Relief is Requested This relief request applies to SONGS 3 RVHP CEDM No. 56, an ASME Code Class 1 component.

2.2 Code Requirements (As stated by the licensee)

ASME XI, IWA-4410(a) states the repair/replacement activities, such as metal removal and welding, shall be performed in accordance with the Owners Requirements and the original Construction Code of the component or system.

The applicable Construction Code is ASME III, 1971 Edition, through the Summer 1971 Addenda.

BASE METAL DEFECT REPAIRS ASME III, NB-4131 states that defects in base metals, such as the RPVH

[penetration] tubes, may be eliminated or repaired by welding, provided the defects are removed, repaired and examined in accordance with the requirements of NB-2500.

ASME III, NB-2538 addresses elimination of base material surface defects and specifies defects are to be removed by grinding or machining. Defect removal must be verified by a magnetic particle or liquid penetrant examination using acceptance criteria of NB-2545 or NB-2546. If the removal process reduces the section thickness below the NB-3000 design thickness, then repair welding per NB-2539 is to be performed.

ASME III, NB-2539.1 addresses removal of defects and requires defects be removed or reduced to an acceptable size by suitable mechanical or thermal methods.

ASME III, NB-2539.4 provides the rules for examination of the base material repair welds and specifies they shall be examined by the magnetic particle or liquid penetrant methods with acceptance criteria per NB-2545 and NB-2546.

Additionally, if the depth of the repair cavity exceeds the lesser of 3/8 inch or 10 percent of the section thickness, the repair weld shall be examined by the radiographic method using the acceptance criteria of NB-5320.

2.3 Relief Requested (As stated by the licensee)

Relief is requested from the requirements of ASME XI, IWA-4410(a), to perform repairs on the RPVH penetrations per the rules of the Construction Code.

Relief is requested from the requirements in ASME III, NB-4131, NB-2538 and NB-2539.1 to eliminate base material defects prior to repair welding.

Relief is requested to use substitute examination methods in lieu of those specified in NB-2539.4 for the following cases:

• In the case of embedded flaw welds on the inside diameter (ID) surface of the penetration tubes, eddy current and ultrasonic examinations will be performed on the overlay repair weld which are surface and volumetric examinations but are different methods than specified in NB-2539.4. The proposed inspection methods are consistent with the NRC Order EA-03-009 [Issuance of Order Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactor] examinations that originally detected the indication. IWB 3600 acceptance criteria are used for the embedded flaws as described in WCAP 15987-P revision 2 [Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations.]

2.4 Licensees Proposed Alternative and Basis for Use The proposed ID weld repair reduces the depth of an outer-diameter (OD) initiated flaw by excavation of 3/16 inch to 1/4 inch from the ID surface. This excavation will be restored to the original ID surface contour by structural weld such that the completed repair will have an embedded flaw less than 72-percent through-wall. The repair weld will extend circumferentially towards the 90-degrees and 270-degrees orientations where penetration tube tensile stresses are minimum.

The repair design, implementation, and inspection is equivalent to the ID weld repairs described in WCAP-15987-P, Revision 2. SCE has performed an ASME Code reconciliation to verify that the bases contained in the WCAP are applicable to SONGS 3. Further, the weld materials, design, and techniques employed for the approved seal weld repair conducted in 2004 are equivalent to the process required for a structural base metal repair weld.

As to the basis for this application, the licensee states partly:

The proposed modification to the approved repair process reflects a new application of that repair. However, the efficacy of the repair to achieve an appropriate level of safety and quality is not adversely affected by this new application.

The proposed modification to examination methods for the ID inlay repair weld has been demonstrated to be adequate for flaw detection and sizing for similar repair geometries.

...Calculations performed in support of SCE Relief Requests ISI-3-13 and ISI-3-21 have shown that the existing flaw will remain within Code margins for at least 10 years.

As described in Attachment 1 to this enclosure, the proposed repair will reduce the indication dimensions and will extend the assured repair life. The methodology used in Attachment Structural Evaluation of the Proposed Embedded Flaw Repair of the Indication in Reactor Vessel Head Penetration No. 56 at SONGS Unit 3, for determining flaw acceptance is consistent with the methodology approved in the Westinghouse Topical Report WCAP 15987-P Revision 2.

The thickness of the weld used to embed the flaw has been set to provide a permanent embedment of the flaw. As shown in Attachment 1 [to SCE letter dated May 11, 2006],

the embedded flaw process imparts less residual stresses than weld repair following the complete removal of the flaw.

Therefore, the embedded flaw repair process is considered to be an alternative to Code requirements that provides an acceptable level of quality and safety, as required by 10 CFR 50.55a(a)(3)(i).

3.0 TECHNICAL EVALUATION

In letter dated May 11, 2006, the licensee requested NRC approval to repair RVHP CEDM No. 56 at SONGS 3 during the Cycle 14 outage if an inspection reveals growth of a previously repaired crack. The proposed repair involves the application of a structural weld inlay on the ID of the No. 56 penetration to repair a flaw located in the base material on the OD of the nozzle at and below the J-groove weld level. The measured length and depth of the flaw during the Cycle 13 outage are 1.96 inches and 0.513 inch, respectively. The ultrasonic testing (UT) measurement uncertainty in the through-wall depth was 0.02 inch. The flaw was originally identified in the Cycle 12 outage as a weld defect having no surface breaking indications with a through-wall depth of 0.44 inch. In the Cycle 13 outage inspection, the flaw was inspected with improved UT equipment and was found to have grown 0.07 inch in one operating cycle under the primary water stress-corrosion cracking (PWSCC) environment. As a result, it was repaired using an embedded flaw process in accordance with that described in WCAP-15987-P, Revision 2. The safety evaluation (SE) on this topical report was issued on July 3, 2003 (ADAMS Accession No. ML031840237).

Since the embedded flaw repair processes described in WCAP-15987-P, Revision 2, have been approved, this staff evaluation focuses on the plant-specific applicability of the WCAP to the current relief request. The licensee identified four exceptions to WCAP-15987-P, Revision 2:

1. Application of an ID weld inlay to mitigate a previously repaired OD-initiated PWSCC flaw,

2. Crediting the inlay weld as a structural weld repair instead of a seal weld,

3. A potential change in residual surface stresses, and future PWSCC susceptibility associated with a repair weld approximately 4/16-inch deep compared to an approved weld depth of at least 3/16-inch deep,

4. A potential decrease in the ability to accurately monitor the residual indication using ultrasonic techniques deployed from the ID surface.

Exception 1 is acceptable because WCAP-15987-P, Revision 2, does not exclude the application of the embedded flaw process to repaired penetrations. Exception 2 is necessary and acceptable, because the crack front now is located at the inlay Alloy 52 weld instead of the Alloy 600 RVHP. Further, the difference in the material properties (Youngs modulus, yield strength, and ultimate strength) specified in the ASME Code for Alloy 690 (the equivalence of Alloy 52 weld) and Alloy 600 are not significant. Hence, the fracture mechanics analysis supporting WCAP-15987-P, Revision 2, with the crack front located at the Alloy 600 RVHP, applies to this relief request, with crack front located at the inlay Alloy 52 weld. The licensee provided in its May 11, 2006, application of the following additional information to supplement the WCAP-15987-P, Revision 2, methodology in support of the relief request: (1) stress and fracture mechanics analysis, (2) laboratory testing, and (3) field experience. The stress and fracture mechanics analysis shows that the ID inlay patch is designed to have its edges located at the low stress or the compressive stress region of the RVHP. The laboratory testing and field experience were provided to support generic Westinghouse weld inlay repairs. This generic information supplements the justification for crediting the inlay weld as a structural weld.

Exception 3 concerns the potential change in residual stresses. The embedded flaw process provides a protective layer of material between the PWSCC environment and the crack tip.

For RVHP No. 56, the layer of PWSCC-resistant Alloy 52 weld will cover enough wetted surface of the ID surface of the nozzle to envelop the area with the indication. This repair eliminates contact between the PWSCC environment and the embedded flaw and should prevent growth of the flaw due to PWSCC. Therefore, the residual stresses, which affect the PWSCC crack growth rate, are irrelevant here. Although residual stresses are considered in the fatigue crack growth calculation, their effect on crack growth is an order of magnitude lower than the PWSCC crack growth. This demonstrates that the potential change in residual stresses is not a concern, and the proposed repair is appropriate for at least one operating cycle.

The last exception of the current relief request from WCAP-15987-P, Revision 2, is the change caused by applying UT techniques from the ID surface. The licensee states that the ability to accurately monitor the residual indication using time of flight tip diffraction ultrasonic examination from the ID surface and eddy current methods for this type of repair has been addressed and has undergone third-party review with satisfactory results. These inspections employ the same Westinghouse tooling and techniques that are in use at SONGS for RVHP inspections required by NRC Order EA-03-009. For inspection of RVHP No. 56, the licensee will use a new mockup with electron machine discharge (EDM) notches of depths 0.02 inch and 0.04 inch in the weld and base metal to demonstrate that the background noise associated with the ID weld repair does not degrade the flaw detection capability. Separately, EDM notches will be made in the base material and beneath the weld inlay repair region to demonstrate the flaw detection capability and the ability to detect flaw growth into a weld repair. The staff questioned whether use of EDM notches, instead of cold isostatically pressed notches, could achieve the stated objectives. To relieve the staff concern, the licensee committed to provide, in a letter dated October 20, 2006:

1) a comparison of the proposed inspection technique to the technique used in the Electric Power Research Institute (EPRI) Material Reliability Program (MRP)-89 demonstration report, and

2) the results of the Westinghouse technical justification analysis[, which will demonstrate]

the capability to monitor the residual indication and detect future growth in a manner consistent with the original EPRI - MRP demonstration.

The NRC staffs concern of the proposed nondestructive examination is resolved because the commitments can confirm that the installed inlay weld is free of defects and the embedded crack, if after a cycle of operation the crack has not grown into the inlay weld,. Therefore, based on the above discussions on the four exceptions, the NRC staff considers Relief Request ISI-3-22 acceptable.

Relief Request ISI-3-22 will only be used if the licensee finds that the flaw growth in the RVHP No. 56 nozzle base material exceeds 0.02 inch during any outage in the third 10-year ISI interval at SONGS 3. If flaw growth occurs, it must be due to something not anticipated and not being addressed in WCAP-15987-P, Revision 2. Consequently, the proposed repair may not be considered as a permanent repair. To respond to this consideration, the licensee, in a letter dated August 15, 2006, revised the duration of the proposed alternative to one cycle following repair, to be used in the Cycle 15 refueling outage. The NRC staff, thus, approves Relief Request ISI-3-22 for the requested duration.

Again, if an inspection during an outage indicates that the crack depth growth is less than 0.02 inch (the measurement resolution capability), Relief Request ISI-3-22 may not be used.

4.0 CONCLUSION

The NRC staff has reviewed the licensees proposal to allow for the use of embedded flaw ID structural weld inlay repair process as an alternative to the ASME Code requirements for the repair of the SONGS 3 RVHP CEDM No. 56, in accordance with 10 CFR 50.55a(a)(3)(i).

Based on its review, the NRC staff finds that the licensees proposal provides an acceptable level of quality and safety for a duration of one cycle of operation following repair, to be used in the Cycle 15 refueling outage. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the proposed alternative to the flaw repair requirements of IWA-4410(a) and related requirements listed in Section 2.2 of this SE, of ASME Code, Sections III and XI, at SONGS 3 for one cycle of post-repair operation during the third 10-year ISI interval. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.

Principal Contributor: S. Sheng Date: February 15, 2007

San Onofre Nuclear Generating Station Units 2 and 3 cc:

Mr. Daniel P. Breig Director, Radiologic Health Branch Southern California Edison Company State Department of Health Services San Onofre Nuclear Generating Station P.O. Box 997414, MS 7610 P.O. Box 128 Sacramento, CA 95899-7414 San Clemente, CA 92674-0128 Resident Inspector/San Onofre NPS Mr. Douglas K. Porter, Esquire c/o U.S. Nuclear Regulatory Commission Southern California Edison Company Post Office Box 4329 2244 Walnut Grove Avenue San Clemente, CA 92674 Rosemead, CA 91770 Mayor Mr. David Spath, Chief City of San Clemente Division of Drinking Water and 100 Avenida Presidio Environmental Management San Clemente, CA 92672 P.O. Box 942732 Sacramento, CA 94234-7320 Mr. James T. Reilly Southern California Edison Company Chairman, Board of Supervisors San Onofre Nuclear Generating Station County of San Diego P.O. Box 128 1600 Pacific Highway, Room 335 San Clemente, CA 92674-0128 San Diego, CA 92101 Mr. James D. Boyd, Commissioner Mark L. Parsons California Energy Commission Deputy City Attorney 1516 Ninth Street (MS 31)

City of Riverside Sacramento, CA 95814 3900 Main Street Riverside, CA 92522 Mr. Ray Waldo, Vice President Southern California Edison Company Mr. Gary L. Nolff San Onofre Nuclear Generating Station Assistant Director - Resources P.O. Box 128 City of Riverside San Clemente, CA 92764-0128 3900 Main Street Riverside, CA 92522 Mr. Brian Katz Southern California Edison Company Regional Administrator, Region IV San Onofre Nuclear Generating Station U.S. Nuclear Regulatory Commission P.O. Box 128 611 Ryan Plaza Drive, Suite 400 San Clemente, CA 92764-0128 Arlington, TX 76011-8064 Mr. Steve Hsu Mr. Michael R. Olson Department of Health Services San Diego Gas & Electric Company Radiologic Health Branch 8315 Century Park Ct. CP21G MS 7610, P.O. Box 997414 San Diego, CA 92123-1548 Sacramento, CA 95899 Mr. A. Edward Scherer Southern California Edison Company San Onofre Nuclear Generating Station P.O. Box 128 San Clemente, CA 92674-0128 March 2006